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atw International Journal for Nuclear Power 2020-08/09

atw - International Journal for Nuclear Power | 08/09.2020 Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information. www.nucmag.com

atw - International Journal for Nuclear Power | 08/09.2020

Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information.

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nucmag.com<br />

<strong>2020</strong><br />

8/9<br />

ISSN · 1431-5254<br />

24.– €<br />

Current Status and<br />

Prospects of <strong>Nuclear</strong> <strong>Power</strong><br />

Plant Decommissioning in<br />

the Republic of Korea<br />

Actual Research and Development<br />

Activities in the Field of Dismantling<br />

Neutronic Study of CAREM-25<br />

Advanced Small Modular Reactor<br />

Using Monte Carlo Simulation


Competence <strong>for</strong><br />

<strong>Nuclear</strong> Services<br />

Operational Waste and D&D<br />

Spent Fuel Management<br />

<strong>Nuclear</strong> Casks<br />

Calculation Services and Consulting<br />

Waste Processing Systems and Engineering<br />

GNS Gesellschaft für Nuklear-Service mbH<br />

Frohnhauser Str. 67 · 45127 Essen · Germany · info@gns.de · www.gns.de


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Implications of the Corona/Covid-19 Crises<br />

on Energy Supply and <strong>Nuclear</strong> <strong>Power</strong><br />

Dear Reader, What began in the early days of this year in Wuhan, China as a local outbreak of a newer <strong>for</strong>m of<br />

Corona virus has now and <strong>for</strong> months on end taken hold of the world.<br />

Our lives are affected by the Corona virus like hardly any<br />

other crisis be<strong>for</strong>e, in private as well as in economy, society,<br />

social life and politics. Restrictions off our daily freedom of<br />

movement are directly noticeable, and economic life has<br />

been badly hit with slumps in gross national product of up<br />

to 13 % in individual OECD countries and around 5 % of<br />

the entire global economy.<br />

The energy supply is also affected. This is clearly<br />

­reflected in the global oil prices. At the beginning of the<br />

year, around US$ 52 per barrel was still being quoted on<br />

the spot market <strong>for</strong> the US grade WTI (West Texas Intermediate).<br />

In the course of the Corona crisis, the price<br />

slumped to US$ 13 and even reached negative values.<br />

Currently, the price of crude oil has stabilized at around<br />

US$ 40. The <strong>International</strong> Energy Agency (IEA) notes a<br />

year-on-year decline in global crude oil demand from<br />

8.1 million barrels a day to currently around 91.9 million<br />

barrels a day, or around minus 9 percent. Electricity<br />

consumption has fallen dramatically in some countries,<br />

particularly in times of complete lockdown and the massive<br />

cutback in all industrial production. In the UK and India,<br />

the year-on-year decline in consumption was thus almost<br />

30 percent, and overall in most OECD countries average<br />

demand was at the lower level of the comparable Sundays<br />

in the previous year, i.e. around 10 percent lower. With the<br />

gradual easing of the complete lockdown from mid-<strong>2020</strong><br />

onwards, electricity demand rose again and in August<br />

reached a level around 5 % below the previous year’s<br />

­figures.<br />

One aspect that is hardly noticed in the overall situation,<br />

as already noted here in issue 5 (<strong>2020</strong>) on the same subject,<br />

is the security of electricity supply. This is because hardly<br />

anyone is aware of how important a secure power supply<br />

was and is <strong>for</strong> communication, which is based almost<br />

entirely on electronic channels, in times of a lack of social<br />

“presence”. Neither telephone, whether line or mobile, nor<br />

the many types of internet communication can do without<br />

electricity. In countries with well-developed infrastructure<br />

and largely existing precautionary measures <strong>for</strong> dealing<br />

with large-scale restrictions caused by infectious diseases,<br />

the continued stable supply of electricity was an expression<br />

of the responsible and <strong>for</strong>ward-looking action of companies<br />

and their employees.<br />

But even if Corona still has us in its grip now, the<br />

expected time after that and the prospects <strong>for</strong> energy<br />

supply and the contribution of nuclear energy should<br />

perhaps be considered all the more intensively than be<strong>for</strong>e.<br />

Let’s take a look at the current situation of nuclear<br />

­power, worldwide: As of August <strong>2020</strong>, 440 nuclear power<br />

plant plants are in operation worldwide. With a gross<br />

capacity of around 400,000 MW, they contribute to around<br />

11 % of global electricity production and thus also to<br />

around one third of low-emission technologies. In terms of<br />

secure supply, i.e. uninfluenced by external, natural<br />

conditions and technically plannable, nuclear power is the<br />

technological leader thanks to its high flexibility, both in<br />

terms of the load gradient, i.e. the speed of ramp-up and<br />

ramp-down, and in terms of the scope of available<br />

capacities, i.e. the available megawatts. This is also<br />

demonstrated today by many plants that support the<br />

integration of renewables in grids with high shares of this<br />

volatile generation. But other results from the 2019<br />

operating year are also impressive. For example, the<br />

97 ­nuclear power plants currently in operation in the<br />

United States reached a further peak value with a<br />

­availability of 93.4 percent (based on gross rated output).<br />

The Taishan 1 EPR reactor with 1,750 MW gross capacity<br />

in its first full year of operation not only led the way in<br />

2019 with around 12.99 billion kilowatt hours of ­electricity<br />

generated, but also slightly exceeded the previous<br />

annual top result of Chooz B1, 1,560 MW, France, with<br />

12.97 ­billion kWh from 2012. All in all: nuclear energy still<br />

has something to offer worldwide.<br />

However, action is needed when it comes to medium<br />

and long-term prospects. The nuclear power plant fleet is<br />

also aging. Although the plants do not reach their ­primarily<br />

technical and economic design limits as quickly as other<br />

technologies, and runtimes of 80 or 100 years are not only<br />

under discussion but a reality internationally, the share of<br />

nuclear energy will decline without action as the world’s<br />

hunger <strong>for</strong> energy continues to rise. The 54 nuclear power<br />

plant plants currently under construction will bring about<br />

75,000 MW of nuclear power plant capacity to the grid in<br />

the coming years due to their capacity size. However,<br />

governments and companies should also quickly tackle<br />

the 104 other projects that are currently in an advanced<br />

planning stage.<br />

The socio-economic benefits of nuclear power, i.e.<br />

secure, competitive and subsidy-free jobs, af<strong>for</strong>dable,<br />

reliable electricity <strong>for</strong> consumers and low-emission power<br />

generation, should be addressed to policy makers. <strong>Nuclear</strong><br />

energy is an investment in the future. It can be safely<br />

implemented in operation with responsibility. The radioactive<br />

residues or waste are not a problem. They are a technical<br />

and scientific challenge <strong>for</strong> the current generation.<br />

And nuclear energy has two further perspectives: One<br />

is the small and medium power reactors, SMR <strong>for</strong> short,<br />

with a wide range of possible technologies. For many years<br />

in the past, SMRs were part vision, part speculation. Today<br />

they are part of the energy future in a whole range of<br />

concrete projects <strong>for</strong> pilot plants - from China and Russia to<br />

the UK, the USA and Canada. And nuclei can not only be<br />

split, they can also be reassembled, in nuclear fusion. This<br />

is where the international fusion project ITER in Cadarche<br />

in the south of France recently reached another milestone<br />

in its construction: with notable political participation, the<br />

interior construction of the ring-shaped nuclear fusion<br />

reactor began at the end of July <strong>2020</strong>. The facility is<br />

expected to be completed by the middle of this decade,<br />

paving the way <strong>for</strong> the energy-producing fusion reactor.<br />

Christopher Weßelmann<br />

– Editor in Chief –<br />

379<br />

EDITORIAL<br />

Editorial<br />

Implications of the Corona/Covid-19 Crises on Energy Supply and <strong>Nuclear</strong> <strong>Power</strong>


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

380<br />

EDITORIAL<br />

Auswirkungen der Corona/Covid-19-Krise<br />

auf Energieversorgung und Kernenergie<br />

Liebe Leserinnen, liebe Leser, was in den ersten Tagen dieses Jahres in Wuhan, China als lokaler Ausbruch einer<br />

neueren Form des Coronavirus begann, hat inzwischen und dies zudem über Monate hinweg die Welt fest im Griff.<br />

Unser Leben wird wie kaum eine andere Krise zuvor vom<br />

­Coronavirus beeinflusst, im privaten Bereich ebenso wie in<br />

Wirtschaft, Gesellschaft, sozialem Leben und Politik. Einschränkungen<br />

unserer täglichen Bewegungsfreiheit sind<br />

direkt spürbar, das Wirtschaftsleben ist schwer getroffen<br />

mit Einbrüchen beim Bruttosozial produkt von bis zu 13 %<br />

in einzelnen OECD-Ländern um rund 5 % der gesamten<br />

Weltwirtschaft.<br />

Auch die Energieversorgung ist betroffen. Offensichtlich<br />

zeigt sich dies bei den weltweiten Ölpreisen. Anfang<br />

des Jahres wurden auf dem Spotmarkt für die U.S.-<br />

amerikanische Sorte WTI (West Texas Intermediate) noch<br />

rund 52 US$ pro Barrel notiert. Im Laufe der Coronakrise<br />

brach der Preis auf 13 US$ ein und erreichte sogar negative<br />

­Werte. Aktuell hat sich der Rohölpreis bei rund 40 US$<br />

­stabilisiert. Beim weltweiten Rohölbedarf notiert die<br />

­<strong>International</strong>e Energieagentur IEA beim Vorjahresvergleich<br />

einen Rückgang von 8,1 auf aktuell rund 91,9 Millionen<br />

­Barrel pro Tag, also etwa 9 Prozent. Der Stromverbrauch verzeichnete<br />

teils dramatische Einbrüche, vor allem in Zeiten<br />

des vollständigen Lockdowns und dem damit ver bundenen<br />

massiven Zurückfahren jeglicher Produktion in der Industrie.<br />

In Großbritannien und Indien lag der Ver­brauchsrückgang<br />

im Vorjahresvergleich so bei fast 30 % und insgesamt<br />

lag in den meisten Ländern der OECD der durchschnittliche<br />

Bedarf auf dem niedrigeren Niveau der vergleichbaren<br />

Vorjahres-Sonntage, also ca. 10 % niedriger. Mit allmäh­licher<br />

Lockerung des vollständigen Lockdowns ab Jahresmitte<br />

<strong>2020</strong> stieg der Strombedarf wieder an und erreichte im<br />

­August ein Niveau rund 5 % unter den ­Vorjahreswerten.<br />

Ein in der Gesamtsituation kaum wahrgenommener<br />

­Aspekt, wie an dieser Stelle schon in der Ausgabe 5 (<strong>2020</strong>)<br />

zum gleichen Themenkomplex angemerkt, ist die Versorgungssicherheit<br />

bei der Stromversorgung. Denn kaum ist<br />

breit bewusst geworden, wie wichtig in Zeiten fehlender<br />

sozialer „Präsenzkontakte“ die sichere Stromversorgung für<br />

die quasi vollständig auf elektronischen Wegen basierende<br />

Kommunikation war und ist. Weder Telefon, ob Festnetz<br />

oder Mobil, noch die vielfältigen Formen der Internetkommunikation<br />

kommen ohne Strom aus. In den Ländern<br />

mit gut ausgebauter Infrastruktur und weitgehend vorhandenen<br />

Vorsorgemaßnahmen im Umgang mit großflächigen<br />

Einschränkungen durch Infektionskrankheiten<br />

war die weiterhin stabile Stromversorgung Ausdruck für das<br />

verantwortungsvolle und vorausschauende Handeln der<br />

Unternehmen und ihrer Mitarbeitenden.<br />

Aber auch wenn uns Corona jetzt noch weiterhin im<br />

Griff hat, sollte über die erwartete Zeit danach und die<br />

Perspektiven der Energieversorgung und dem Beitrag<br />

der Kernenergie vielleicht umso intensiver als vorher nachgedacht<br />

werden.<br />

Dazu ein Blick auf die aktuelle Situation der Kernkraft,<br />

weltweit: Aktuell sind mit Stand August <strong>2020</strong> weltweit<br />

440 Kernkraftwerksblöcke in Betrieb. Sie tragen mit rund<br />

400.000 MW Bruttoleistung zu rund 11 % der weltweiten<br />

Stromproduktion bei und damit auch etwa zu einem Drittel<br />

der emissionsarmen Technologien. Bei der gesicherten<br />

­Versorgung, d. h. von äußeren, natürlichen Gegebenheiten<br />

unbeeinflusst und technisch planbar, ist die Kernenergie<br />

durch ihren hohen Grad an Flexibilität, sowohl was den<br />

Lastgradienten, also die Geschwindigkeit des Hoch- und<br />

­Abfahrens, betrifft, als auch beim Umfang der verfügbaren<br />

Editorial<br />

Implications of the Corona/Covid-19 Crises on Energy Supply and <strong>Nuclear</strong> <strong>Power</strong><br />

Kapazitäten, also den verfügbaren Megawatt, technologischer<br />

Spitzenreiter. Dies zeigen auch heutzutage viele<br />

Anlagen, die die Integration von Erneuerbaren in Netzen mit<br />

hohen Anteilen dieser volatilen Erzeugung unterstützen.<br />

Doch auch weitere Ergebnisse aus dem Betriebsjahr 2019<br />

beeindrucken. So haben die in den USA laufenden 97 Kernkraftwerke<br />

mit einer Arbeitsverfügbarkeit von 93,4 Prozent<br />

(bezogen auf die Brutto-Nennleistung) einen weiteren<br />

Spitzenwert erreicht. Der EPR-Reaktor Taishan 1 mit<br />

1.750 MW Bruttoleistung war im ersten vollen Betriebsjahr<br />

mit rund 12,99 Milliarden erzeugten Kilowattstunden<br />

Strom nicht nur führend in 2019, sondern hat auch das<br />

­bisherige Jahres-Spitzenergebnis von Chooz B1, 1.560 MW,<br />

Frankreich, mit 12,97 Mrd. kWh aus dem Jahr 2012, leicht<br />

übertroffen. In Summe: Kernenergie hat weltweit noch<br />

etwas zu bieten.<br />

Was allerdings die mittel und langfristigen Perspektiven<br />

betrifft, ist Handeln er<strong>for</strong>derlich. Auch der Kernkraftwerkspark<br />

altert. Zwar kommen die Anlagen nicht so schnell an<br />

ihre vor allem technisch-wirtschaftliche Auslegungsgrenze<br />

wie andere Technologien, und international sind Laufzeiten<br />

von 80 bzw. 100 Jahren nicht nur in der Diskussion sondern<br />

Realität, aber dennoch wird der Kernenergieanteil ohne<br />

Handeln mit dem weiter steigenden Energiehunger in der<br />

Welt sinken. Die 54 aktuell in Bau befindlichen Kernkraftwerksblöcke<br />

werden aufgrund ihrer Leistungsgröße in den<br />

kommenden Jahren rund 75.000 MW Kernkraftwerkskapazität<br />

ans Netz bringen. Die Regierungen und Unternehmen<br />

sollten aber darüber hinaus die 104 weiteren<br />

Projekte, die sich heute in einem weit <strong>for</strong>tgeschrittenen<br />

­Planungsstadium befinden, zügig angehen.<br />

Mit den sozioökonomischen Vorteilen der Kernenergie,<br />

also sicheren, wettbewerbsfähigen und subventionsfreien<br />

Arbeitsplätzen, preisgünstigem, versorgungssicherem Strom<br />

für den Verbraucher und einer emissionsarmen Energieerzeugung<br />

sollte an die politischen Entscheidungsträger<br />

appelliert werden. Kernenergie ist eine Zukunftsinvestition.<br />

Sie ist mit Verantwortung sicher im Betrieb umsetzbar.<br />

Die radioaktiven Reststoffe oder Abfälle sind kein Problem.<br />

Sie sind eine technisch-wissenschaftlich lösbare Heraus<strong>for</strong>derung<br />

für die aktuelle Generation.<br />

Und Kernenergie hat zwei weitere Perspektiven: Diese<br />

eine sind die Reaktoren kleiner und mittlerer Leistung,<br />

kurz SMR, mit einem breiten Spektrum von möglichen<br />

­Technologien. Über viele Jahre der Vergangenheit waren<br />

SMR teils Vision, teils Spekulation. Heute sind sie in einer<br />

ganzen Reihe von konkreten Projekten für Pilotanlagen –<br />

von China über Russland bis hin zu Großbritannien, die USA<br />

und Kanada – Teil der Energiezukunft. Und Kerne lassen<br />

sich nicht nur spalten, sie lassen sich auch neu zusammenfügen,<br />

bei der Kernfusion. Hier hat das internationale<br />

­Fusionsprojekt ITER im Südfranzösischen Cadarche kürzlich<br />

einen weiteren Meilenstein bei der Errichtung erreicht:<br />

Unter namhafter politischer Teilnahme begann Ende Juli<br />

<strong>2020</strong> der Innenausbau des ringförmigen Kernfusionsreaktors.<br />

Mitte dieses Jahrzehnts soll die Anlage fertiggestellt<br />

sein und den Weg zum Energie liefernden Fusionsreaktor<br />

bereiten.<br />

Christopher Weßelmann<br />

– Chefredakteur –


Kommunikation und<br />

Training für Kerntechnik<br />

Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik?<br />

Wählen Sie aus folgenden Themen: Dozent/in Termin/e Ort<br />

3 Atom-, Vertrags- und Exportrecht<br />

Atomrecht – Das Recht der radioaktiven Reststoffe und Abfälle RA Dr. Christian Raetzke 20.10.<strong>2020</strong> Berlin<br />

Export kerntechnischer Produkte und Dienstleistungen –<br />

Chanchen und Regularien<br />

Atomrecht – Was Sie wissen müssen<br />

Atomrecht – Ihr Weg durch Genehmigungs- und<br />

Aufsichtsverfahren<br />

RA Kay Höft M.A. (BWL) 04.11.<strong>2020</strong> Berlin<br />

RA Dr. Christian Raetzke<br />

Akos Frank LL. M.<br />

11.11.<strong>2020</strong> Berlin<br />

RA Dr. Christian Raetzke 20.01.2021 Berlin<br />

3 Kommunikation und Politik<br />

Public Hearing Workshop –<br />

Öffentliche Anhörungen erfolgreich meistern<br />

Dr. Nikolai A. Behr 10.11. - 11.11.<strong>2020</strong> Berlin<br />

3 Rückbau und Strahlenschutz<br />

In Kooperation mit dem TÜV SÜD Energietechnik GmbH Baden-Württemberg:<br />

3 <strong>Nuclear</strong> English<br />

Stilllegung und Rückbau in Recht und Praxis<br />

Das Strahlenschutzrecht und<br />

seine praktische Umsetzung<br />

Dr. Stefan Kirsch<br />

RA Dr. Christian Raetzke<br />

Dr. Maria Poetsch<br />

RA Dr. Christian Raetzke<br />

23.<strong>09</strong>. - 24.<strong>09</strong>.<strong>2020</strong> Berlin<br />

29.10. - 30.10.<strong>2020</strong> Berlin<br />

English <strong>for</strong> the <strong>Nuclear</strong> Industry Angela Lloyd 16.03. - 17.03.2021 Berlin<br />

3 Wissenstransfer und Veränderungsmanagement<br />

Erfolgreicher Wissenstransfer in der Kerntechnik –<br />

Methoden und praktische Anwendung<br />

Veränderungsprozesse gestalten –<br />

Heraus<strong>for</strong>derungen meistern, Beteiligte gewinnen<br />

Dr. Tanja-Vera Herking<br />

Dr. Christien Zedler<br />

Dr. Tanja-Vera Herking<br />

Dr. Christien Zedler<br />

05.10. - 06.10.<strong>2020</strong> Berlin<br />

24.11. - 25.11.<strong>2020</strong> Berlin<br />

Haben wir Ihr Interesse geweckt? 3 Rufen Sie uns an: +49 30 498555-30<br />

Kontakt<br />

INFORUM Verlags- und Verwaltungs gesellschaft mbH ı Robert-Koch-Platz 4 ı 10115 Berlin<br />

Petra Dinter-Tumtzak ı Fon +49 30 498555-30 ı Fax +49 30 498555-18 ı Seminare@KernD.de<br />

Die INFORUM-Seminare können je nach<br />

Inhalt ggf. als Beitrag zur Aktualisierung<br />

der Fachkunde geeignet sein.


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

382<br />

CONTENTS<br />

Issue 8/9 | <strong>2020</strong><br />

August/September<br />

Contents<br />

Editorial<br />

Implications of the Corona/Covid-19 Crises<br />

on Energy Supply and <strong>Nuclear</strong> <strong>Power</strong> E/G 379<br />

Inside <strong>Nuclear</strong> with NucNet<br />

South Africa / Policy Announcement Opens Door<br />

to all Types of <strong>Nuclear</strong> Technology 384<br />

Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . .385<br />

Cover:<br />

Dismantling the concrete shell of the<br />

containment of NPP Philippsburg 1.<br />

Courtesy of EnBW Kernkraft GmbH<br />

Contents:<br />

Manipulator Operated Laser Ablation<br />

(MANOLA).<br />

Courtesy of KIT Institute of Technology<br />

Calendar . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .386<br />

Feature | Decommissioning And Waste Management<br />

Current Status and Prospects of <strong>Nuclear</strong> <strong>Power</strong> Plant<br />

Decommissioning in the Republic of Korea 387<br />

Spotlight on <strong>Nuclear</strong> Law<br />

Cost Correction <strong>for</strong> Site Selection Procedures –<br />

It Remains Confusing G 392<br />

Decommissioning and Waste Management<br />

Actual Research and Development Activities<br />

in the Field of Dismantling 394<br />

A Geopolymer Waste Form <strong>for</strong> Technetium, Iodine<br />

and Hazardous Metals 397<br />

Decommissioning of <strong>Nuclear</strong> <strong>Power</strong> Plants: Waste Streams<br />

and Release Measurements 400<br />

Ventilation Concepts <strong>for</strong> <strong>Nuclear</strong> Decommissioning 403<br />

Steam Generator Rip and Ship – a Valuable Contribution<br />

to Decommissioning and Dismantling of <strong>Nuclear</strong> <strong>Power</strong> Plants 406<br />

Casks and Cask Stacks in Interim Storage Facilities<br />

under Earthquake Loads 4<strong>09</strong><br />

Radioactivity Calculation of the Concrete Shielding<br />

of the Petten LFR and the Dodewaard BWR 414<br />

Quality Assurance and Data Analysis in Automated Radiological<br />

Characterization of Large Soil Volumes 418<br />

Construction of a Dismantling Hall <strong>for</strong> Large Components<br />

at Entsorgungswerk für Nuklearanlagen in Lubmin G 421<br />

Environment and Safety<br />

Current Procedure <strong>for</strong> Determining Release Parameters <strong>for</strong> a Plane<br />

Crash on a <strong>Nuclear</strong> Facility in the Context of Accident Analyses 426<br />

Research and Innovation<br />

Evaluation of MACST Strategies <strong>for</strong> Extended Loss<br />

of AC Electric <strong>Power</strong> Event in OPR1000 <strong>Nuclear</strong> <strong>Power</strong> Plants 430<br />

Neutronic Study of CAREM-25 Advanced Small Modular Reactor<br />

Using Monte Carlo Simulation 435<br />

KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .440<br />

Report<br />

Operating results 2019 441<br />

News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .456<br />

<strong>Nuclear</strong> Today<br />

Europe Can’t Discard <strong>Nuclear</strong> Investment<br />

in Quest <strong>for</strong> a Clean Energy Future 462<br />

G<br />

E/G<br />

= German<br />

= English/German<br />

Imprint 393<br />

Insert: ICOND <strong>2020</strong><br />

Contents


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Feature<br />

Feature | Decommissioning<br />

And Waste Management<br />

383<br />

CONTENTS<br />

387 Current Status and Prospects<br />

of <strong>Nuclear</strong> <strong>Power</strong> Plant Decommissioning<br />

in the Republic of Korea<br />

Joo Hyun Moon<br />

Decommissioning and Waste Management<br />

394 Actual Research and Development Activities<br />

in the Field of Dismantling<br />

Sascha Gentes and Nadine Gabor<br />

397 A Geopolymer Waste Form <strong>for</strong> Technetium, Iodine<br />

and Hazardous Metals<br />

Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg<br />

400 Decommissioning of <strong>Nuclear</strong> <strong>Power</strong> Plants:<br />

Waste Streams and Release Measurements<br />

Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann<br />

Research and Innovation<br />

435 Neutronic Study of CAREM-25 Advanced Small Modular Reactor<br />

Using Monte Carlo Simulation<br />

Saeed Zare Ganjaroodi and Ali Pazirandeh<br />

Report<br />

441 Operating results 2019<br />

Contents


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

384<br />

South Africa / Policy Announcement Opens Door to<br />

all Types of <strong>Nuclear</strong> Technology<br />

Country wants 2,500 MW of new reactors, but financing remains an obstacle<br />

INSIDE NUCLEAR WITH NUCNET<br />

South Africa’s announcement that it is developing a roadmap <strong>for</strong> 2,500 MW of nuclear- powered<br />

generating capacity signals a policy revival that opens the door to all types of technologies and reactor sizes<br />

from 1,000 MW at the higher end to Generation IV small modular reactors (SMRs) that range from 50-300 MW.<br />

The plan is to allow vendors to self-finance 100 % of the cost<br />

which means the national government will not provide any<br />

funding.<br />

Mineral resources and energy minister Gwede Mantashe<br />

said his agency will issue a request <strong>for</strong> in<strong>for</strong>mation to assess the<br />

market with a focus on SMRs. However, he said all options are<br />

being explored and if the market indicates one design is more<br />

af<strong>for</strong>dable and can be built more efficiently, he will go with it.<br />

He did not say when his agency expects a vendor to break<br />

ground, nor did he specify a preference <strong>for</strong> any particular<br />

reactor designs.<br />

He said: “We may give a company a right to develop a<br />

nuclear station on a build, operate, and transfer basis. It<br />

means there is no immediate funding from the state.”<br />

The nuclear plans are a significant development on what<br />

has been a long and sometimes controversial ef<strong>for</strong>t by South<br />

Africa to add to its existing commercial nuclear capacity.<br />

In 2018 plans to expand nuclear capacity by building up to<br />

9,600 MW of new plants were put on hold with nuclear<br />

excluded from an integrated resource plan (IRP) because the<br />

government saw electricity generation from other sources as<br />

cheaper and because there was a lower demand <strong>for</strong> electricity<br />

than <strong>for</strong>ecast in an earlier plan in 2010.<br />

The IRP called <strong>for</strong> nuclear capacity to remain at 1,860 MW<br />

(net) by 2030, which means there will be no change.South<br />

­Africa’s only nuclear station at Koeberg has two pressurised<br />

water reactor units that have been in commercial operation<br />

since 1984 and 1985. Their output accounts <strong>for</strong> 2.5 % of the<br />

country’s energy generation.<br />

The plan was drawn up a year after the Western Cape High<br />

Court said a series of preliminary procurement deals <strong>for</strong> new<br />

nuclear between the government of South Africa and Russia,<br />

China, the US, South Korea and France were illegal.<br />

The court ruled in April 2017 that the procurement ­process<br />

was unconstitutional and illegal as it was not ­sufficiently ­public<br />

and did not involve adequate environmental and financial<br />

­assessments. After theverdict, then energy minister Mmamoloko<br />

Kubayi said the government would go ahead with the signing of<br />

five new intergovernmental agreements with potential international<br />

partners, but this never happened and ambitious plans<br />

<strong>for</strong> up to 9,600 MW of new nuclear ­capacity were dropped.<br />

That plan, put <strong>for</strong>ward by then president Jacob Zuma,<br />

would have seen South Africa sign a deal with Russia’s state<br />

­nuclear corporation Rosatom <strong>for</strong> eight 1,200 MW VVER ­nuclear<br />

reactors at a projected cost of between $30-to-$50 bn. Rosatom’s<br />

terms were that it would provide 50 % of the financing.<br />

The plan died <strong>for</strong> three reasons. The first was that South<br />

Africa couldn’t af<strong>for</strong>d it, even with generous financial terms<br />

from any vendor. The second was that Mr Zuma’s administration<br />

was rife with allegations of corruption and nepotism. The<br />

third was the lack of transparency related to how the procurement<br />

process <strong>for</strong> the deal was done. It was said to have come<br />

about as a result of a “secret” meeting between Mr Zuma and<br />

Russian president Vladimir Putin in a side meeting at a development<br />

conference in Brazil. No tender had been released <strong>for</strong><br />

the project be<strong>for</strong>e that meeting.<br />

Separately, the nation’s economy has been hobbled by a<br />

series of electricity blackouts due to a lack of electrical power<br />

and an aging grid infrastructure.<br />

Eskom, the state-owned utility that also operates Koeberg’s<br />

two reactors, has been thwarted in its requests to raise rates as<br />

the government uses cheap electricity as a way to address<br />

poverty in the country. The government has also declined to<br />

subsidise Eskom directly.<br />

A proposed turnaround plan <strong>for</strong> Eskom has been put on hold<br />

due to the coronavirus pandemic. Eskom’s turnaround plan<br />

includes proposed debt transfer to the government, cost containment,<br />

operational re<strong>for</strong>ms and the company’s unbundling<br />

into three separate entities: generation, transmission and<br />

distribution.<br />

The latest announcement from Mr Mantashe immediately<br />

ran into significant challenges. Shadow minister of mineral<br />

resources and energy Kevin Mileham questioned whether the<br />

100 % vendor financed approach would work and said the timeline<br />

<strong>for</strong> issuing and evaluating tenders might not be feasible.<br />

Additionally, he pointed to the IRP, which he said makes no<br />

mention of nuclear energy at least <strong>for</strong> the next decade. In fact,<br />

the IRP makes brief mention of “preparations <strong>for</strong> nuclear<br />

­energy”, but does not mention a specific level of generating<br />

capacity or a timeline <strong>for</strong> a procurement.<br />

Meanwhile, the need <strong>for</strong> new baseload generation remains.<br />

Africa’s inability to generate enough electricity continues to<br />

hamper economic growth, cutting 2 % to 4 % off GDP every<br />

year, according to the Africa Progress Panel.<br />

<strong>Nuclear</strong> could be part of the solution, but financing ­remains<br />

an issue. <strong>Nuclear</strong> plants are relativelt cheap to operate, but are<br />

expensive to build and require significant upfront capital with a<br />

long wait <strong>for</strong> any return on investment.<br />

The <strong>Nuclear</strong> Industry Association of South Africa (­Niasa)<br />

has proposed six possible funding options <strong>for</strong> new nuclear, but<br />

government officials have suggested the most likely is a “build,<br />

own, operate and transfer” (Boot) model similar to that used<br />

by Russia <strong>for</strong> project including Akkuyu in Turkey.<br />

Niasa had earlier welcomed Mr Mantashe’s announcement<br />

of plans to produce a roadmap <strong>for</strong> new nuclear power plants,<br />

saying it gives the requisite policy certainty which “enables<br />

industry to respond accordingly”.<br />

The government remains cautious. Mr Mantashe said it<br />

will “test the market” to hear what potential investors and<br />

consortia have to say about building a new nuclear facility.<br />

Options being considered include giving a “right to develop<br />

a modular nuclear station on a build, own, operate and transfer<br />

basis,” which means there may be no imme diate call <strong>for</strong><br />

state funding” he said.<br />

“We are going to explore all options, when there is appetite<br />

<strong>for</strong> nuclear in the market we will go ahead with it,”<br />

Mr ­Mantashe added.<br />

An <strong>International</strong> Energy Agency report said electricity<br />

generation from nuclear energy is likely to increase only<br />

slightly in South Africa by 2040, but could see a bigger jump if<br />

policies are enacted to develop the continent’s energy sector.<br />

Those policies would include faster economic expansion<br />

accompanied by the full achievement of key sustainable development<br />

goals by 2030. Those goals, including full access to<br />

electricity, would allow economies to grow “strongly, sustainably<br />

and inclusively”, the IEA said.<br />

Author<br />

NucNet – The Independent Global <strong>Nuclear</strong> News Agency<br />

Editors responsible <strong>for</strong> this story:<br />

David Dalton and Dan Yurman<br />

Avenue des Arts 56 2/C<br />

1000 Bruxelles<br />

www.nucnet.org<br />

www.neutronbytes.com<br />

Inside <strong>Nuclear</strong> with NucNet<br />

South Africa / Policy Announcement Opens Door to all Types of <strong>Nuclear</strong> Technology


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Did you know...?<br />

Enabling deep decarbonisation:<br />

The main drivers of nuclear new build cost and ways to reduce them<br />

In July <strong>2020</strong> the <strong>Nuclear</strong> Energy Agency (NEA) of the OECD released the<br />

report “Unlocking Reductions in the Construction Costs of <strong>Nuclear</strong>: A Practical<br />

Guide <strong>for</strong> Stakeholders” assessing opportunities to reduce construction costs<br />

at the plant-level. The report highlights the role of the different stakeholders<br />

– particularly policymakers. It aims to unlock a positive learning trend in<br />

nuclear construction to enable the full potential of nuclear power as integral<br />

part of future low-carbon energy generation and decarbonisation pathways.<br />

The report con textualises the nuclear challenge with the necessity to reduce<br />

the average carbon intensity of OECD electricity generation from 430 gCO 2 /<br />

kWh in 2016 to less than 50 gCO 2 /kWh in 2050 and with the Sustainable<br />

Development Scenario of the <strong>International</strong> Energy Agency that requires the<br />

commissioning of 15 GW new nuclear capacity globally per year. The main<br />

conclusions and advice derived from the NEA case and study analysis of<br />

which some key parameters are presented in the graphs below are the<br />

following:<br />

p Capitalise on lessons learnt from recent Gen III construction projects. A<br />

window of opportunity exists <strong>for</strong> cost reductions with recently realized Gen III<br />

designs where the nuclear industry and its supply chain have in large part<br />

redeveloped their capabilities in several OECD countries. This requires timely<br />

new-build decisions in the early <strong>2020</strong>s.<br />

p Prioritise design maturity and regulatory stability. Policy support mechanisms<br />

should include requirements <strong>for</strong> design maturity and construction readiness.<br />

The regulatory framework needs to remain stable and predictable throughout<br />

construction.<br />

p Consider committing to a standardised nuclear programme. A standardised<br />

nuclear programme is the most promising way to achieve cost reductions by<br />

the series effect, multi-unit con struction and design and process optimisation.<br />

p Enable and sustain supply chain development and industrial per<strong>for</strong>mance.<br />

New-build ambitions have to consider supply chain constraints and ensure<br />

continuous activity to enable and sustain development in the framework of a<br />

long-term energy policy commitment.<br />

p Foster innovation, talent development and collaboration at all levels. Cost<br />

reduction can arise from innovative nuclear technologies (i.e. SMRs and<br />

Gen-IV reactors). Governments can support this by the timely development of<br />

demonstration projects and an adequate licensing framework.<br />

p Support robust and predictable market and financing frameworks. <strong>Nuclear</strong><br />

projects require long-term government commitments and market regulations.<br />

In addition, financial support is currently essential in western OECD<br />

countries to address market failure and deliver cost-competitive new nuclear<br />

construction.<br />

p Encourage concerted stakeholder ef<strong>for</strong>ts. Governments should create an<br />

environment that fosters a social contract with industry and society to reduce<br />

nuclear construction costs as shown by the <strong>Nuclear</strong> Sector Deal in the United<br />

Kingdom.<br />

p Tailor government involvement to programme needs. Government financial<br />

support <strong>for</strong> fleet programmes will likely decrease gradually with industry<br />

maturity and perceived risk levels falling, but restarting a nuclear programme<br />

likely will require further government support.<br />

DID YOU EDITORIAL KNOW...?<br />

385<br />

Cost breakdown <strong>for</strong> nuclear power<br />

levelised cost of electricity (LCOE)<br />

13 %<br />

11 %<br />

9 %<br />

20 %<br />

47 %<br />

IDC and return of investment are<br />

financing cost, i. e. 67 percent of total cost<br />

p Interest during construction (IDC)<br />

p Return of capital<br />

p Overnight construction costs (OCC)<br />

p Operations and maintenance (O&M)<br />

p Fuel<br />

Key drivers of Flamanville 3 EPR cost overruns<br />

8,000<br />

6,000<br />

4,000<br />

2,000<br />

0<br />

1,577<br />

1,923<br />

1,771<br />

Flamanville ex-post<br />

607<br />

p Budgeted cost<br />

p Design maturity<br />

p Project management<br />

(FOAK effect)<br />

p Regulatory changes<br />

p Delays<br />

1,697 2,900<br />

Taishan ex-post, pro-<strong>for</strong>ma adjusted<br />

Source: “Unlocking Reductions in the Construction Costs of <strong>Nuclear</strong>”, NEA, <strong>2020</strong><br />

Calculations based on OCC of USD 4.500 per kilowatt of electrical capacity (/kWe), a load<br />

factor of 85%, 60-year lifetime and 7-year construction time at a real discount rate of 9%.<br />

Source: NEA (<strong>2020</strong>), based on Folz (2019),<br />

“Rapport au Président Directeur Général d’EDF:<br />

La construction de l’EPR de Flamanville.”<br />

LCOE of a new nuclear power plant project according to the cost of capital<br />

160<br />

p Fuel cycle costs p Operation & maintenance costs p Overnight construction costs p Cost of capital<br />

120<br />

80<br />

40<br />

0<br />

0 % 1 % 2 % 3 % 4 % 5 % 6 % 7 % 8 % 9 % 10 % 11 % 12 %<br />

For further details<br />

please contact:<br />

Nicolas Wendler<br />

KernD<br />

Robert-Koch-Platz 4<br />

10115 Berlin<br />

Germany<br />

E-mail: presse@<br />

KernD.de<br />

www.KernD.de<br />

Did you know...?


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

386<br />

Calendar<br />

This is not a full list. Dates are subject to change.<br />

Please check the listed websites <strong>for</strong> updates.<br />

<strong>2020</strong><br />

CALENDAR<br />

07.10. – <strong>08</strong>.10.<strong>2020</strong><br />

9 th ENPPS - EU <strong>Nuclear</strong> <strong>Power</strong> Plant Simulation<br />

Forum. Brussels, Belgium, NRG Events,<br />

www.nrg-events.com<br />

11.10. – 15.10.<strong>2020</strong><br />

RRFM – European Research Reactor Conference.<br />

Helsinki, Finland, European <strong>Nuclear</strong> Society,<br />

www.euronuclear.org<br />

Virtual Meetings 15.10. and 21.10.<strong>2020</strong><br />

Implementing Digital Innovation in a <strong>Nuclear</strong><br />

Environment <strong>2020</strong>. <strong>Nuclear</strong> Institute,<br />

www.nuclearinst.com<br />

19.10. – 23.10.<strong>2020</strong><br />

<strong>International</strong> Conference on the Management<br />

of Naturally Occurring Radioactive Materials<br />

(NORM) in Industry. Vienna, Austria, IAEA,<br />

www.iaea.org<br />

Cancelled<br />

ATH'<strong>2020</strong> – <strong>International</strong> Topical Meeting<br />

on Advances in Thermal Hydraulics.<br />

Paris, France, SFEN, www.sfen-ath<strong>2020</strong>.org<br />

26.10. – 30.10.<strong>2020</strong><br />

NuMat <strong>2020</strong> – 6 th <strong>Nuclear</strong> Materials Conference.<br />

Gent, Belgium, IAEA, www.iaea.org<br />

Virtual Meeting 29.10. – 30.10.<strong>2020</strong><br />

<strong>Nuclear</strong> Decommissioning & Used Fuel Strategy.<br />

Reuters Events, www.nuclearenergyinsider.com<br />

Virtual Meeting 04.11. – 06.11.<strong>2020</strong><br />

The <strong>Power</strong> & Electricity World Africa <strong>2020</strong>.<br />

Terrapinn, www.terrapinn.com<br />

<strong>08</strong>.11. – 12.11.<strong>2020</strong><br />

Advancing Geological Repositories<br />

from Concept to Operation. Helsinki, Finland,<br />

OECD, <strong>Nuclear</strong> Energy Agency, www.oecd-nea.org<br />

<strong>09</strong>.11. – 13.11.<strong>2020</strong><br />

<strong>International</strong> Conference on Radiation Safety:<br />

Improving Radiation Protection in Practice.<br />

Vienna, Austria, IAEA, www.iaea.org<br />

Virtual Meeting 15.11. – 19.11.<strong>2020</strong><br />

ANS Winter Meeting and Technology of Fusion<br />

Energy (TOFE <strong>2020</strong>). American <strong>Nuclear</strong> Society,<br />

www.ans.org<br />

30.11. – 02.12.<strong>2020</strong><br />

European <strong>Power</strong> Strategy & Systems Summit.<br />

Prague, Czech Republic, European <strong>Power</strong><br />

Generation, www.europeanpowergeneration.eu<br />

07.12. – 10.12.<strong>2020</strong><br />

SAMMI <strong>2020</strong> – Specialist Workshop on Advanced<br />

Measurement Method and Instrumentation<br />

<strong>for</strong> enhancing Severe Accident Management in<br />

an NPP addressing Emergency, Stabilization and<br />

Long-term Recovery Phases. Fukushima, Japan,<br />

NEA, www.sammi-<strong>2020</strong>.org<br />

<strong>08</strong>.12. – 10.12.<strong>2020</strong><br />

World <strong>Nuclear</strong> Exhibition <strong>2020</strong>. Paris Nord<br />

Villepinte, France, Gifen,<br />

www.world-nuclear-exhibition.com<br />

Virtual Meeting 17.12. – 18.12.<strong>2020</strong><br />

ICNESPP <strong>2020</strong> – 14. <strong>International</strong> Conference on<br />

<strong>Nuclear</strong> Engineering Systems and <strong>Power</strong> Plants.<br />

WASET, www.waset.org<br />

2021<br />

07.03. – 11.03.2021<br />

WM2021 – Waste Management Symposia.<br />

Phoenix, Arizona, USA, X-CD Technologies,<br />

www.wmsym.org<br />

17.03. – 19.03.2021<br />

KONTEC 2021 – 15 th <strong>International</strong> Symposium<br />

“Conditioning of Radioactive Operational &<br />

Decommissioning Wastes”. Dresden, Germany,<br />

atm, www.kontec-symposium.de<br />

23.03. – 26.03.2021<br />

7 th international conference on Education and<br />

Training in Radiation Protection. Groningen,<br />

Netherlands, FuseNet, www.etrap.net<br />

26.04. – 27.04.2021<br />

AtomExpo 2021. Sochi, Russia, Rosatom,<br />

www.2021.atomexpo.ru<br />

26.04. – 30.04.2021<br />

European <strong>Nuclear</strong> Young Generation Forum<br />

(ENYGF). Tarragona, Spain, ENYGF, www.enygf.org<br />

Postponed to 02.06. – 04.06.2021<br />

HTR<strong>2020</strong> – 10 th <strong>International</strong> Conference<br />

on High Temperature Reactor Technology.<br />

Yogyakarta, Indonesia, Indonesian <strong>Nuclear</strong> Society,<br />

www.htr<strong>2020</strong>.org<br />

Postponed to 30.<strong>08</strong>. – 03.<strong>09</strong>.2021<br />

<strong>International</strong> Conference on Operational Safety<br />

of <strong>Nuclear</strong> <strong>Power</strong> Plants. Beijing, China, IAEA,<br />

www.iaea.org<br />

Postponed to <strong>08</strong>.<strong>09</strong>. – 10.<strong>09</strong>.2021<br />

3 rd <strong>International</strong> Conference on Concrete<br />

Sustainability. Prague, Czech Republic, fib,<br />

www.fibiccs.org<br />

27.<strong>09</strong>. – 01.10.2021<br />

NPC 2021 <strong>International</strong> Conference on <strong>Nuclear</strong><br />

Plant Chemistry. Antibes, France, SFEN Société<br />

Française d’Energie Nucléaire,<br />

www.sfen-npc2021.org<br />

Postponed to 07.<strong>09</strong>. – 10.<strong>09</strong>.2021<br />

<strong>International</strong> Forum on Enhancing a Sustainable<br />

<strong>Nuclear</strong> Supply Chain. Helsinki, Finland, Foratom,<br />

https://events.<strong>for</strong>atom.org/mstf<strong>2020</strong>/<br />

Postponed to 30.11. – 02.12.2021<br />

Enlit (<strong>for</strong>mer European Utility Week and<br />

POWERGEN Europe). Milano, Italy,<br />

www.powergeneurope.com<br />

Postponed to 2021<br />

The Frédéric Joliot/Otto Hahn Summer School<br />

on <strong>Nuclear</strong> Reactors “Physics, Fuels and Systems”.<br />

Aix-en-Provence, France, CEA & KIT, www.fjohss.eu<br />

Postponed to 2021<br />

INDEX <strong>2020</strong>: <strong>International</strong> <strong>Nuclear</strong> Digital<br />

Experience. Paris, France, SFEN,<br />

www.sfen-index<strong>2020</strong>.org<br />

Postponed to 2021<br />

4 th CORDEL Regional Workshop – Harmonization<br />

to support the operation and new build of NPPs<br />

including SMR. Lyon, France, World <strong>Nuclear</strong><br />

Association, www.events.<strong>for</strong>atom.org<br />

2022<br />

18.11. – 19.11.<strong>2020</strong><br />

INSC — <strong>International</strong> <strong>Nuclear</strong> Supply Chain<br />

Symposium. Munich, Germany, TÜV SÜD,<br />

www.tuvsud.com<br />

23.11. – 25.11.<strong>2020</strong><br />

KELI <strong>2020</strong> – Conference <strong>for</strong> Electrical Engineering,<br />

I&C and IT in generation plants. Bremen, Germany,<br />

VGB <strong>Power</strong>Tech e.V., www.vgb.org<br />

Postponed to 10.05. – 15.05.2021<br />

FEC <strong>2020</strong> – 28 th IAEA Fusion Energy Conference.<br />

Nice, France, IAEA, www.iaea.org<br />

Postponed to 29.05. – 05.06.2021<br />

BEPU<strong>2020</strong> – Best Estimate Plus Uncertainty <strong>International</strong><br />

Conference, Giardini Naxos. Sicily, Italy,<br />

NINE, www.nineeng.com<br />

24.11. – 26.11.<strong>2020</strong><br />

ICOND <strong>2020</strong> – 9 th <strong>International</strong> Conference on<br />

<strong>Nuclear</strong> Decommissioning. Aachen, Germany,<br />

AiNT, www.icond.de<br />

Postponed to 31.05. – 04.06.2021<br />

20 th WCNDT – World Conference on<br />

Non-Destructive Testing. Incheon, Korea,<br />

The Korean Society of Nondestructive Testing,<br />

www.wcndt<strong>2020</strong>.com<br />

KERNTECHNIK 2022.<br />

Germany, KernD and KTG,<br />

www.kerntechnik.com<br />

Calendar


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Current Status and Prospects of<br />

<strong>Nuclear</strong> <strong>Power</strong> Plant Decommissioning<br />

in the Republic of Korea<br />

Joo Hyun Moon<br />

Introduction As of August <strong>2020</strong>, there are 30 nuclear power reactors in the Republic of Korea as shown in<br />

Table 1 and Figure 1. Among those reactors, 24 nuclear power reactors are in operation; four reactors (Shin Hanul<br />

units 1 and 2 and Shin Kori units 5 and 6) are under construction, and two reactors (Kori unit 1 and Wolsong unit 1) are<br />

permanently shut down.<br />

South Korea is now facing the problem of safe decommissioning<br />

of Kori unit 1 because it is the first commercial<br />

power reactor to be decommissioned. Wolsong unit 1 was<br />

declared to be permanently shut down in June 2018, and<br />

an application <strong>for</strong> a permit of change in its operating<br />

license, which is an initial step to decommission a nuclear<br />

facility required by nuclear safety regulations in Korea,<br />

was approved by the nuclear safety and security commission<br />

(NSSC) that is the national nuclear regulatory<br />

­authority in December 2019.<br />

No. Name Reactor<br />

Type<br />

Net Capacity<br />

(MWe)<br />

Although it has experience on completing the decommissioning<br />

of two research reactors (TRIGA MARK II and<br />

III), Korea does not have any experience on the whole<br />

decommissioning process of a commercial nuclear power<br />

reactor. The Korean government has been preparing <strong>for</strong><br />

the safe decommissioning of Kori unit 1 since several years<br />

prior to the expiration date of its renewed operating<br />

license. The government has considered decommissioning<br />

Kori unit 1 to be an opportunity to develop new technologies<br />

to decommission nuclear facilities and extend its<br />

Issue Date of Operating License<br />

(First Critical Date)<br />

1 Kori 1 PWR 587 1972.05.31<br />

(1977.06.19)<br />

Expiration Date<br />

of Operating License<br />

Design Life<br />

(Year)<br />

Status<br />

2017.06.18 30 Permanent<br />

shutdown<br />

2 Kori 2 PWR 650 1983.<strong>08</strong>.10 2023.<strong>08</strong>.<strong>09</strong> 40 Operational<br />

3 Kori 3 PWR 950 1984.<strong>09</strong>.29 2024.<strong>09</strong>.28 40 Operational<br />

4 Kori 4 PWR 950 1985.<strong>08</strong>.07 2025.<strong>08</strong>.06 40 Operational<br />

5 Wolsong 1 PHWR 679 1978.02.15<br />

(1982.11.21)<br />

2022.11.20 30 Permanent<br />

shutdown<br />

6 Wolsong 2 PHWR 700 1996.11.02 2026.11.01 30 Operational<br />

7 Wolsong 3 PHWR 700 1997.12.30 2027.12.29 30 Operational<br />

8 Wolsong 4 PHWR 700 1999.02.<strong>08</strong> 2029.02.07 30 Operational<br />

9 Hanbit 1 PWR 950 1985.12.23 2025.12.22 40 Operational<br />

10 Hanbit 2 PWR 950 1986.<strong>09</strong>.12 2026.<strong>09</strong>.11 40 Operational<br />

11 Hanbit 3 PWR 1,000 1994.<strong>09</strong>.<strong>09</strong> 2034.<strong>09</strong>.<strong>08</strong> 40 Operational<br />

12 Hanbit 4 PWR 1,000 1995.06.02 2035.06.01 40 Operational<br />

13 Hanbit 5 PWR 1,000 2001.10.24 2041.10.23 40 Operational<br />

14 Hanbit 6 PWR 1,000 2002.07.31 2042.07.30 40 Operational<br />

15 Hanul 1 PWR 1,000 1987.12.23 2027.12.22 40 Operational<br />

16 Hanul 2 PWR 1,000 1988.12.29 2028.12.28 40 Operational<br />

17 Hanul 3 PWR 1,000 1997.11.<strong>08</strong> 2037.11.07 40 Operational<br />

18 Hanul 4 PWR 1,000 1998.10.29 2038.10.28 40 Operational<br />

19 Hanul 5 PWR 1,000 2003.10.20 2043.10.19 40 Operational<br />

20 Hanul 6 PWR 1,000 2004.11.12 2044.11.11 40 Operational<br />

21 Shin Kori 1 PWR 1,000 2010.05.19 2050.05.18 40 Operational<br />

22 Shin Kori 2 PWR 1,000 2011.12.02 2051.12.01 40 Operational<br />

23 Shin Kori 3 PWR 1,400 2015.10.30 2075.10.29 60 Operational<br />

24 Shin Kori 4 PWR 1,400 2019.02.01 2079.01.31 60 Operational<br />

25 Shin Wolsong 1 PWR 1,000 2011.12.02 2051.12.01 40 Operational<br />

26 Shin Wolsong 2 PWR 1,000 2014.11.14 2054.11.13 40 Operational<br />

27 Shin Hanul 1 PWR 1,400 - - 60 Under<br />

construction<br />

28 Shin Hanul 2 PWR 1,400 - - 60 Under<br />

construction<br />

29 Shin Kori 5 PWR 1,400 - - 60 Under<br />

construction<br />

30 Shin Kori 6 PWR 1,400 - - 60 Under<br />

construction<br />

387<br />

FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT<br />

| Tab. 1.<br />

<strong>Nuclear</strong> Reactors in the Republic of Korea. (As of August <strong>2020</strong>)<br />

Feature<br />

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FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 388<br />

| Fig. 1.<br />

<strong>Nuclear</strong> power plant sites in the Republic of Korea.<br />

own nuclear industrial capabilities to cover all areas of the<br />

nuclear fuel cycle.<br />

In April 2019, the Korean government established ‘a<br />

business strategy to promote domestic industry in the field<br />

of nuclear facility decommissioning’ [1] to technically<br />

support the decommissioning of Kori unit 1 and to develop<br />

the domestic nuclear industry’s technical competence in<br />

regards to the decommissioning of nuclear facilities. With<br />

this strategy, the Korean government set an ambitious<br />

target to foster the domestic nuclear industry’s technical<br />

competence; the goal is to be a contender in the global<br />

market <strong>for</strong> nuclear facility decommissioning by 2035.<br />

The Korea Hydro and <strong>Nuclear</strong> <strong>Power</strong> Corporation<br />

( KHNP) operates all nuclear power plants in the Republic<br />

of Korea and is responsible <strong>for</strong> the safe and successful<br />

decommissioning of Kori unit 1. KHNP came up with and is<br />

now collecting opinions from the local residents on the<br />

decommissioning plan (DP) <strong>for</strong> Kori unit 1 as required by<br />

our national nuclear safety regulations. Main decommissioning<br />

activities will start immediately after the DP is<br />

approved by the <strong>Nuclear</strong> Safety and Security Commission<br />

(NSSC). For successful decommissioning of the first<br />

commercial nuclear power reactor in Korea in cooperation<br />

with domestic and <strong>for</strong>eign corporations, the KHNP has<br />

prepared extensively from the development of technologies<br />

needed <strong>for</strong> decommissioning and securing funds <strong>for</strong><br />

decommissioning to organizing a department responsible<br />

<strong>for</strong> leading the decommissioning project. These developments<br />

have been in motion since several years be<strong>for</strong>e the<br />

expiration date of Kori unit 1’s renewed operating license.<br />

Few papers have been written on the preparation status<br />

of the decommissioning of Kori unit 1 and prospects<br />

regarding the decommissioning of nuclear power plants in<br />

the Republic of Korea. Hence, this paper will examine the<br />

preparation status of the decommissioning of Kori unit 1<br />

by the KHNP and the Korean government, and the<br />

prospects of decommissioning of other nuclear power<br />

reactors in the Republic of Korea.<br />

Status of Kori unit 1<br />

Kori unit 1 is the first commercial nuclear power reactor in<br />

Korea and operated <strong>for</strong> about 40 years starting June 1977<br />

when it reached first criticality. Kori unit 1 had a capacity<br />

of 587 MW and produced a total of 155,260 GWh of electricity<br />

in that period. The original operating lifetime <strong>for</strong><br />

Kori unit 1 was 30 years, but was extended by 10 years<br />

through license renewal in 2007. In June 2015, the board<br />

of directors of KHNP decided to permanently shut<br />

down Kori unit 1 with a renewed operating license<br />

­termination date of 18 June 2017.<br />

For its permanent shutdown, pursuant with nuclear<br />

safety regulations, the KHNP had to submit a written<br />

application <strong>for</strong> a permit of change in operating license on<br />

24 June 2016 to the NSSC, and won the approval from<br />

­NSSC on 9 June 2017. The KHNP then shut down Kori unit<br />

1 permanently on 18 June 2017. Then, spent nuclear fuel<br />

was discharged from the reactor and transported to the<br />

temporary storage water pool. Since then, Kori unit 1 had<br />

been kept in the cold shutdown state.<br />

Pursuant to an immediate dismantling strategy, KHNP<br />

decided to complete the decommissioning of Kori unit 1<br />

with a budget of 812.9 billion Korean Won (KWN) as the<br />

2019-year present value (about 677 million US dollars at<br />

exchange rate of $1 = 1,200 KWN) within about 15 years.<br />

The strategy consists of four main stages: 1) management<br />

of permanent shutdown conditions and preparation <strong>for</strong><br />

decommissioning; 2) beginning of decommissioning and<br />

construction of facilities <strong>for</strong> radioactive waste treatment;<br />

3) decontamination/dismantling and waste treatment;<br />

and 4) site restoration and report of completion of<br />

decommissioning, as shown in Figure 2 [2]. The main<br />

tasks in each stage are as follows:<br />

Stage 1<br />

p Management of spent nuclear fuel and inspections<br />

­conducted by a regulatory body;<br />

| Fig. 2.<br />

Four main stages of decommissioning of Kori unit 1 [2].<br />

Feature<br />

Current Status and Prospects of <strong>Nuclear</strong> <strong>Power</strong> Plant Decommissioning in the Republic of Korea ı Joo Hyun Moon


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 3.<br />

Development status of technologies and devices <strong>for</strong> decommissioning of Kori unit 1 [2].<br />

p Development of DP and the design and procurement of<br />

waste treatment facilities;<br />

p Preparation of decommissioning planning documents<br />

and application <strong>for</strong> DP approval from the NSSC.<br />

Stage 2<br />

p Dismantling of non-radioactive areas, installation and<br />

operation of utilities <strong>for</strong> decommissioning;<br />

p Construction of radioactive waste treatment facilities;<br />

p Transport of spent nuclear fuel to off-site storage<br />

facility.<br />

Stage 3<br />

p Decontamination and dismantling of radioactive<br />

­systems and structures;<br />

p Operation of radioactive waste treatment facilities<br />

( decontamination, cutting, volume reduction, packaging<br />

etc.);<br />

p Evaluation and verification of radioactivity measurements.<br />

Stage 4<br />

p Site restoration;<br />

p Final status survey and inspection <strong>for</strong> closure of decommissioning<br />

processes;<br />

p Termination of operating license <strong>for</strong> Kori unit 1.<br />

To secure technical competence in the decommissioning of<br />

Kori unit 1, the KHNP <strong>for</strong>mulated and implemented the<br />

roadmap <strong>for</strong> the development of 17 decommissioning<br />

technologies in 2017, as shown in Figure 3. The KHNP<br />

developed the technology tree to decommission nuclear<br />

power plants and identified 58 technologies to complete<br />

decommissioning. The gap analysis identified 17 technologies<br />

that were lacking or insufficient <strong>for</strong> use onsite and<br />

must be developed with urgency. As of December 2017,<br />

13 of 17 technologies were under development and the<br />

development of the other four technologies will be<br />

completed by 2021. In parallel with technologies development,<br />

the key decommissioning devices with high value<br />

added are being developed.<br />

Although the DP <strong>for</strong> Kori unit 1 is subject to change<br />

because it has not been approved by the NSSC, the KHNP is<br />

going to carry <strong>for</strong>ward with the decommissioning project<br />

using the project management system shown in Figure 4.<br />

In this system, KHNP is in overall charge of this project and<br />

will be a licensee <strong>for</strong> the decommissioning of Kori unit 1.<br />

The maintenance of Kori unit 1 till initiation of the main<br />

decommissioning activities will be undertaken by the<br />

KHNP in cooperation with the existing contractors.<br />

For efficient project management, the KHNP would<br />

categorize the decommissioning projects into the six<br />

elementary businesses: comprehensive design of decommissioning;<br />

system decontamination, construction and<br />

operation of waste treatment facilities and decontamination/dismantling;<br />

cutting and dismantling of the nuclear<br />

reactor system; radiation measurement, evaluation, and<br />

verification; and site restoration. For each elementary<br />

business, the KHNP will choose and enter into a business<br />

contract with a well-equipped company or consortium<br />

through competitive bids.<br />

There are three matters that might have significant<br />

­influence on the schedule of the Kori unit 1 decommissioning<br />

project. The first is transport of spent nuclear fuel out<br />

of the temporal storage pool. Pursuant to nuclear safety<br />

regulations, the main decommissioning activities shall not<br />

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FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 390<br />

| Fig. 4.<br />

Project management system <strong>for</strong> decommissioning of Kori unit 1 [2].<br />

| Fig. 5.<br />

Number of nuclear reactors to be decommissioned by time period.<br />

be started until all spent nuclear fuel stored at the temporal<br />

storage pool of Kori unit 1 has been transported to other<br />

storage pools or sites. At present, it seems to be hard to<br />

secure enough space at the other sites to accommodate all<br />

nuclear fuel accumulated at the temporal storage pool of<br />

Kori unit 1. It is also uncertain when the interim storage<br />

site <strong>for</strong> spent nuclear fuel could be secured due to an<br />

aversion to spent nuclear fuel in Korea. Due to this fact, the<br />

decommissioning of Kori unit 1 may be delayed by several<br />

years.<br />

The second is disposal of radioactive waste. KHNP has<br />

set 14,500 200L drums as the target <strong>for</strong> final disposal<br />

amount of all low- and intermediate-level radioactive<br />

waste generated over the whole decommissioning period.<br />

Be<strong>for</strong>e applying any volume reduction techniques, about<br />

several hundred thousand drums of decommissioning<br />

waste including intermediate-, low-, and very low-level<br />

waste would be generated during the whole process<br />

of decommissioning, which is huge amount even <strong>for</strong> a<br />

relatively short period of about 15 years. It is big challenge<br />

to classify, condition, and treat huge amounts of waste<br />

­adequately and finally reduce the waste volume to about a<br />

tenth of raw waste volume. Even if successful in reducing<br />

the volume, 14,500 drums would be also big burden to the<br />

operator of the disposal facility because all drums need to<br />

be inspected to comply with waste acceptance criteria.<br />

Because unit disposal costs per 200L drum of low- and<br />

intermediate-level waste is about 15.2 million KWN (about<br />

12,667 US $ at exchange rate of $1 = 1,200 KWN) in<br />

Korea, the more drums to be disposed of, the higher the<br />

disposal cost and the higher the total cost of decommissioning.<br />

To reduce disposal burden and cut down total<br />

costs of decommissioning Kori unit 1 as much as possible,<br />

it is essential to minimize the quantity of waste drums to be<br />

finally disposed.<br />

The third matter is the related impacts on the operation<br />

of neighboring nuclear reactors. At the Kori site, there are<br />

a total of four nuclear reactors including Kori unit 1.<br />

Because Kori unit 2 is immediately adjacent to Kori unit 1,<br />

the decommissioning works of Kori unit 1 might influence<br />

the operation of Kori unit 2. Hence, the decommissioning<br />

works should carry <strong>for</strong>ward only within the limits so as to<br />

not cause inconvenience in the operation of Kori unit 2; the<br />

whole schedule of Kori unit 1 could be affected by the<br />

operation schedule of Kori unit 2.<br />

Prospect of nuclear decommissioning in Korea<br />

In December 2017, the Ministry of Trade, Industry and<br />

­Energy (MOTIE) released ‘8 th Basic Plan <strong>for</strong> Long-term<br />

Electricity Supply and Demand (2017–2031) (8 th plan)’<br />

[3]. According to the 8 th plan, license renewal beyond the<br />

designated lifetimes of any existing nuclear reactor is not<br />

allowed. Without considering lifetime extension, the<br />

number of nuclear power reactors to be permanently shut<br />

down with each coming decade is shown in Figure 5.<br />

The annual expenses expected to be spent on<br />

the decommissioning of the 26 reactors in Korea are<br />

roughly estimated without considering inflation. Among<br />

30 reactors in Table 1, the decommissioning start dates of<br />

four reactors which are being under construction were<br />

not able to be fixed and are there<strong>for</strong>e not included in this<br />

estimation. For this estimation, the main assumptions are<br />

as follows:<br />

Considering the 8 th plan, the license renewal <strong>for</strong> the<br />

designated lifetime extension <strong>for</strong> all nuclear power plants<br />

is not considered. The decommissioning of a nuclear<br />

Feature<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 6.<br />

Annual expenses expected to be spent on the decommissioning of nuclear power plants in Republic of Korea.<br />

power reactor will be completed within 15 years. Stage 1<br />

<strong>for</strong> decommissioning would be initiated immediately after<br />

the permanent shutdown date.<br />

The total expense <strong>for</strong> decommissioning of a nuclear<br />

­reactor will be 812.9 billion KRW which is equivalent to<br />

about $677M US at the 2019-year present value, which is<br />

subject to review and revision every two years by MOTIE.<br />

The decommissioning cost consists of labor costs,<br />

dismantling costs, decontamination costs, waste disposal<br />

costs (including waste transportation costs), and miscellaneous<br />

costs such as insurance fees, taxes, and utility costs.<br />

The annual expense varies according to activities that<br />

would be per<strong>for</strong>med by year.<br />

Based on the above assumptions, the annual decommissioning<br />

cost of the 26 nuclear power reactors were shown<br />

in Figure 6. Accordingly, the annual decommissioning cost<br />

and the decommissioning market size will grow gradually<br />

and reach the maximum size in 2037 when the decommissioning<br />

of 13 nuclear reactors will be carried out<br />

simultaneously. From that point, the decommissioning<br />

market size will gradually decrease until 2069 when the<br />

decommissioning of Shin Wolsong unit 2 is completed.<br />

There would be no decommissioning works over the next<br />

five years from 2070 to 2074, and in 2075, the decommissioning<br />

of Shin Kori unit 3 would be initiated. Those<br />

estimated timings might be altered due to changes in the<br />

future circumstances.<br />

Conclusion<br />

This paper reviews the current status of the decommissioning<br />

process of Kori unit 1. Because Kori unit 1 is the<br />

first commercial nuclear reactor to be decommissioned in<br />

the Republic of Korea, the Korean government and the<br />

KHNP have prepared to ensure the safety of the project.<br />

In particular, the Korean government considered this to<br />

be an opportunity to strengthen its nuclear industry’s<br />

competence and foster domestic specialized companies in<br />

the field of decommissioning of nuclear facilities. Thus,<br />

the Korean government <strong>for</strong>mulated ‘a business strategy to<br />

promote the domestic industry in field of nuclear facility<br />

decommissioning.’ The KHNP has made preparations to<br />

decommission Kori unit 1 through measure such as the<br />

<strong>for</strong>mulation and implementation of roadmaps <strong>for</strong> technological<br />

developments, <strong>for</strong>mulation of a decommissioning<br />

plan and making license-related documents, and etc. This<br />

paper estimates the number of nuclear power reactors to<br />

be permanently shut down with each coming decade and<br />

the expected annual expenses of decommissioning 26<br />

­nuclear power reactors in Korea from 2017 to 2<strong>09</strong>4.<br />

­Although the estimations presented are rough, they show<br />

that the size of the decommissioning market will grow<br />

gradually and reach the maximum in 2037 when the<br />

decommissioning of 13 nuclear reactors would be carried<br />

out simultaneously. This paper is expected to be helpful to<br />

get insight on the status and prospects of decommissioning<br />

of nuclear facilities in the Republic of Korea.<br />

Acknowledgements<br />

This present research was conducted by the research fund<br />

of Dankook University in 2019.<br />

References<br />

[1] Korean Government, “a business strategy to promote the domestic industry in field of nuclear facility<br />

decommissioning (in Korean),” April 2019.<br />

[2] Young Gi Choi, “Plan <strong>for</strong> Decommissioning of Kori Unit 1 (in Korean),” <strong>Nuclear</strong> Industry, 2017(12),<br />

pp.40-49, December 2017.<br />

[3] Ministry of Trade, Industry and Energy, “8th Basic Plan <strong>for</strong> Long-term Electricity Supply and Demand,”<br />

December 2017.<br />

Author<br />

Joo Hyun Moon<br />

jhmoon86@dankook.ac.kr<br />

Department of <strong>Nuclear</strong> Engineering<br />

Dankook University<br />

Cheonan-Si,<br />

Chungnam 31116, Rep. of Korea<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

392<br />

Kostenkorrekturen für Standortauswahlverfahren –<br />

Es bleibt unübersichtlich<br />

Tobias Leidinger<br />

SPOTLIGHT ON NUCLEAR LAW<br />

Gesetzliche Kostenregelungen sind häufig kompliziert, unübersichtlich und zuweilen auch nicht (mehr)<br />

sachgerecht. Mit dem Ende Juli <strong>2020</strong> vom Bundeskabinett beschlossenen „Gesetz zur Anpassung der Kostenvor­schriften<br />

im Bereich der Entsorgung radioaktiver Abfälle sowie zur Änderung weiterer Vorschriften“ zielt der Gesetzgeber u. a.<br />

auf eine nachträgliche Korrekturmöglichkeit bei der Kostenverteilung für das Standortauswahlverfahren nach<br />

­Standortauswahlgesetz (StandAG) ab. Dieser grundsätzlich begrüßenswerte Schritt fällt bei näherer Betrachtung<br />

allerdings nicht konsequent aus.<br />

Kostentragung für die Endlagersuche<br />

Die Kosten für das komplexe Standortauswahlverfahren<br />

in Deutschland für das Endlager für hochradioaktive<br />

­Abfälle werden nach § 29 Abs. 1 StandAG auf die Umlagepflichtigen<br />

anteilig umgelegt. Das bedeutet konkret,<br />

dass das Bundesumweltministerium (BMU) jedes Jahr<br />

Bescheide zur Festsetzung der im vorangegangenen Jahr<br />

umlagefähigen Kosten einerseits und zur Vorauszahlung<br />

dieser Kosten für das laufende Jahr andererseits gegenüber<br />

den Beitragsverpflichteten erhebt.<br />

Das Problem: Überholter Kostenschlüssel<br />

Problematisch ist diese Kostenregelung deshalb, weil der<br />

Maßstab, nach dem sich der jeweils zu entrichtende Anteil<br />

eines Umlagepflichtigen an den umlagefähigen Kosten bemisst,<br />

bereits seit längerer Zeit nicht mehr „passt“. § 29<br />

Abs. 2 StandAG legt insoweit fest, dass dafür der Aufwandschlüssel<br />

aus § 6 der Endlagervorausleistungsverordnung<br />

(EndlagerVlV) entsprechend heranzuziehen ist. Dieser<br />

„Schlüssel“ ist für die danach Zahlungspflichtigen der<br />

Gruppe der „Nicht-KKW-Betreiber“ (§ 6 Abs. 1 Nr. 2 lit. c)<br />

EndlagerVlV) seit geraumer Zeit nicht mehr aufwandsgerecht.<br />

Denn der diesem Schlüssel zugrundeliegende<br />

Sachverhalt hat sich geändert: Das Endlager Konrad<br />

für schwach- und mittelradioaktive Abfälle wurde 2007<br />

rechtskräftig planfestgestellt, der „Schlüssel“ in § 6<br />

­EndlagerVIV – und dementsprechend auch die Umlagebescheide<br />

des BMU nach StandAG – gehen indes noch<br />

­immer von einer Addition der Abfälle für das Endlager<br />

Konrad und das Endlager für hochradioaktive Abfälle<br />

(HAW) aus („Ein-Endlager-Prinzip“). Seit der Bestandskraft<br />

des Planfeststellungsbeschlusses für das Endlager<br />

Konrad am 26. März 2007, dem Tag der Zurückweisung<br />

der Nicht-Zulassungsbeschwerde gegen das die Klagen<br />

­zurückweisende Urteil des OVG Lüneburg durch das<br />

BVerwG, steht indes fest, dass die Konradmengen nicht im<br />

HAW-Endlager endgelagert werden.<br />

Das StandAG unterstellt indes ausweislich seiner<br />

Gesetzes­begründung – vgl. BT-Drs. 17/13471, S. 19 – dass<br />

nur die nicht in das Endlager Konrad einzulagernden<br />

­Abfälle in das nach StandAG zu suchende Endlager eingebracht<br />

werden sollen. Es verweist in § 29 Abs. 2 StandAG<br />

aber gleichwohl auf den bisherigen Kostenschlüssel in<br />

§ 6 der EndlagerVIV, der diese Trennung gerade nicht<br />

­berücksichtigt. Eine verursachungsgerechte Verteilung<br />

der Kosten hätte mithin seit März 2007 (also vor mehr als<br />

10 Jahren!) erfolgen müssen. Denn eine Addition<br />

der ­Mengen für Konrad und das HAW-Endlager ist<br />

seitdem nicht mehr zulässig. Die Regelung nach § 6 Abs. 3<br />

Satz 3 EndlagerVIV ist mithin nicht aufwandsgerecht.<br />

Diese sachlich überholte Rechtslage führt dazu, dass<br />

die ­Umlagebescheide nach StandAG, die auf den<br />

Kostenschlüssel in § 6 EndlagerVlV aufsetzen, zumeist mit<br />

Rechtsmitteln angegriffen werden.<br />

Die Lösung: Nachträgliche Korrektur<br />

ermöglichen – Richtiger Maßstab fehlt aber<br />

nach wie vor<br />

Das geplante „Kostenanpassungsgesetz“ sieht nunmehr<br />

mit der Neu-Einführung von § 35a StandAG eine Regelung<br />

vor, die die abschließende Berechnung der Umlagebeträge<br />

am Ende der Standortsuche für das Endlager (also nach<br />

2031) vorsieht. In diese Neu-Berechnung können die nach<br />

dem 1. Januar 2021 festgesetzten oder in der Vergangenheit<br />

erhobene Beiträge einbezogen werden, bei denen<br />

die Festsetzung entweder noch nicht bestandskräftig ist<br />

( wegen Einlegung von Rechtsbehelfen) oder bei denen die<br />

Festsetzung rechtswidrig erfolgt ist, also die Voraussetzungen<br />

für eine Rücknahme der Bescheide nach § 48<br />

des Verwaltungsverfahrensgesetzes (VwVfG) vorliegen.<br />

Diese Neuregelung ist im Grundsatz zu begrüßen, weil<br />

damit eine nachträgliche „Umverteilung“ des Kostenaufwands,<br />

also eine Korrektur unrichtiger Beitragserhebungen,<br />

ermöglicht wird.<br />

Die Neuregelung in § 35a StandAG ist aber nicht<br />

­ausreichend: Es fehlt jeglicher Anhaltspunkt sowohl im<br />

Gesetzestext als auch in der Gesetzesbegründung, nach<br />

welchem inhaltlichen Maßstab die spätere Neu berechnung<br />

der Umlagebeträge, d. h. nach der finalen Feststellung des<br />

Endlagerstandorts erfolgen soll. Da eine Veränderung bei<br />

einem Zahlungspflichtigen dazu führt, dass sich die<br />

­Umlagebeträge auch für alle anderen Zahlungspflichtigen<br />

ändern, bedarf es für eine Neuberechnung eines klar<br />

­definierten inhaltlichen Maßstabs. Daran fehlt es indes.<br />

Vielmehr bleibt auch weiterhin der längst überholte<br />

­Kostenschlüssel in § 6 EndlagerVlV maßgebend, wie sich<br />

aus dem Verweis der Neuregelung in § 35a StandAG auf<br />

die unverändert in Bezug genommenen Regelungen in<br />

§§ 29, 31, 32 und 35 StandAG ergibt. Das verwundert<br />

schon. Zumindest in der Gesetzesbegründung wäre ein<br />

Hinweis zu erwarten gewesen, dass der bisherige Schlüssel<br />

aus § 6 EndlagerVlV absehbar novelliert wird, so dass er<br />

auf den Sachverhalt nach StandAG „passt“. Trotz der Neuregelung<br />

in § 35a StandAG ist also festzustellen, dass es<br />

auch danach an einem realitätsgerechten Maßstab für die<br />

Bestimmung der Umlagen nach StandAG fehlt und zwar<br />

nicht nur für die Zeit bis zur finalen Feststellung des<br />

Endlagerstandorts, sondern auch für den Zeitpunkt der<br />

abschließenden Berechnung nach § 35a StandAG selbst.<br />

Keine Zinsregelung vorgesehen<br />

Unter beitragsrechtlichen Gesichtspunkten nicht unproblematisch<br />

ist die Tatsache, dass – trotz des sehr langen<br />

Zeitraums, der noch bis zur endgültigen Berechnung der<br />

Spotlight on <strong>Nuclear</strong> Law<br />

Cost Correction <strong>for</strong> Site Selection Procedures – It Remains Confusing ı Tobias Leidinger


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

finalen Umlagebeiträge nach § 35a StandAG verstreichen<br />

wird – eine Zinsregelung für überzahlte, d. h. am Ende<br />

zurück zu erstattende Beiträge nicht vorgesehen ist. Die<br />

Gesetzesbegründung zu § 35a StandAG verweist dazu<br />

­lediglich auf § 34 StandAG, also eine unverzinste Rückerstattung.<br />

Damit weicht § 35a StandAG für das HAW-­<br />

Endlager von der Regelung für das Endlager Konrad ab:<br />

§ 8 EndlagerVlV sieht hier eine Verzinsung von 3 % über<br />

dem Basiszinssatz gesetzlich vor, wenn überzahlte Beträge<br />

an die Beitragspflichtigen zurück zu erstatten sind. Warum<br />

dies nicht auch im Rahmen der Erstattung von Zahlungen<br />

nach § 35a StandAG gelten soll, erschließt sich nicht. Eine<br />

Begründung dafür fehlt.<br />

Fazit<br />

Die nach § 35a StandAG eröffnete Korrektur der Kostenverteilung<br />

nach Abschluss des Standortauswahlverfahrens<br />

ist grundsätzlich zu begrüßen. Damit ist es möglich,<br />

­sachliche Veränderungen in die abschließende Umlageberechnung<br />

einfließen zu lassen. Das neue Gesetz lässt<br />

aber den für die finale Berechnung er<strong>for</strong>derlichen inhaltlichen<br />

Maßstab vermissen. Darüber hinaus bleibt der für<br />

die weiteren Festsetzungs- und Vorauszahlungsbescheide<br />

nach StandAG sachlich längst überholte „Schlüssel“ nach<br />

§ 6 EndlagerVlV auch weiterhin unverändert. Damit sind<br />

entsprechende Umlagebescheide auch zukünftig angreifbar.<br />

Die Anzahl der Rechtsbehelfe wird also kaum<br />

­abnehmen, die Situation für die Beitragspflichtigen bleibt<br />

damit bis auf Weiteres unübersichtlich. Von einer „echten<br />

Lösung“ kann also aktuell keine Rede sein, sie wurde<br />

vielmehr in die weite Zukunft verschoben.<br />

Imprint<br />

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SPOTLIGHT ON NUCLEAR LAW 393<br />

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ISSN 1431-5254<br />

Spotlight on <strong>Nuclear</strong> Law<br />

Cost Correction <strong>for</strong> Site Selection Procedures – It Remains Confusing ı Tobias Leidinger


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

394<br />

DECOMMISSIONING AND WASTE MANAGEMENT<br />

Actual Research and Development<br />

Activities in the Field of Dismantling<br />

Sascha Gentes and Nadine Gabor<br />

The shutdown of nuclear facilities is attracting more and more public attention, not only because of their limited<br />

life cycle, but primarily due to the political decision to phase out nuclear power. For the engineers involved, the<br />

complete deconstruction and decommissioning of such facilities represents an extremely complex problem with<br />

countless constraints and variables which have constantly to be taken into account and incorporated into the process.<br />

Deconstruction work often relies on standard construction equipment, but this has to be modified and refined <strong>for</strong> each<br />

application and <strong>for</strong> each part of the structure.<br />

| Fig. 1.<br />

Patented novel milling cutter to remove highly rein<strong>for</strong>ced concrete.<br />

The deconstruction costs run into<br />

several hundred million euros, depending<br />

on the actual design of the<br />

power station, and the demolition<br />

work itself takes about ten years. The<br />

phasing out of nuclear energy and the<br />

attendant switching off of Germany’s<br />

nuclear power stations by 2022 have<br />

focused public attention even more on<br />

this particular issue. More than 440<br />

nuclear power stations are in operation<br />

around the globe, and they will<br />

all need to be decommissioned at<br />

some point. These facts serve to illustrate<br />

the great potential and the vast<br />

amount of research which this field<br />

entails.<br />

The professorship of Deconstruction<br />

and Decommissioning of Conventional<br />

and <strong>Nuclear</strong> Buildings at<br />

Karlsruhe Institute of Technology<br />

(KIT) was established in 20<strong>08</strong> and is<br />

dedicated exclusively to the last life<br />

cycle of a building, its deconstruction<br />

and decommissioning.<br />

There is no doubt that a nuclear<br />

power station can be safely decommissioned<br />

nowadays. Neither is there any<br />

doubt that optimization potential<br />

exists <strong>for</strong> many technologies and processes.<br />

There is also scope to further<br />

enhance the effectiveness of the automation<br />

and the robotics.<br />

This is precisely where this department<br />

comes in, concentrating its<br />

ef<strong>for</strong>ts on R&D projects, and always<br />

working in close collaboration with<br />

industry. These projects can then<br />

be comprehensively tested in the<br />

Institute’s own testing facility.<br />

The individual R&D projects address<br />

the following issues:<br />

p Reduction of secondary waste<br />

p Automation and remote operation<br />

of the processes<br />

p Per<strong>for</strong>mance optimization of existing<br />

processes<br />

p Development of new technologies<br />

p Management methods <strong>for</strong> decommissioning<br />

and deconstruction<br />

By way of example, we present a<br />

brief overview of the DefAhS (Defined<br />

ablation of highly rein<strong>for</strong>ced concrete<br />

structures) project, a collaborative<br />

project between Kraftanlagen Heidelberg<br />

GmbH, Herrenknecht AG, and<br />

KIT. The project objective was the<br />

in-depth removal of highly rein<strong>for</strong>ced<br />

concretes using one single tool and in<br />

one single operation. In 2019, the project<br />

achieved 2 nd place in the Innovation<br />

Awards at the Bauma, the world’s<br />

| Fig. 2.<br />

Detail of the combination of impact cutters and steel cutting inserts.<br />

largest construction machinery trade<br />

fair, and the innovation has been<br />

granted patents in numerous countries.<br />

The equipment innovatively<br />

combines cutting tools <strong>for</strong> concrete<br />

and steel in one milling drum. For the<br />

first time ever, it is now possible to<br />

automate the deep milling of cracks,<br />

<strong>for</strong> example.<br />

Another of our current projects is<br />

MASK (Magnetic Separation Method<br />

<strong>for</strong> the Reduction of Secondary Waste<br />

from the Water Abrasive Suspension<br />

Cutting Technique (MASK)), because<br />

the cutting up and disposal of the<br />

­reactor pressure vessel (RPV) and its<br />

related installations poses a considerable<br />

challenge during the decommissioning<br />

of a nuclear facility. One of<br />

the cutting techniques is the Water<br />

­Abrasive Suspension Cutting Technique<br />

(WASS), which is characterized<br />

by the high degree of flexibility of its<br />

modular application and the fact that<br />

it is impervious to the mechanical and<br />

thermal stresses in the material being<br />

cut. The abrasive that has to be<br />

Decommissioning and Waste Management<br />

Actual Research and Development Activities in the Field of Dismantling ı Sascha Gentes and Nadine Gabor


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 3.<br />

Prototype rig <strong>for</strong> the separation.<br />

admixed to facilitate the cutting process<br />

combines with the metal cuttings<br />

from the RPV, and this mixture itself<br />

has then also to be disposed of. The<br />

amount of secondary waste produced<br />

thereby is considerable, roughly doubling<br />

the total volume of radioactive<br />

waste. The disposal of secondary<br />

waste is very expensive and so this<br />

method of cutting up the reactor<br />

pressure vessel has fallen out of use<br />

despite its technical advantages.<br />

MASK was preceded by the research<br />

project known as NENAWAS (New<br />

Disposal Methods <strong>for</strong> the Secondary<br />

Waste of the Water Abrasive Suspension<br />

Cutting Technique), a close<br />

­collaboration between KIT and ­AREVA<br />

GmbH which succeeded in developing<br />

a separation method which can<br />

process the secondary waste from the<br />

water abrasive suspension cutting<br />

technique. It utilizes a prototype magnetic<br />

separation rig to separate the<br />

metal cuttings from the mixture of<br />

steel and abrasive material produced<br />

in the cutting process. The microscopic<br />

analysis of the separated abrasive<br />

shows that it is still contaminated<br />

with metal cuttings, however. Be<strong>for</strong>e<br />

the bulk of the secondary waste can be<br />

cleared <strong>for</strong> release, further investigation<br />

is required, and this is being done<br />

in the current MASK project. The<br />

objective here is to undertake basic<br />

research which allows the quality of<br />

the separation to be optimized to such<br />

a degree that the secondary waste<br />

can be disposed of in a conventional<br />

manner and thus allow the cutting<br />

technique to be used in the future <strong>for</strong><br />

| Fig. 4.<br />

Secondary waste: generation.<br />

the large number of decommissioning<br />

projects still to be undertaken. To<br />

this end, further experiments (with<br />

different types of steel to be cut and<br />

different processed mixtures) are<br />

being carried out on the existing test<br />

rig. Moreover, a numerical flow simulation<br />

of the magnetic filter is being<br />

produced. In the controlled area of<br />

the laboratory belonging to our project<br />

partner (Institute <strong>for</strong> <strong>Nuclear</strong><br />

Waste Disposal), the separation is<br />

being tested and evaluated under<br />

realistic conditions with a small,<br />

laboratory-scale test setup with active<br />

and activated materials to assess its<br />

suitability <strong>for</strong> the treatment of the<br />

radioactive waste.<br />

Robotics is also making greater<br />

and greater inroads into the field of<br />

deconstruction and decommissioning.<br />

This has led to the setting up of<br />

the ROBDEKON (Robotic Systems<br />

<strong>for</strong> Decontamination in Hazardous<br />

Environments) project, a competence<br />

center dedicated to research into fully<br />

autonomous and semi-autonomous<br />

robotic systems. The aim <strong>for</strong> the future<br />

is <strong>for</strong> such systems to carry out<br />

decontamination work autonomously,<br />

obviating the need <strong>for</strong> humans to<br />

­enter hazardous zones. As of mid-<br />

June 2018, the Federal Ministry of<br />

Education and Research has provided<br />

ROBDEKON with twelve million euros<br />

of funding as part of the »Research<br />

<strong>for</strong> Civil Security« program. It will<br />

ini tially run <strong>for</strong> four years, but the aim<br />

is <strong>for</strong> the competence center to continue<br />

in the long term. ROBDEKON<br />

is coordinated by the Fraunhofer<br />

Institute of Optronics, System Technologies<br />

and Image Exploitation<br />

IOSB. In addition to the Fraunhofer<br />

IOSB, other research institutions<br />

involved in the project are Karlsruhe<br />

Institute of Technology (KIT), the<br />

­German Research Center <strong>for</strong> Artificial<br />

Intelligence (DFKI), and the FZI<br />

Research Center <strong>for</strong> In<strong>for</strong>mation<br />

Technology. The industrial partners<br />

in the consortium are Götting KG,<br />

Kraftanlagen Heidelberg GmbH, ICP<br />

Ingenieurgesellschaft Prof. Czurda<br />

und Partner mbH, and KHG Kerntechnische<br />

Hilfsdienst GmbH.<br />

The intention is <strong>for</strong> ROBDEKON<br />

to become the national point of<br />

contact <strong>for</strong> issues appertaining to<br />

robotic systems <strong>for</strong> decontamination<br />

in hazardous environments. The<br />

competence center aims to set up a<br />

network of experts and users, and<br />

create an innovative environment <strong>for</strong><br />

new technologies <strong>for</strong> robot-assisted<br />

decontamination <strong>for</strong> its partners from<br />

science and industry.<br />

ROBDEKON has been established<br />

to explore and develop novel types of<br />

robotic systems <strong>for</strong> decontamination<br />

tasks. Its research topics are mobile<br />

robots <strong>for</strong> difficult terrain, autonomous<br />

construction equipment,<br />

robotic manipulators, and also decontamination<br />

concepts, planning<br />

algorithms, multi-sensorial 3D environment<br />

mapping, and teleoperation<br />

by means of virtual reality. Methods<br />

from the field of artificial intelligence<br />

enable the robot to per<strong>for</strong>m the<br />

tasks assigned to it either autonomously<br />

or semi-autonomously. While<br />

DECOMMISSIONING AND WASTE MANAGEMENT 395<br />

Decommissioning and Waste Management<br />

Actual Research and Development Activities in the Field of Dismantling ı Sascha Gentes and Nadine Gabor


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 396<br />

| Fig. 5.<br />

Prototype of automated milling system.<br />

the competence center is being set up,<br />

the work will initially concentrate on<br />

three relevant areas of application:<br />

the remediation of landfills and contaminated<br />

sites, the decommissioning<br />

and deconstruction of nuclear facilities,<br />

and the decontamination of<br />

facility components. By involving<br />

users at an early stage, we ensure that<br />

practical systems which reduce the<br />

risk <strong>for</strong> human operators and protect<br />

them from hazards are developed<br />

expeditiously.<br />

The Department of Deconstruction<br />

and Decommissioning of Conventional<br />

and <strong>Nuclear</strong> Buildings at Karlsruhe<br />

Institute of Technology is to explore<br />

and develop an approach that will<br />

help find a solution which uses robotic<br />

systems <strong>for</strong> the automated decontamination<br />

and clearance measurement<br />

of building structures in nuclear facilities.<br />

Automated decontamination<br />

focuses on treating a contaminated<br />

concrete wall and contaminated indoor<br />

spaces with the aid of a mobile<br />

work plat<strong>for</strong>m.<br />

To this end, the department is<br />

developing and constructing an<br />

automated surveying system which<br />

will utilize the latest measurement<br />

systems to explore its surroundings.<br />

It will display the contaminated locations<br />

on a 3D chart of its environment<br />

as a function of the radiation level,<br />

thus allowing the subsequent decontamination<br />

work to be carried out<br />

with greater efficiency.<br />

This overview is intended to show<br />

that R&D work is still required in the<br />

field of nuclear facility decommissioning,<br />

even in <strong>2020</strong>. The emphasis<br />

here is always on the involvement of,<br />

and close collaboration with industry,<br />

since this is the only way to work out a<br />

practical solution.<br />

In addition to issues relating<br />

­spe­cifically to nuclear facilities, great<br />

importance is also attached to all<br />

aspects concerning the demolition of<br />

conventional buildings. Some brief<br />

details are provided below. Basically,<br />

the waste products from the demolition<br />

process (asphalt, concrete, masonry,<br />

asbestos, man-made mineral<br />

fiber, …) are categorized as either<br />

hazardous or non-hazardous waste.<br />

The laws, ordinances and guidelines<br />

are extremely extensive, specific to a<br />

particular federal state in some cases,<br />

and subject to continual revisions<br />

and amendments. The thresholds<br />

above which asphalt is classified as<br />

hazardous waste differ by several<br />

­hundred mg PAH/kg across the individual<br />

federal states, <strong>for</strong> example.<br />

The figures stated below show<br />

just how important the demolition<br />

of conventional buildings is. In<br />

­Germany, around 2<strong>09</strong> million tonnes<br />

of waste are classified as „construction<br />

and demolition waste“ every year. In<br />

2015, this amounted to over 50 percent<br />

of the total waste produced<br />

(402.2 million tonnes). In conjunction<br />

with the Circular Economy Act (Kreislaufwirtschaftsgesetz)<br />

and the stipulations<br />

on recycling rates, this fact<br />

clearly emphasizes the considerable<br />

research potential in this field. The<br />

research topics addressed are the<br />

automated separation of different<br />

types of waste, optimization potentials<br />

in relation to environmental<br />

release and pollution during demolition,<br />

and also automation and remote<br />

handling. This relates in particular<br />

to the handling of “hazardous waste”.<br />

The task is to recognize these potentials,<br />

develop optimization approaches,<br />

and implement pilot projects<br />

with a specific objective.<br />

Selective demolition requires that<br />

questions about how to handle<br />

hazardous waste and pollutants, and<br />

how to comply with the stipulations<br />

regarding type-specific collection and<br />

disposal of the demolition material,<br />

be clarified in advance. All these<br />

issues mean the demolition of conventional<br />

buildings is an exciting field<br />

with extensive research potential.<br />

The redevelopment of existing<br />

building structures and building on<br />

“brown field sites”, too, will become<br />

ever-expanding fields of research in<br />

the future. To do justice to the<br />

demands, the Department of Deconstruction<br />

and Decommissioning of<br />

Conventional and <strong>Nuclear</strong> Buildings is<br />

addressing this future-oriented field<br />

in its research, science and teaching.<br />

Now more than ever, the graduates of<br />

today must have a good general<br />

grounding in topics such as the construction,<br />

operation, and also the<br />

decommissioning of buildings. To<br />

illustrate this, we have included part<br />

of a survey carried out at universities<br />

and universities of applied sciences<br />

on the topic of “demolition-related<br />

course content in civil engineering”.<br />

The survey reveals that “demolition”<br />

and “decommissioning” and other<br />

related subjects are not covered in<br />

detail. Lectures on more detailed<br />

issues of decommissioning and deconstruction<br />

are not usually part of the<br />

curriculum.<br />

This is precisely where we come in,<br />

by offering lectures which train young<br />

graduates in these important and<br />

sustainable subjects.<br />

The deconstruction and decommissioning<br />

of conventional and<br />

nuclear buildings is there<strong>for</strong>e a<br />

discipline whose graduates will be<br />

in great demand and which will<br />

guarantee secure employment in the<br />

future.<br />

Authors<br />

Prof. Dr.-Ing. Sascha Gentes<br />

sascha.gentes@kit.edu<br />

Dr.-Ing. Nadine Gabor<br />

Karlsruhe Institut of Technology<br />

Institute of Technology and<br />

Management in Construction<br />

Am Fasanengarten<br />

76131 Karlsruhe, Germany<br />

Decommissioning and Waste Management<br />

Actual Research and Development Activities in the Field of Dismantling ı Sascha Gentes and Nadine Gabor


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

A Geopolymer Waste Form <strong>for</strong><br />

Technetium, Iodine and Hazardous Metals<br />

Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg<br />

We have developed geopolymer waste <strong>for</strong>ms <strong>for</strong> the solidification of waste streams (Table 1)<br />

that will be generated at the Han<strong>for</strong>d site in the State of Washington, USA. At the Han<strong>for</strong>d site,<br />

about 205,000 m 3 of liquid radioactive waste is stored in 177 underground tanks. These wastes will<br />

be immobilized at the US Department of Energy’s (DOE) “ Han<strong>for</strong>d Tank Waste Treatment and<br />

Immobilization Plant” (WTP), which is under construction. The wastes will be separated into<br />

high-level waste (HLW) and low-activity waste (LAW), both of which will be vitrified. The vitrified<br />

highlevel waste will be disposed in an offsite underground repository. The vitrified LAW will be<br />

disposed onsite in the “Integrated Disposal Facility” (IDF).<br />

Vitrification of LAW and HLW will generate secondary ­liquid<br />

wastes, called Han<strong>for</strong>d secondary waste (HSW), which includes<br />

process condensates and liquid effluents from ­off-gas<br />

treatment systems. HSW will be sent to an ­effluent treatment<br />

­facility (ETF) <strong>for</strong> further treatment and solidification<br />

(not vitrification) and then disposed in the IDF as well [1].<br />

To support the evaluation and selection of waste <strong>for</strong>ms<br />

suitable <strong>for</strong> the solidification and disposal of ­secondary<br />

waste streams from the WTP, the Pacific Northwest ­National<br />

Laboratory (PNNL) in Richland, WA, conducted a waste<br />

<strong>for</strong>m testing program <strong>for</strong> the DOE [2, 3]. Our geopolymers<br />

(patented as DuraLith [4]) and two other materials (cast<br />

stone [5], and ceramicrete [6]) par ticipated and were tested<br />

under the conditions laid out by PNNL. Cast Stone consists<br />

of Portland cement, ground granulated blast furnace slag<br />

and Class F fly ash (about 10, 45, 45 wt. %) and Ceramicrete<br />

of 12.5 wt. % MgO, 42.5 wt. % KH 2 PO 4 and, 45.0 wt. %<br />

Class C fly ash. For the DuraLith geopolymer the most<br />

important results will be reported here.<br />

Materials and methods:<br />

DuraLith geopolymer waste <strong>for</strong>ms consist of two binders,<br />

i.e., ground granulated blast furnace slag (FS) and<br />

metakaolin (MK), an additional source of silica, river sand,<br />

and an activator, here the waste solution with extra sodium<br />

hydroxide added. To support fixation of 99 Tc (replaced by<br />

ReO 4 - ) SnF 2 was added to reduce Re 7+ (Tc 7+ ) to Re 4+<br />

(Tc 4+ ), which <strong>for</strong>ms insoluble ReO 2 (TcO 2 ). IONEX Ag 900,<br />

an Ag-zeolite (Ag-Z) was used to precipitate I - as AgI in the<br />

nano pores of the zeolite.<br />

HSW will be generated as soon as vitrification operations<br />

begin. Table 1 shows four projected compositions of<br />

HSW streams (S1 to S4) that were provided <strong>for</strong> testing by<br />

PNNL. S1 to S3 are alkaline solutions (high OH - ) and of<br />

similar composition. S4 is not alkaline and contains<br />

ammonia salts and has the highest concentrations of<br />

­iodine and technetium. All wastes contain the heavy<br />

­metals Cr, Cd, Pb, Ag, As, Hg with concentrations ranging<br />

from about 1 ppm (Cd) to 1000 ppm (Cr). Mercury was not<br />

included in this study <strong>for</strong> safety reasons.<br />

The solid constituents were mixed and the activator<br />

solution was added. Mixing was continued until a pour able<br />

paste <strong>for</strong>med, which took a few minutes. The paste was<br />

transferred into cylindrical plastic molds where it hardened<br />

within 1 to 3 hours, depending on composition. All samples<br />

were cured at room temperature <strong>for</strong> at least 28 days be<strong>for</strong>e<br />

tests were conducted. No bleeding, swelling, salt deposition,<br />

or cracking was observed. Example recipes <strong>for</strong> MKand<br />

FS-based waste <strong>for</strong>ms used <strong>for</strong> waste stream S1 are<br />

Planned entry <strong>for</strong><br />

shown in Table 2 <strong>for</strong> a typical lab-scale batch of about 4 kg.<br />

Loading of waste solids was about 2.5 wt. %.<br />

IDF waste acceptance requires that materials testing<br />

follows the pro cedures referred to below, that the quality<br />

assurance programs NQA-1 is in place, and that the laboratory<br />

is audited. Our laboratory meets these requirements.<br />

Compressive strength was measured after 28 days<br />

­following ASTM C39. Pieces of the crushed cylinders were<br />

used to prepare the required particle size fraction <strong>for</strong> the<br />

“Toxicity characteristic leaching procedure” (TCLP),<br />

SW846 Test Method 1311. This procedure was applied to<br />

measure the release of hazardous metals. The method<br />

yields data that can be compared to the U.S. Environmental<br />

Protection Agency”s Universal Treatment Standards<br />

(UTS) [7], which are part of the waste acceptance criteria<br />

<strong>for</strong> HSW waste <strong>for</strong>ms at the IDF. Direct current plasma<br />

atomic emission ­spectroscopy (DCP-AES) was ­employed to<br />

determine the release of Cr, Ag, As, Cd, Cu, Pb, Sn from the<br />

DuraLith waste <strong>for</strong>ms during leaching. A small amount of<br />

the crushed cylinders was used to prepare samples <strong>for</strong><br />

SEM/EDS analysis.<br />

Best Paper<br />

Award<br />

The papers<br />

“A geopolymer waste<br />

<strong>for</strong>m <strong>for</strong> technetium,<br />

iodine and hazardous<br />

metals” by<br />

Werner Lutze,<br />

Weiliang Gong, Hui<br />

Xu and Ian L. Pegg<br />

and “Code and data<br />

enhancements of the<br />

MURE C++ environment<br />

<strong>for</strong> Monte-Carlo<br />

simulation and<br />

depletion” by<br />

Dr. Maarten Becker<br />

(featured in previous<br />

<strong>atw</strong>) have been<br />

awarded as<br />

Best Papers of<br />

KERNTECHNIK <strong>2020</strong>,<br />

which un<strong>for</strong>tunately<br />

had to be cancelled<br />

due to Covid-19.<br />

Constituents S1 S2 S3 S4<br />

Al(OH) 3 9.39E-02 1.14E-01 9.22E-02 4.24E-02<br />

Si 1.88E-03 2.04E-03 7.74E-04 1.39E-02<br />

K 5.82E-04 6.51E-04 2.18E-03 2.87E-02<br />

NH 4<br />

+<br />

- - - 4.41E-01<br />

OH - 3.98E-01 4.35E-01 2.45E-01 1.02E-<strong>08</strong><br />

NO 3<br />

-<br />

3.28E-01 1.90E-01 3.97E-01 1.13E+00<br />

CO 3<br />

2-<br />

2.28E-02 4.66E-02 3.94E-02 1.04E-02<br />

Cl - 2.25E-02 2.17E-02 2.91E-02 1.04E-02<br />

NO 2<br />

-<br />

1.20E-02 1.05E-02 3.83E-02 4.31E-02<br />

PO 4<br />

3-<br />

6.87E-03 4.85E-03 6.03E-03 5.10E-03<br />

SO 4<br />

2-<br />

4.41E-03 5.81E-03 5.14E-03 4.36E-02<br />

F - 5.57E-04 3.75E-04 4.42E-04 1.02E-<strong>08</strong><br />

Cr 2.03E-04 2.03E-04 2.03E-04 1.<strong>09</strong>E-03<br />

Ag 6.27E-06 6.27E-06 6.27E-06 2.35E-05<br />

As 3.48E-05 3.48E-05 3.48E-05 1.61E-05<br />

Cd 1.57E-06 1.57E-06 1.57E-06 2.16E-05<br />

Hg 1.13E-05 1.13E-05 1.13E-05 5.30E-06<br />

Pb 8.99E-06 8.99E-06 8.99E-06 8.28E-06<br />

Tc 1.81E-05 1.81E-05 1.81E-05 5.59E-04<br />

I - 4.62E-06 4.62E-06 4.62E-06 6.29E-05<br />

Total organic carbon 9.39E-02 1.14E-01 9.22E-02 4.42E-02<br />

| Tab. 1.<br />

Compositions of Han<strong>for</strong>d secondary waste stream simulants (mol/l).<br />

DECOMMISSIONING AND WASTE MANAGEMENT 397<br />

Decommissioning and Waste Management<br />

A Geopolymer Waste Form <strong>for</strong> Technetium, Iodine and Hazardous Metals ı Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 398<br />

Order<br />

of addition<br />

Chemical/<br />

material<br />

Assay<br />

S1-2xMKR<br />

Mass (g)<br />

S1-2xFSR<br />

1 HSW simulant 1.00 658.4 654.7<br />

2 SnF2 0.98 15.0 15.0<br />

3 KOH 0.90 499.4 232.6<br />

4 NaOH 0.98 59.5 118.6<br />

5 Fumed silica 0.96 466.1 333.2<br />

6 Metakaolin 0.96 833.5 477.9<br />

6 Furnace slag 1.00 547.5 1183.8<br />

6 Fine river sand 1.00 733.0 723.3<br />

6 Ag-Z 900 1.00 38.6 38.1<br />

7 Silica fume filler 1.00 77.2 38.1<br />

| Tab. 2.<br />

Example recipes <strong>for</strong> FS-based and MK-based DuraLith waste <strong>for</strong>ms applied to S1.<br />

Sample ID Ag As Cd Cr Cu Pb Sn<br />

S1-2xFS


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Table 4 shows the results obtained with the ­ANSI/ANS-<br />

16.1 test <strong>for</strong> the MK-based geopolymer samples. The results<br />

<strong>for</strong> the FS-based samples are similar. Results are<br />

shown <strong>for</strong> rhenium and sodium. Iodine will be discussed<br />

separately. Most of the LI values <strong>for</strong> rhenium are >9, i.e.,<br />

greater than the required lower limit of 9.0 <strong>for</strong> technetium.<br />

The LI values <strong>for</strong> sodium are about 9, i.e., three orders of<br />

magnitude higher than the required minimum value of 6.<br />

The objective of this work was to develop a waste <strong>for</strong>m<br />

that is easy to make at ambient temperature and exhibits<br />

high retention <strong>for</strong> technetium, iodine and the hazardous<br />

metals contained in HSW. To find out how well the<br />

simulation of technetium by rhenium would be, we shared<br />

a MK-based DuraLith recipe with Pierce et al. [8] who<br />

conducted testing with 99 Tc. These authors used the EPA<br />

Method 1315, which is equivalent to ANSI/ANS-16.1.<br />

Pierce et al. [8] reported an average leaching index of<br />

about 11, which is one to two orders of magnitude higher<br />

than our index <strong>for</strong> Re. In another study, Mattigod et al. [9]<br />

conducted leaching experiments with 99 Tc and Re, using<br />

one of our FS-based DuraLith recipes. The authors<br />

concluded that rhenium does not simulate technetium<br />

very well. The leachability indices <strong>for</strong> Re were one to three<br />

orders of magnitude lower than those <strong>for</strong> 99 Tc.<br />

Ag-Z was employed to fixate iodine by precipitating AgI<br />

in the nanopores of the zeolite. We found that at least some<br />

of our chemicals and/or raw materials used to prepare the<br />

geo polymer waste <strong>for</strong>ms contained iodide as an impurity.<br />

The level of the contamination was significant enough to<br />

compromise the ANSI/ANS-16.1 leach tests. There<strong>for</strong>e,<br />

the results <strong>for</strong> iodine could not be quantified correctly.<br />

Pierce et al. [8] reported an iodine LI of around 7 <strong>for</strong> a<br />

MK-based ­DuraLith geopolymer <strong>for</strong> S1 with Ag-Z, spiked<br />

with stable 127 I. This value is com parable to data collected<br />

by Pierce et al. <strong>for</strong> two other waste <strong>for</strong>ms, Cast Stone and<br />

Ceramicrete [8, 10, 11]. None of these waste <strong>for</strong>ms met the<br />

minimum iodine LI of 11 <strong>for</strong> disposal at the IDF. An evaluation<br />

by PNNL concluded that Cast Stone should be the<br />

­final choice, because of its greater maturity concerning<br />

large-scale production of the waste <strong>for</strong>m.<br />

We have studied iodide fixation by Ag-Z directly in the<br />

waste solution (S1). For example, with a molar ratio<br />

Ag/I = 5, Ag-Z sequestered 99.9 % of the iodide in the<br />

simulated HSW solution (2M Na + ), spiked with 100 ppm<br />

iodide. Though silver-based scavengers are very efficient in<br />

removing iodine from HSW, its long-term stability may be<br />

impaired as the AgI encapsulated in the zeolite may be<br />

destabilized by sulfide released ­during alkali-activation of<br />

blast furnace slag. More work needs to be done to evaluate<br />

the significance of this process. ­Solidification of the waste<br />

stream ­after iodide separation may yield a sufficiently high<br />

LI <strong>for</strong> the geo polymer.<br />

Scale-up testing:<br />

The feasibility of larger-scale batches of FS-based geopolymer<br />

waste <strong>for</strong>ms was tested [12]. For example, a batch<br />

was produced to fill a 55-gallon (2<strong>08</strong> liters) drum.<br />

Core-drilling and cutting showed that a visibly homogeneous<br />

monolith was obtained. Further more, on the<br />

same scale we studied effects of composition changes on<br />

workability and initial setting with 6 M Na + HSW simulant<br />

(S1). Blast furnace slag varied in concen tration and parts<br />

of the slag were ­replaced by Class F fly ash (ASTM C618).<br />

The less reactive class F fly ash extended the workable time<br />

(up to the time of initial setting) up to 15 hours, depending<br />

on the weight fraction of fly ash. This ­observation gave rise<br />

to an in-depth study in the laboratory on effects of composition<br />

on pro duction-related and some other properties<br />

Sample ID Concentration (mg/l) Leachability index<br />

of these geopolymer waste <strong>for</strong>ms. This study has been<br />

completed and will be reported elsewhere.<br />

Conclusions<br />

A geopolymer waste <strong>for</strong>m was ­de­veloped to immobilize<br />

Han<strong>for</strong>d secondary lowlevel radioactive waste streams.<br />

Two other materials were ­under consideration <strong>for</strong> a final<br />

selection. The final decision was between Cast Stone and<br />

the geopolymer, which per<strong>for</strong>med equally well. Cast Stone<br />

was selected because of greater experience with technicalscale<br />

production. None of the waste <strong>for</strong>ms complied with<br />

the required degree of fixation of iodine. We have shown<br />

that iodine separation by Ag-Z prior to ­solidification is very<br />

effective. The decontamination factor is about 10 3 . We<br />

have not yet tested whether adding the precipitated iodine<br />

during solidification of the depleted waste stream would<br />

increase LI to 11.<br />

References<br />

[1] Pacific Northwest National Laboratory, Han<strong>for</strong>d Site Secondary Waste Roadmap, PNNL-18196<br />

(20<strong>09</strong>) Pacific Northwest National Laboratory, Richland, WA<br />

[2] R.L. Russell, M.J. Schweiger, J.H. Westsik, Jr., P.R. Hrma, D.E. Smith, A.B. Gallegos, M.R. Telander,<br />

S.G. Pitman, Low temperature waste immobilization testing, PNNL-16052 Rev 1 (2006), Pacific<br />

Northwest National Laboratory, Richland, WA<br />

[3] E.M. Pierce, R.J. Serne, W. Um., S.V. Mattigod, J.P. Icenhower, N.P. Qafoku, J.H. Westsik, Jr., R.D.<br />

Scheele, Review of potential candidate stabilization technologies <strong>for</strong> liquid and solid secondary<br />

waste streams, PNNL-19122 (2010), Pacific Northwest National Laboratory, Richland, WA<br />

[4] W. Gong, W. Lutze, I.L. Pegg, U. S. Patent No. 7,855,313 B2 (2010)<br />

[5] S.K. Sundaram, K.E. Parker, M.E. Valenta, S.G. Pitman, J. Chun, C.-W. Chung, M.L. Kimura, C.A.<br />

Burns, W. Um, J.H. Westsik, Jr., Secondary waste <strong>for</strong>m development and optimization -Cast<br />

Stone, PNNL-20159 (2011), Pacific Northwest National Laboratory, Richland, WA<br />

[6] D. Singh, R. Ganga, J. Gaviria, Y. Yusufoglu, Secondary waste <strong>for</strong>m testing: Ceramicrete<br />

phosphate bonded ceramics, ANL-11/16 (2011), Argonne National Laboratory, Chicago, IL<br />

[7] 40 CFR 268. 2002, Land disposal restrictions; Code of Federal Regulations, U.S. Environmental<br />

Protection Agency, Washington, DC<br />

[8] E.M. Pierce, R.J. Serne, W. Um., S.V. Mattigod, J.P. Icenhower, N.P. Qafoku, J.H. Westsik, Jr., R.D.<br />

Scheele, Review of potential candidate stabilization technologies <strong>for</strong> liquid and solid secondary<br />

waste streams, PNNL-19122 (2010), Pacific Northwest National Laboratory, Richland, WA<br />

[9] S.V. Mattigod, J.H. Westsik, Jr., C.W. Chung, M.J. Lindberg, and K.E. Parker, Waste acceptance<br />

testing of secondary waste <strong>for</strong>ms: Cast Stone, Ceramicrete and DuraLith, PNNL-20632 (2011),<br />

Pacific Northwest National Laboratory, Richland, WA<br />

[10] S.K. Sundaram, K.E. Parker, M.E. Valenta, S.G. Pitman, J. Chun, C.-W. Chung, M.L. Kimura, C.A.<br />

Burns, W. Um, J.H. Westsik, Jr., Secondary waste <strong>for</strong>m development and optimization -Cast<br />

Stone, PNNL-20159 (2011), Pacific Northwest National Laboratory, Richland, WA<br />

[11] D. Singh, R. Ganga, J. Gaviria, Y. Yusufoglu, Secondary waste <strong>for</strong>m testing: Ceramicrete<br />

phosphate bonded ceramics, ANL-11/16 (2011), Argonne National Laboratory, Chicago, IL<br />

[12] G.B. Josephson, J.H. Westsik, Jr., R.P. Pires, J.L. Bick<strong>for</strong>d, M.W. Foote, Engineering-scale<br />

demonstration of DuraLith and Ceramicrete waste <strong>for</strong>ms, PNNL20751 (2011), Pacific Northwest<br />

National Laboratory, Richland, WA<br />

Authors<br />

Prof. Werner Lutze<br />

Weiliang Gong<br />

Hui Xu<br />

Ian L. Pegg<br />

The Catholic University of America<br />

620 Michigan Ave NE<br />

20064 Washington, DC, USA<br />

Na Re Na Re<br />

S1-2xMKR-L01 4.56 0.004 9.6 10.1<br />

S1-2xMKR-L02 5.30 0.002 9.3 10.6<br />

S1-2xMKR-L03 9.91 0.003 9.3 10.8<br />

S1-2xMKR-L04 10.24 0.016 9.2 9.3<br />

S1-2xMKR-L05 7.86 0.041 9.2 8.2<br />

S1-2xMKR-L06 14.60 0.041 8.5 8.1<br />

S1-2xMKR-L07 90.82 0.041 8.4 8.0<br />

S1-2xMKR-L<strong>08</strong> 90.86 0.078 8.7 9.3<br />

S1-2xMKR-L<strong>09</strong> 114.2 0.077 8.7 9.5<br />

S1-2xMKR-L10 127.11 0.073 8.6 9.6<br />

| Tab. 4.<br />

Results of ANSI/ANS-16.1 leach test <strong>for</strong> selected DuraLith waste <strong>for</strong>ms.<br />

DECOMMISSIONING AND WASTE MANAGEMENT 399<br />

Decommissioning and Waste Management<br />

A Geopolymer Waste Form <strong>for</strong> Technetium, Iodine and Hazardous Metals ı Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 400<br />

Planned entry <strong>for</strong><br />

Decommissioning of <strong>Nuclear</strong> <strong>Power</strong><br />

Plants: Waste Streams and Release<br />

Measurements<br />

Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann<br />

The material from nuclear power<br />

plants must be checked <strong>for</strong> radioactive<br />

contamination be<strong>for</strong>e it can be recycled.<br />

Roughly spoken, three classes<br />

exist in respect to radioactivity:<br />

release material, material with short<br />

living radioactive isotopes and material<br />

with long living radioactive isotopes<br />

like Uranium or Plutonium.<br />

The release measurement procedure<br />

is of highest importance because<br />

the material can be immediately<br />

recycled and does not need special<br />

handling and storages.<br />

This paper will give a survey over<br />

measurement methods, per<strong>for</strong>mance<br />

consideration, data management and<br />

quality control. All these topics have<br />

also to be considered under economical<br />

aspects.<br />

Measurement methods<br />

The inventory of nuclear power plants<br />

consists of isotopes emitting gammas,<br />

beta particles, alpha particles and<br />

neutrons (other modes like internal<br />

conversion, spontaneous fission etc.<br />

are not considered here).<br />

Alpha particles are difficult to<br />

measure with high sensitivity because<br />

of the high stopping power by material<br />

or air. Neutrons can easily penetrate<br />

material, but the detection limit<br />

is very high because of the weak<br />

During the next 20 years a large number of nuclear installations have to be taken out of operation<br />

and have to be decommissioned up to the level of “green field”. According to German laws, all<br />

material should be recycled, whenever possible. Otherwise the material has to be transported to<br />

long term storages or disposals, which are actually not planed finally. From the economical point of<br />

view, the recycling is there<strong>for</strong>e the preferred solution.<br />

interaction with the detectors. For this<br />

reason, alphas and neutrons are rarely<br />

measured. This issue can be solved,<br />

because almost all isotopes emitting<br />

alphas or neutrons also emit gamma<br />

radiation. The situation concerning<br />

beta particles is similar. The main<br />

difference is that there are some isotopes<br />

emitting betas, but without<br />

measurable gammas like Sr-90. In<br />

this case the key nuclide method<br />

must be applied, which means, that<br />

gamma emission of those isotopes is<br />

measured, which mostly occur<br />

together with the hidden beta emitter<br />

(e.g. Cs-137 is a key nuclide <strong>for</strong> Sr-90).<br />

In principle the betas can be<br />

measured with beta counters, but this<br />

is only possible <strong>for</strong> isotopes near the<br />

surface of the material and is not good<br />

enough to estimate the total beta<br />

emitters in the waste.<br />

In conclusion the measurement of<br />

gamma emitters is the best way<br />

because it is non-invasive method and<br />

is partly isotope specific. There is also<br />

the possibility to analyse the material<br />

in a radiochemical laboratory, where<br />

alpha or neutron emitters can be<br />

measured with high sensitivity and<br />

precision. But this is not a 100 %<br />

­efficiency measurement. There<strong>for</strong>e,<br />

the radiochemical analysis is only<br />

used <strong>for</strong> quality control, where the big<br />

ef<strong>for</strong>t <strong>for</strong> sample taking, transport to<br />

the laboratory, waiting <strong>for</strong> results has<br />

not the crucial role.<br />

To measure the gamma emission<br />

detectors are required, which have<br />

high sensitivity and high energy<br />

resolution. The energy resolution is<br />

important to identify the gamma<br />

emitting isotopes and to measure<br />

gamma lines with low detection<br />

limits. Nowadays only one detector<br />

type exists that completely fulfils<br />

these requirements: the HPGe detector.<br />

The price which must be paid <strong>for</strong><br />

its excellent characteristics is the<br />

cooling of the detector crystal. This<br />

can be made by using liquid Nitrogen<br />

or by electric cooling systems. Twenty<br />

years be<strong>for</strong>e, these detectors were<br />

utilised only in laboratory environments<br />

because of their sensitivity to<br />

sound waves and electric fields from<br />

servo-drives, etc. Actually, devices are<br />

available which are resistant against<br />

rough environmental conditions and<br />

ready <strong>for</strong> field application.<br />

The measurement, data readout<br />

and spectra analysis can be per<strong>for</strong>med<br />

automatically, including the un folding<br />

of spectra with overlapping peaks<br />

from different isotopes. Different<br />

programs to analyse the spectra,<br />

which were tested by the IAEA [1], are<br />

existing today. One of these programs<br />

| Fig. 1.<br />

Waste ideal to build a waste stream (after processing: i.e. crushing, sieving etc.).<br />

| Fig. 2.<br />

Waste to be pre-processed be<strong>for</strong>e building a waste stream.<br />

Decommissioning and Waste Management<br />

Decommissioning of <strong>Nuclear</strong> <strong>Power</strong> Plants: Waste Streams and Release Measurements ı Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 3.<br />

Automated system <strong>for</strong> crashing, sieving, weighing and filling on a conveyor belt.<br />

is the program GAMMA-W, which was<br />

upgraded to calculate uncertainties<br />

according the current norm DIN ISO<br />

11929 [2].<br />

Waste streams<br />

The measured activities do not only<br />

depend on the intensity of measured<br />

signals of the detector, but also<br />

from absorption effects of the waste.<br />

There<strong>for</strong>e, the measured activities<br />

have to be corrected to compensate<br />

such ­influences. Considering large<br />

quantities of waste under measurement<br />

during decommissioning, waste<br />

streams have to be defined, which<br />

consist of similar materials in respect<br />

to absorption properties and mass<br />

densities (Figure 1 and Figure 2).<br />

Inside a stream similar correction can<br />

be applied. Typical streams are rubble<br />

from building demolition, soil, metal<br />

waste, etc. The homogeneity of the<br />

material can be enhanced by crushing<br />

and sieving of the material (Figure 3).<br />

The material preparation <strong>for</strong> the<br />

measurement is an important aspect<br />

because this influences the reliability<br />

of the results and allows increases the<br />

material throughput.<br />

For large quantities two solutions<br />

exist in respect to the measurement<br />

arrangement: The first solution is the<br />

| Fig. 4.<br />

Drum Measurement System.<br />

filling of the material from a stream<br />

into drums with a typical volume<br />

between 200 l and 400 l. The drums<br />

are placed on a rotary table and the<br />

HPGe detector is placed at the side of<br />

the drum (Figure 4).<br />

Different measurement modes are<br />

possible like the measurement of the<br />

whole drum or a slice measurement<br />

by using a horizontal slit collimator<br />

and changing the HPGe position vertically.<br />

An alternative solution is the filling<br />

of the material on a conveyor band as<br />

a thin (typically 10 cm high) layer<br />

(Figure 5). The material moves by the<br />

conveyor and passes the detector<br />

system mounted above the conveyor<br />

(Figure 6). The absorption effects are<br />

minimized because no steel is used<br />

like in the case of drums and the waste<br />

thickness is small. To enhance the<br />

­detection efficiency, the detector<br />

system can be constructed by an array<br />

of HPGe detectors.<br />

The advantage of this construction<br />

is the high material throughput of up<br />

to 100 tons per hour (practical value<br />

of tests at Hanau, Germany). The<br />

disadvantage is that the measurement<br />

time cannot be changed freely during<br />

the measurement. There<strong>for</strong>e, it is<br />

important to run the system with<br />

material of the same material property<br />

(i.e. from the same waste stream).<br />

The throughput <strong>for</strong> a drum configuration<br />

is smaller: using a 400 l<br />

drum filled with material having a<br />

density of 2.3 g/cm 3 , measuring<br />

3 minutes and with additional<br />

2 ­minutes <strong>for</strong> drum handling (transport<br />

to the turn table, positioning at<br />

the table, removing from the table and<br />

transport to a store position) gives a<br />

throughput of 400*2.3/1000/5*60<br />

= 11 tons per hour. This seems to be<br />

an upper limit where the drum filling<br />

and emptying procedure are not taken<br />

into account.<br />

The values as given above are<br />

taken from a realistic scenario. The<br />

requirements in respect to detection<br />

limits, matrix properties of the waste<br />

and configuration details like the<br />

number of detectors etc. can influence<br />

the throughput.<br />

The metal waste can have three<br />

­different final destinations: unrestricted<br />

use, melting and filling at<br />

disposal sites. In all these cases the<br />

waste is released after measurement<br />

of radioactive materials.<br />

For the different sub-classification<br />

of released material (i.e. free release,<br />

restricted release, disposal) the<br />

measured data have to be analysed.<br />

The analysis starts with the data taken<br />

DECOMMISSIONING AND WASTE MANAGEMENT 401<br />

| Fig. 5.<br />

Rubber (left) and soil* (right) filled on a conveyor belt (*with courtesy of FBFC <strong>International</strong>).<br />

Decommissioning and Waste Management<br />

Decommissioning of <strong>Nuclear</strong> <strong>Power</strong> Plants: Waste Streams and Release Measurements ı Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 402<br />

| Fig. 6.<br />

HPGe detectors, liquid Nitrogen cooling (left), electrical cooling (right).<br />

<strong>for</strong> individual batches: a batch can be<br />

a drum filled with waste or a certain<br />

section of the conveyor belt. The batch<br />

is characterised by the material and<br />

the mass of waste contained. After<br />

that, <strong>for</strong> each batch the activity limit is<br />

calculated as the upper limit under<br />

consideration of all possible errors.<br />

This limit value must be lower than<br />

the limit values as given by law.<br />

If batches with very low limits are<br />

separated from batches with higher<br />

limits and if the two groups are filled<br />

into separate containers, then the<br />

content of the low value container<br />

may have significantly lower activity<br />

limits compared with the individual<br />

batches. The reason is that the average<br />

mass is increased and equalises the<br />

fluctuation of the values of the individual<br />

batches. In case of the use of<br />

conveyor belts, this analysis can be<br />

made online and can separate the<br />

batches into streams <strong>for</strong> the different<br />

release classifications.<br />

Up to this point, data (i.e. spectra)<br />

are analysed from measurement of<br />

the complete batch. It is also possible,<br />

to read out the spectra in short time<br />

intervals and analyse the measured<br />

data in respect to Hot Spots. This<br />

special analysis is parallel to the batch<br />

activity analysis. If Hot Spots are<br />

found, the batch is separated, the Hot<br />

Spot can be eliminated by e.g. hand<br />

measurements and the rest of the<br />

batch is given back <strong>for</strong> a new run.<br />

Quality control<br />

The measurement results depend<br />

from the programs used to analyse<br />

measured data and from the error estimation.<br />

Because the detection limits<br />

<strong>for</strong> release waste are very low, small<br />

differences in the used algorithms or<br />

in the method of error calculation may<br />

influence the decision <strong>for</strong> release.<br />

There<strong>for</strong>e, it is very important to<br />

check the quality of measurements<br />

independently. This can be done by<br />

sampling and analysing in a radiochemical<br />

laboratory. As long as the<br />

activities measured in the radiochemical<br />

laboratory are lower than<br />

the values from the batch measurements<br />

there is no reason to distrust<br />

the values from the batch measurements.<br />

Otherwise, safety factors have<br />

to be introduced into the data analysis.<br />

The details of data analysis and<br />

result interpretation have to be agreed<br />

with the state authorities.<br />

Economic considerations<br />

The release measurement is a balance<br />

act between environmental requirements<br />

(protecting the population<br />

from radioactive radiation) and financial<br />

expenses <strong>for</strong> the measurement.<br />

The most important parameters are<br />

detection limits and throughput.<br />

It must be considered that detection<br />

limits depend not only from detector<br />

sensitivities and matrix absorption<br />

effects, but also from the natural<br />

occurring radioactive materials at the<br />

location where the measurement system<br />

should be erected. If necessary<br />

special shielding has to be provided.<br />

The throughput depends not only<br />

from a high-speed measurement but<br />

also from the speed of material preparation<br />

like crashing, sieving or<br />

weighing of the batches and from<br />

the removal of the material after<br />

measurement. The slowest process<br />

­defines the throughput.<br />

The investment <strong>for</strong> the equipment<br />

is high and there<strong>for</strong>e it has also to be<br />

considered what to do with the<br />

equipment after the measurement<br />

campaign. A modular structured<br />

system which can be containerised<br />

<strong>for</strong> easy transport is a good solution<br />

to use the equipment <strong>for</strong> a long time<br />

period. The concept <strong>for</strong> decontamination<br />

of the equipment has also to be<br />

included in the considerations.<br />

Conclusion<br />

For the release measurement of a<br />

large amount of waste there are two<br />

general solutions: drum systems and<br />

conveyor belt systems. Both solutions<br />

are complementing each other. The<br />

conveyor belt system is favoured in<br />

respect to the highest throughput.<br />

The waste stream concept enhances<br />

the quality of the measurement and<br />

optimises the throughput at the same<br />

time. From economical point of view,<br />

the careful planning of material<br />

throughput as well as the reuse of the<br />

installed systems <strong>for</strong> next projects<br />

helps to save financial resources. The<br />

installations must be carefully designed<br />

and agreed with the responsible<br />

authorities. Data available <strong>for</strong> commissioned<br />

installations are helpful to<br />

optimize the layout.<br />

References<br />

[1] M. Blaauw et al., The 1995 IAEA intercomparison of gamma<br />

ray spectrum analysis software, <strong>Nuclear</strong> Instruments and<br />

Methods in Physics Research A 387 (1997) 416-432<br />

[2] Dr. Westmeier GmbH, Private in<strong>for</strong>mation from Dr. Westmeier,<br />

Ebsdorfergrund-Mölln, Germany<br />

Authors<br />

Dr. Marina Sokcic-Kostic<br />

marina.sokcic-kostic@<br />

nukemtechnologies.de<br />

Dr. Christoph Klein<br />

Dr. Frank Scheuermann<br />

NUKEM Technologies Engineering<br />

Services GmbH<br />

Industriestr. 13<br />

63755 Alzenau, Germany<br />

Decommissioning and Waste Management<br />

Decommissioning of <strong>Nuclear</strong> <strong>Power</strong> Plants: Waste Streams and Release Measurements ı Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Ventilation Concepts<br />

<strong>for</strong> <strong>Nuclear</strong> Decommissioning<br />

Tobias Finken and Peter Hausch<br />

During decommissioning, both the functions and functional areas of the various building<br />

­sections must to some ­extent be considered in significantly different ways and extensively adapted<br />

to the requirements of the decommis sioning. Particularly in the case of nuclear facilities, the<br />

planning of which often did not take into account the eventual needs of decommissioning, there are<br />

great challenges in making such buildings as safe as technically possible during decommissioning.<br />

Importantly, these considerations must also cover the air distribution concepts of the building<br />

sections. The original ventilation must be shut down be<strong>for</strong>e the power plant is decommissioned<br />

because it is necessary to put in place a new air distribution concept that suits all necessary requirements<br />

<strong>for</strong> the period to follow.<br />

In order to design an air distribution<br />

concept <strong>for</strong> the decommissioning<br />

phase, a detailed inventory of both the<br />

previous ventilation system and all<br />

components of the building is necessary.<br />

Moreover, the decommissioning<br />

concept itself must be analysed in<br />

terms of what ventilation technology<br />

it may require. As the work pro­gresses,<br />

an increasingly large proportion of the<br />

total area is incorporated from the<br />

black zone into the white zone.<br />

Inside the white zone, which has<br />

already been made free of contaminants,<br />

the focus is on the building<br />

itself and on it lasting through the<br />

decommissioning phase, which may<br />

take several years. For the actual areas<br />

of operation, the focus is on the safety<br />

of the people working there and,<br />

on the buildings’, immediate surroundings.<br />

All areas of the system<br />

must be designed using a suitable<br />

control concept so that, on the one<br />

hand, safety is guaranteed at all times<br />

even if individual parts fail and, on the<br />

other hand, interdependent system<br />

components are automatically shut<br />

down in the event of partial failures.<br />

In addition to the applicable legal<br />

regulations, aspects of public perception<br />

and opinion must also be<br />

taken into account in nuclear engineering<br />

projects. Especially given the<br />

awareness that the building exhaust<br />

automatically means some interaction<br />

with the environment, there is a<br />

legitimate public interest that, in<br />

­addition to the best possible filtering<br />

of the air, there is also continuous<br />

monitoring so that any abnormality,<br />

even if it is within the legal limits, is<br />

noted and, in case of any doubt,<br />

triggers a safe shutdown of the system<br />

with a cause analysis to follow.<br />

Another aspect that distinguishes<br />

nuclear projects from conventional<br />

applications is the necessary period<br />

of observation. There can be a<br />

| Fig. 1.<br />

Schematic representation of the entire building.<br />

comparatively long period of time<br />

between the start of planning and its<br />

implementation.<br />

If further technological developments<br />

that can better increase safety<br />

take place within this time frame, these<br />

developments must be taken into<br />

consideration in the overall concept.<br />

As a last point to be considered, the<br />

costs associated with implementing<br />

air distribution systems must not be<br />

overlooked. For nuclear engineering<br />

projects, these generally take a back<br />

seat to safety aspects, but the public<br />

may have legitimate questions about<br />

the use of public funds, be it through<br />

direct expenditure or necessary<br />

subsidies.<br />

If one now takes a closer look at the<br />

air distribution requirements during<br />

decommissioning, there are two main<br />

areas of application:<br />

p The ventilation of the buildings<br />

themselves with their different<br />

areas (permanent negative pressure<br />

separation of areas via<br />

pressure cascades), and<br />

p Local work area-related ventilation<br />

with detection and filtering of<br />

radioactive particles and potentially<br />

contaminated dust particles<br />

as close as possible to the point of<br />

release.<br />

Planned entry <strong>for</strong><br />

In the following, both fields of application<br />

will be explained in more detail<br />

using specific examples.<br />

Figure 1 shows a schematic representation<br />

of a cross-section of a reactor<br />

building and auxiliary buildings with<br />

an air distribution concept suitable<br />

<strong>for</strong> decommissioning already implemented.<br />

Since the focus here is no<br />

longer on the com<strong>for</strong>t of the indoor<br />

climate, additional processing of the<br />

supply air can often be dispensed<br />

with. Instead, additional appropriately<br />

dimensioned inflow openings are<br />

added to the buildings. Here the air<br />

enters the building through simple<br />

­filter walls. Decen­tralized, mobile<br />

­HEPA filter de­vices come into use<br />

directly wherever the decommissioning<br />

or dismantling of old systems is taking<br />

place. In addition, an exhaust air<br />

system is connected from the outside in<br />

a separate con tainer building to ensure<br />

continuous negative pressure and<br />

­filtered air ­discharge from the building<br />

in order to prevent carryover of radioactive<br />

particles. The machine exhaust<br />

air shown in the ­figure is used to filter<br />

the total air inside the building.<br />

Figure 2 shows a top view of<br />

a newly constructed exhaust air<br />

container. This is a completely prefabricated<br />

module with a filter wall<br />

DECOMMISSIONING AND WASTE MANAGEMENT 403<br />

Decommissioning and Waste Management<br />

Ventilation Concepts <strong>for</strong> <strong>Nuclear</strong> Decommissioning ı Tobias Finken and Peter Hausch


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 404<br />

| Fig. 2.<br />

Exhaust air building (container).<br />

made of F9 + H13 filter elements and<br />

a fan that blows directly into a newly<br />

constructed exhaust air stack. In order<br />

to ensure that the building exhaust<br />

air released into the environment is<br />

actually safe, a fail-safe measurement<br />

system <strong>for</strong> identifying radioactive<br />

paricles in the air has been integrated<br />

into the exhaust air stack ( Figure 3).<br />

| Fig. 3.<br />

Measuring system <strong>for</strong> monitoring of the central building exhaust air.<br />

By installing pressure relief dampers,<br />

pressure cascades and differing<br />

areas of use can be put into effect at<br />

different phases of the decommissioning.<br />

For the safety of the people working<br />

in black zones, the most important<br />

point during decommissioning and<br />

dismantling is that the large amounts<br />

of dust present there accumulate<br />

safely in order to ensure exhaust air or<br />

recirculated air that is completely free<br />

of contamination.<br />

While particulate matter filters<br />

(HEPA filters) are able to filter even the<br />

smallest particles


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 5.<br />

Bag in/bag out technology on particulate removal and when changing the filter.<br />

| Fig. 7.<br />

Pressure curve of differential pressure across filters in the test with plasma cutting.<br />

­necessary filter changes, through<br />

its recleaning mechanism. Figure 5<br />

shows how the safe changing of<br />

the filter cells takes place using bag<br />

in/bag out technology. Here too, it is<br />

very important to consider unplanned<br />

operating situations and their associated<br />

risks. For systems that are<br />

connected to a central exhaust air<br />

system, there is the risk that negative<br />

pressure in the housing could develop<br />

due to a leak or a defective flap. The<br />

grooves into which the fastening<br />

bands are tensioned are there<strong>for</strong>e<br />

shaped in such a way that they<br />

can hold onto the protective bag<br />

even when it is sucked into the filter<br />

housing.<br />

For the optimal design of the system<br />

<strong>for</strong> online cleaning of the filter<br />

cells, it is first important to consider<br />

the buildup of dust deposits on the<br />

surface of the filter material. Figure 6<br />

shows schematically how such dust<br />

deposits are distributed during the<br />

cleaning process. It has been tested<br />

that the high working pressure of >5<br />

bar is necessary <strong>for</strong> the dust to be effectively<br />

removed from the tightly<br />

folded filter paper. Since there is<br />

always a small amount of dust in the<br />

| Fig. 6.<br />

Build-up on the filter material.<br />

filter cell, a ­special dust is applied as<br />

precoating when it is first put into operation<br />

and after a filter change. This<br />

<strong>for</strong>ms the base layer on the filter and<br />

enables more effective cleaning of the<br />

actual radioactive dust.<br />

To determine the most sensible<br />

parameters <strong>for</strong> a basic setting, investigations<br />

were carried out using plasma<br />

dust from a cutting process. The<br />

diagram in Figure 7 shows the<br />

­pressure curve of a recleanable filter<br />

stage over several hours of continuous<br />

operation.<br />

As can be clearly seen, the chosen<br />

triggering pressure difference of<br />

1500 Pa leads to a uni<strong>for</strong>m cleaning<br />

frequency and to a constant negative<br />

discharge pressure. With additional<br />

offline cleaning at the end of each<br />

assignment, the system can virtually<br />

return to its initial pressure loss of<br />

ca. 500 Pa. In this way, only the radioactive<br />

dust is collected, and the filter<br />

cell does not have to be replaced.<br />

To ensure safety, numerous parameters<br />

are continuously monitored.<br />

These include the filter load ­capacity<br />

and filter breakage of both stages. If a<br />

filter cell per<strong>for</strong>ates, the system immediately<br />

shuts down automatically.<br />

If the damaged cell is from the first<br />

stage, the automatic cleaning that<br />

usually occurs after the system stops is<br />

also prevented.<br />

For systems that are further<br />

connected to a central exhaust air system,<br />

it can also make sense to connect<br />

them to the central control room in<br />

order to test external release signals or<br />

to react automatically to error messages.<br />

Through the combination of<br />

centralized systems <strong>for</strong> building air<br />

distribution and local mobile systems<br />

in individual work areas <strong>for</strong> air conditioning<br />

or <strong>for</strong> filtering radioactive<br />

dusts close to the place of origin,<br />

overall safe operations of such air<br />

distribution systems can be achieved<br />

both <strong>for</strong> workers and a plant’s<br />

immediate surroundings. The special<br />

challenges that the decommissioning<br />

phase of a nuclear facility presents can<br />

thus likewise be met.<br />

Authors<br />

Tobias Finken<br />

Dr.-Ing. Peter Hausch<br />

Krantz GmbH<br />

Uersfeld 24<br />

52072 Aachen, Germany<br />

DECOMMISSIONING AND WASTE MANAGEMENT 405<br />

Decommissioning and Waste Management<br />

Ventilation Concepts <strong>for</strong> <strong>Nuclear</strong> Decommissioning ı Tobias Finken and Peter Hausch


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 406<br />

Planned entry <strong>for</strong><br />

Steam Generator Rip and Ship – a Valuable<br />

Contribution to Decommissioning and<br />

Dismantling of <strong>Nuclear</strong> <strong>Power</strong> Plants<br />

Heiko Herbell, Arne Larsson, Gregor Krause and Véronique Bouilly<br />

Practical examples are given from<br />

past projects such as Ringhals NPP<br />

(Sweden) or Stade NPP (Germany) but<br />

also from ongoing developments such<br />

as the Fessenheim decommissioning<br />

plan (France). These examples are<br />

used to understand major challenges<br />

during removal of the SGs and transportation<br />

e.g. shipment of the SGs.<br />

The radionuclide history and<br />

inventory is of major interest <strong>for</strong> the<br />

further treatment. Necessary input<br />

in<strong>for</strong>mation <strong>for</strong> transportation and<br />

offsite treatment of the SG is<br />

described.<br />

Concluding, the advantages are<br />

described <strong>for</strong> off-site treatment of<br />

large components.<br />

2 Description of rip & ship<br />

concept<br />

The rip & ship concept can be divided<br />

into the rip and the ship part. The<br />

rip part summarizes all activities<br />

necessary to remove the SG from its<br />

original position and place it outside<br />

the containment at ground level.<br />

Here, the ship part begins.<br />

a Removal of steam<br />

generators<br />

Fessenheim decommissioning plan is<br />

described in order to understand the<br />

challenges during removal of SG. This<br />

1 Introduction The removal and management of large components like steam generators<br />

(SGs) is one of the major tasks during decommissioning and dismantling of a PWR nuclear power<br />

plant. In case the decommissioning schedule is a key parameter to meet the budget, which it usually<br />

is, transportation and treatment of large components to an external facility (Rip & Ship) is<br />

­beneficial. This paper gives a short introduction of existing available technology <strong>for</strong> Rip & Ship<br />

combined with advanced treatment.<br />

example does not necessarily fits <strong>for</strong> all<br />

plant geometries and conditions but<br />

can be used to outline a specific<br />

sce nario <strong>for</strong> NPPs without the practical<br />

possibility to remove SGs as one<br />

piece.<br />

For the SG decommissioning EDF<br />

chose the following scenario:<br />

p Cutting of each steam generator in<br />

2 pieces (i.e. separation of steam<br />

dome) inside the reactor building<br />

as <strong>for</strong> the previously replaced<br />

steam generators).<br />

p It generates 6 pieces of less than<br />

200 t of which three are only<br />

slightly contaminated and three<br />

contains more than 99 % of the<br />

component radioactivity<br />

p Handling out of reactor building by<br />

a specific handling plat<strong>for</strong>m or<br />

polar crane,<br />

p Storage in a dedicated building on<br />

site (same than the one used <strong>for</strong><br />

steam generator replacement),<br />

p Treatment off site in a EDF Group<br />

facility aiming <strong>for</strong> recycling of<br />

material and minimization of the<br />

waste volume<br />

Prior to the removal of the SGs from<br />

the reactor hall, the previously<br />

replaced SGs will have to be eva cuated<br />

from the storage building and sent off<br />

site <strong>for</strong> treatment. This is scheduled to<br />

take place near term.<br />

The definition and implementation<br />

of optimized waste routes, be<strong>for</strong>e<br />

starting the dismantling, improves the<br />

project schedule, cost effectiveness<br />

and its success. This solution contributes<br />

to the:<br />

p reduction of interfaces and risks<br />

in decommissioning projects by<br />

inte gration across the value chain,<br />

p reduction of waste management &<br />

disposal costs,<br />

p optimization of scarce radioactive<br />

disposal capacity.<br />

For the rest of the French fleet, it’s not<br />

necessary to remove the steam dome<br />

from the steam generators to bring<br />

them out of the reactor building. Onepiece<br />

removal is the preference.<br />

These examples show that the rip<br />

& ship concept can be used <strong>for</strong> intact<br />

SGs or be combined with a partial<br />

segmentation of the SG in order to<br />

ease transportation within building<br />

structures.<br />

b Transportation of steam<br />

generators<br />

Any treatment outside the reactor<br />

building will likely, independent of<br />

distance, require a transport qualification<br />

& licensing as well as the<br />

involvement of a heavy lifting and<br />

transport company. The additional<br />

practical ef<strong>for</strong>ts <strong>for</strong> the licensee<br />

| Steam generator replacement upper part handling in the reactor building.<br />

| SG lower part handling in the reactor building.<br />

Decommissioning and Waste Management<br />

Steam Generator Rip and Ship – a Valuable Contribution to Decommissioning and Dismantling of <strong>Nuclear</strong> <strong>Power</strong> Plants ı Heiko Herbell, Arne Larsson, Gregor Krause and Véronique Bouilly


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 1.<br />

Ringhals SG entering the treatment facility (Cyclife Sweden AB).<br />

to remove the SGs from the site<br />

compared to an on-site waste<br />

treatment facility are fairly limited.<br />

However, the movement of large<br />

contaminated components over long,<br />

trans-boundary distances will require<br />

pre-studies and qualifications, special<br />

licenses, stakeholder interactions and<br />

practical transport arrangements.<br />

Cyclife has, both in practice and<br />

in studies, demonstrated that the by<br />

road or water <strong>for</strong> large contami nated<br />

components within Europe are open<br />

or can be opened when necessary.<br />

The same is expected also <strong>for</strong> other<br />

parts of the world, based on studies<br />

conducted.<br />

In total, hundreds of large, contaminated<br />

components have been<br />

shipped to the Swedish facility <strong>for</strong><br />

treatment. More than 20 of them have<br />

exceeded 300 tonnes in weight.<br />

3 Description of treatment<br />

technology<br />

Already in the early phases of the<br />

planning <strong>for</strong> a SG treatment project, it<br />

is important to understand the radionuclide<br />

inventory and distribution.<br />

Whether the SG has been chemically<br />

decontaminated as a part of a full<br />

­system decontamination, a specific<br />

object decontamination or only has<br />

decayed down to reasonable levels<br />

<strong>for</strong> handling and off-site treatment<br />

has a major impact on the optimisation<br />

of the treatment.<br />

A chemical decontamination prior<br />

to disconnection from the primary<br />

­circuit significantly decreases the<br />

collective dose exposure during dismantling,<br />

handling and treatment as<br />

well as <strong>for</strong> the return of the treatment<br />

residues (as most of the inventory was<br />

removed already at the NPP).<br />

Also detailed drawings, engineering<br />

details, metallurgic data,<br />

operational history records that are<br />

available are of great importance <strong>for</strong><br />

the handling, segmentation and decontamination<br />

activities and recycling.<br />

A key parameter is data on the level<br />

<strong>for</strong> contamination, if any at all of any<br />

significance, on the secondary side.<br />

Other, in most cases even more<br />

important parameters related to the<br />

operational history, are the data<br />

regarding the tubing: number and<br />

location of inserted sleeves and plugs.<br />

Also, how the sleeves and plugs are<br />

designed and have been fixed to the<br />

tubes are of prime interest.<br />

Cyclife Sweden, <strong>for</strong>merly Studsvik<br />

Waste Treatment, has per<strong>for</strong>med<br />

several SG treatment projects, in total<br />

treated 13 full size SGs.<br />

p Nine SGs of Westinghouse design<br />

originating from the Ringhals NPP<br />

in Sweden, approximately 300 Mg<br />

each.<br />

p Four SGs of Siemens design<br />

originating from the Stade NPP in<br />

Germany, approximately 165 Mg<br />

each.<br />

The conducted projects demonstrate<br />

that SGs can be authorized <strong>for</strong><br />

shipment and shipped as one piece,<br />

to an external treatment plant.<br />

The treatment process <strong>for</strong> recycling<br />

of full-size SG’s was started up in<br />

2005 and has been enhanced stepwise<br />

over the years. The driving <strong>for</strong>ce <strong>for</strong><br />

treatment of those components was<br />

primarily to reduce the waste volume<br />

<strong>for</strong> final disposal, although there are<br />

| Fig. 2.<br />

Cross section of 300 Mg SG of Westinghouse design.<br />

other benefits such as reduction of<br />

on-site/project related risks as well<br />

the elimination of a time critical<br />

project during on-going decommissioning.<br />

SGs received at the treatment<br />

facility are handled, further characterized,<br />

decontaminated and size<br />

reduced. The entire secondary<br />

side material do not usually require<br />

any special decontamination (but<br />

careful separation of non-contaminated<br />

parts) as the contamination on<br />

the secondary side should be very<br />

limited, if any. The parts exposed to<br />

the pri mary circuit, i.e. the water<br />

chamber, parts of the tube plate and<br />

the tubes, need special decontamination<br />

or material segregation to be<br />

candidates <strong>for</strong> clearance after treatment,<br />

if possible at all. The main<br />

challenge is to make the tubes candidate<br />

<strong>for</strong> clearance. It is currently<br />

not within reach, in practice, <strong>for</strong><br />

Inconel-600 tubes due to the intergranular<br />

cracks with embedded<br />

activity the treatment.<br />

Depending on the cost <strong>for</strong> waste<br />

storage and disposal, as well as the<br />

availability of repository volume, the<br />

plant owners may have different<br />

preferences. If the disposal cost is low,<br />

it may be sufficient to treat the entire<br />

secondary side <strong>for</strong> clearance and<br />

dispose the tubes, the tube plate and<br />

the water chamber as contaminated<br />

material. On the other hand, if the<br />

disposal cost is high, there is likely an<br />

interest to minimise the amount of<br />

waste also at higher treatment cost. In<br />

this case the preference may be that<br />

only the tubes and the residues from<br />

the treatment should be returned <strong>for</strong><br />

disposal.<br />

The secondary side material may, if<br />

con sidered free from contamination<br />

of any significance, be decontaminated<br />

and undergo clearance procedure<br />

<strong>for</strong> recycling back to industry<br />

either by implementing melting or by<br />

direct clearance. Both alternatives are<br />

available, the Cyclife preferred concept<br />

is to combine direct clearance<br />

and clearance after melting. In most<br />

DECOMMISSIONING AND WASTE MANAGEMENT 407<br />

Decommissioning and Waste Management<br />

Steam Generator Rip and Ship – a Valuable Contribution to Decommissioning and Dismantling of <strong>Nuclear</strong> <strong>Power</strong> Plants ı Heiko Herbell, Arne Larsson, Gregor Krause and Véronique Bouilly


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 4<strong>08</strong><br />

| Fig. 3.<br />

Treatment of Ringhals SG in the Cyclife Swedish facility.<br />

cases a vast majority of the secondary<br />

side tonnage goes to direct clearance.<br />

By applying melting on all the metal,<br />

the results are more predictable but at<br />

a somewhat higher cost.<br />

The primary side parts will likely<br />

have to be mechanically decontaminated<br />

by a physical removal of the<br />

surface layer to be candidates <strong>for</strong><br />

clearance after treatment. This has<br />

to be optimized with the customers<br />

needs. Based on the past experiences,<br />

the decontamination methods have<br />

been enhanced incorporating new<br />

technologies and new equipment.<br />

The tube plates need a careful<br />

analysis prior to selection of the<br />

treatment approach. Multiple technologies<br />

are applied by the SG<br />

­manufacturers to fix the tubes in the<br />

tube plate and to secure that no<br />

leakage will occur. Of these reasons,<br />

the treatment of the tube plates will<br />

have to be tailored. Depending on<br />

the complexity the ef<strong>for</strong>ts will differ<br />

to make the tube plate a candidate<br />

<strong>for</strong> clearance.<br />

By application of chemical decontamination<br />

prior to SG removal,<br />

the dose rates <strong>for</strong> handling and<br />

­treatment can be reduced significantly<br />

and the management of the residual<br />

waste simplified. The residual waste<br />

contains most of the radioactive<br />

­inventory but in a significantly lower<br />

volume.<br />

The SG recycling rate is expected<br />

to be in the order of<br />

60-80 % depending on the degree<br />

of treatment aiming <strong>for</strong> clearance,<br />

i.e. 20-40 % of the original tonnage<br />

will have to be disposed as radioactive<br />

waste.<br />

The residues from the treatment<br />

will have to be returned as radioactive<br />

waste <strong>for</strong> disposal. Figure 4 shows<br />

how the secondary waste from treatment<br />

will look like.<br />

Likely all tubes will have to be<br />

returned as waste. Depending on<br />

customer preferences, the tube<br />

material can be compacted, chopped<br />

and flattened or melted as illustrated<br />

in Figure 5.<br />

The treatment time <strong>for</strong> one,<br />

300 Mg steam generator, using the<br />

latest technology is estimated to<br />

3-5 months (depending on the<br />

selected scope of services and the<br />

­specific SG properties).<br />

In a volume perspective, which is<br />

the most important as most waste<br />

repositories charge per volume<br />

­disposed, only 7-20 % of the original<br />

volume <strong>for</strong> a full size SG will have to<br />

be disposed. The percentage depends<br />

both on the degree of treatment <strong>for</strong><br />

clearance but also on the ef<strong>for</strong>ts to<br />

­reduce the final volume.<br />

4 Conclusion<br />

Examples from past projects such<br />

as Ringhals NPP or Stade NPP but<br />

also from ongoing projects such as<br />

Fessenheim underline flexibility of<br />

the rip & ship concept in order<br />

to overcome challenges during<br />

on-site treatment as eq logictic <strong>for</strong><br />

>1000 tonnes of SG material during<br />

treatment in the narrow reactor<br />

building.<br />

The Fessenheim projects also<br />

highlights the issue with previously<br />

replaced SGs stored on site. Removal<br />

of them from the storage is an<br />

important decommissioning preparation<br />

activity to be done well<br />

ahead of the dismantling start.<br />

Transportation of large contaminated<br />

components over long, transboundary<br />

distances require prestudies<br />

and qualifications, special<br />

licenses, stakeholder interactions and<br />

practical packaging arrangements.<br />

Besides these ef<strong>for</strong>ts, offsite treatment<br />

of large components enable<br />

­significant advantages such as:<br />

p Less waste handling and treatment<br />

on site.<br />

p Potential <strong>for</strong> a significant reduction<br />

of the decommissioning schedule.<br />

p Risk mitigation by transfer of work<br />

to specialists with proven pro cesses<br />

and experienced work<strong>for</strong>ce in facilities<br />

that have been designed<br />

and built <strong>for</strong> the purpose.<br />

p A significantly lower volume of<br />

waste <strong>for</strong> disposal. In addition, this<br />

option has the benefit of recycling.<br />

p A predictable result and a fixed<br />

cost.<br />

Abbreviations<br />

EDF – Électricité de France<br />

NPP – <strong>Nuclear</strong> <strong>Power</strong> Plant<br />

PWR – Pressurized Water Reactor<br />

SG – Steam Generator<br />

SGR – Steam Generator Replacement<br />

| Fig. 4.<br />

Secondary waste from treatment.<br />

Authors<br />

Dr. Heiko Herbell<br />

heiko.herbell@framatome.com<br />

Framatome GmbH<br />

Paul Gossen Str. 100<br />

91058 Erlangen, Germany<br />

Arne Larsson<br />

Gregor Krause<br />

Cyclife Sweden AB<br />

Box 610<br />

SE-61110 Nyköping, Sweden<br />

Véronique Bouilly<br />

Cyclife Engineering<br />

196, Avenue Thiers<br />

69006 Lyon, France<br />

| Fig. 5.<br />

Alternatives <strong>for</strong> the tube material not subject to clearance.<br />

Decommissioning and Waste Management<br />

Steam Generator Rip and Ship – a Valuable Contribution to Decommissioning and Dismantling of <strong>Nuclear</strong> <strong>Power</strong> Plants ı Heiko Herbell, Arne Larsson, Gregor Krause and Véronique Bouilly


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Casks and Cask Stacks in Interim Storage<br />

Facilities under Earthquake Loads<br />

Nina Wieczorek<br />

Interim storage of casks In Germany, there are interim storage facilities built directly Planned entry <strong>for</strong><br />

next to nuclear ­power plants as well as three central interim storage facilities sited in Ahaus,<br />

­Gorleben and near Lubmin. These interim storage facilities differ in the classification of casks<br />

stored, differentiating between radioactive waste with negligible heat generation and dry cask<br />

storage of spent fuel and heat-generating waste. Depending on the type of waste, different types of<br />

casks are stored. Besides containers of various types, cylindrical casks like e.g. shipping and storage casks of type<br />

­CASTOR® V/19, V/52, MTR3 and cast iron casks of type MOSAIK® are used. The containers and casks are stored<br />

separately as well as stacked, whereat common ISO edges are used <strong>for</strong> containers. The casks are stacked directly or with<br />

stacking aids. This paper deals with cylindrical casks stored separately as well as stacked.<br />

Guidelines and standards<br />

ESK guidelines<br />

The ESK guidelines (ESK: <strong>Nuclear</strong><br />

Waste Management Commission)<br />

contain protection goals with which<br />

the storage of radioactive waste has to<br />

comply. The requirements resulting<br />

from these are the avoidance of<br />

any unnecessary radiation exposure<br />

or contamination of man and the<br />

environment and the minimization of<br />

any radiation exposure or contamination<br />

of man and the environment.<br />

Depending on the type of radioactive<br />

waste, the ESK established<br />

guidelines to ensure the compliance<br />

with the protection goals, see [1, 2],<br />

whereat external hazards like the<br />

“safety shutdown earthquake” (SSE)<br />

focussed in this paper is addressed as<br />

well. Ensuring the compliance with<br />

the protection goals <strong>for</strong> the load case<br />

SSE is feasible by proof of stability of<br />

the containers and casks.<br />

KTA series 2201<br />

The KTA series consists of six parts<br />

and deals with the design of nuclear<br />

power plants against seismic events,<br />

whereat its scope of application is<br />

extended by the ESK guidelines<br />

[1, 2]. KTA 2201.1 [3] contains<br />

general requirements concerning the<br />

design basis earthquake (DBE) and<br />

the ­verification. KTA 2201.2 [4] deals<br />

with the subsoil. While KTA 2201.3<br />

[5] contains detailed requirements<br />

concerning the verification of civil<br />

structures, KTA 2201.4 [6] contains<br />

detailed requirements concerning the<br />

verification of components.<br />

All six parts had been amended<br />

successively between 2011 to 2015.<br />

Especially with the new version of<br />

KTA 2201.1 [3], the requirements<br />

regarding the DBE were augmented<br />

compared to the previous version.<br />

Thus, the DBE has to be specified with<br />

an intensity of at least VI and a probability<br />

of exceedance of at least 10 -5 /a.<br />

External hazard – “safety<br />

shutdown earthquake”<br />

Concerning the existing interim<br />

storage facilities, seismological expert's<br />

reports define a site-specific<br />

DBE and seismo-engineering parameters<br />

(e.g. , intensity, probability of<br />

exceedance and strong motion<br />

­duration) on the basis of KTA 2201.1<br />

[3] as well as soil parameters (e.g.<br />

shear modulus, density, Poisson's<br />

­ratio, damping ratio). According to<br />

KTA 2201.1 [3], the DBE is defined<br />

as a free field response spectrum,<br />

meaning a ground acceleration<br />

response spectrum <strong>for</strong> a reference<br />

horizon in the subsoil, where the<br />

­oscillation properties are not influenced<br />

by building structures. No<br />

soil-structure interaction is considered.<br />

Generally, the reference<br />

horizon is equal to the ground level or<br />

the geological layer boundary of a<br />

­sufficiently stiff ground layer. The<br />

scatter band of the soil profile as well<br />

as uncertainties are covered within<br />

the computations according to KTA<br />

2201.1 [3]. The DBE results from<br />

smoothed, broadened and enveloped<br />

spectra and is defined <strong>for</strong> the horizontal<br />

resultant as well as the<br />

horizontal and vertical component.<br />

Figure 1 (a) contains exemplarily free<br />

field ­response spectra <strong>for</strong> a damping<br />

ratio of 5 % <strong>for</strong> an arbitrary site in<br />

Germany. The peak ground acceleration<br />

(pga) <strong>for</strong> the horizontally resultant<br />

direction a hr is 0.26 g, <strong>for</strong><br />

the vertical component a v it is 50 % of<br />

a hr (0.13 g).<br />

Floor response spectra applied<br />

as design spectra<br />

KTA 2201.4 [6] defines the design<br />

spectrum to be enveloping, widened<br />

DECOMMISSIONING AND WASTE MANAGEMENT 4<strong>09</strong><br />

a<br />

| Fig. 1.<br />

Free field response spectra <strong>for</strong> a site in Germany (a) and FEM model of an interim storage facility (b).<br />

b<br />

Decommissioning and Waste Management<br />

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 410<br />

and smoothed. One differentiates<br />

between ground acceleration spectrum<br />

as primary spectrum, building<br />

response spectrum as secondary<br />

spectrum, and component response<br />

spectrum as tertiary spectrum.<br />

a<br />

Generally, based on the sitespecific<br />

DBE, floor response spectra<br />

(FRS) <strong>for</strong> the base plate and the<br />

crane runway are computed on a FEM<br />

model of the building of the interim<br />

storage facility, exemplarily shown in<br />

Figure 1 (b). In order to consider<br />

soil-structure interaction, besides the<br />

structural building the soil is contained<br />

in the numerical model as well.<br />

According to KTA 2201.1 [3] and<br />

2201.3 [5], diverse variations of the<br />

| Fig. 2.<br />

Floor response spectra of the base plate of an interim storage facility <strong>for</strong> a damping ratio of 4 %; x-direction (a), y- direction (b) und z- direction (c); comparison of the required response<br />

spectra and the response spectra trans<strong>for</strong>med back from the compatible time histories (TH, two per direction).<br />

b<br />

c<br />

a<br />

| Fig. 3.<br />

Comparison of time histories from structural analyses (SA; blue line) with time histories compatible with the required response spectra (orange line); exemplarily plotted <strong>for</strong> the model<br />

„ maximum shear modulus (stiff soil) / maximum load“; x-direction (a), y- direction (b) and z- direction (c).<br />

b<br />

c<br />

a<br />

b<br />

| Fig. 4.<br />

FEM model of a cask of type CASTOR® V/52 (a), stacked casks of type MOSAIK® without stacking aids (b), stacked casks of type CASTOR® MTR3<br />

with stacking aids (c), stacked casks of type CASTOR® MTR3 with stacking aids (section of whole model) (d); (a) to (c) with base plate section, respectively (green).<br />

c<br />

d<br />

Decommissioning and Waste Management<br />

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

reference model are investigated considering<br />

the scatter band of the soil<br />

profile and the various loading conditions.<br />

Concerning the soil, the<br />

average shear modulus (G mid ) is<br />

varied between a lower (G min ) and<br />

upper (G max ) bound. The various<br />

loading conditions are considered by<br />

additional masses ranging from a<br />

lower (M min ) and upper bound (M max )<br />

as well.<br />

Response spectra are computed<br />

<strong>for</strong> the various models, whereat the<br />

excitation is applied as time histories<br />

compatible with the DBE according<br />

to KTA 2201.1 [3] and 2201.3 [5].<br />

­Afterwards, FRS are identified as<br />

smoothed, broadened and enveloping<br />

design response spectra (DRS).<br />

Figure 2 contains DRS with a<br />

damping ratio of 4 % <strong>for</strong> the base<br />

plate of an interim storage facility.<br />

Furthermore, Figure 2 contains<br />

response spectra trans<strong>for</strong>med back<br />

from time histories, which are<br />

compatible with the DRS according to<br />

KTA 2201.1 [3] and 2201.3 [5],<br />

and taken as a basis <strong>for</strong> component<br />

verification.<br />

Time histories compatible<br />

with required/design response<br />

spectrum<br />

If quasi-static methods are not<br />

­sufficient, time history analyses are<br />

per<strong>for</strong>med in order to verify the<br />

­stability of the cask/cask stack <strong>for</strong> the<br />

load case SSE with the help of a<br />

FEM model of the cask/cask stack.<br />

As ­mentioned above, time histories<br />

compatible with the required response<br />

spectrum (RRS) are generated<br />

complying with the requirements of<br />

KTA 220.1 [3] and used as excitation<br />

<strong>for</strong> the FEM model, see Figure 2.<br />

Time histories taken directly<br />

from structural analysis<br />

Besides the approach of time histories<br />

compatible with the RRS, KTA 2201.1<br />

[3] allows an alternative approach by<br />

applying time histories taken directly<br />

from the structural analyses of the<br />

building. Here, it has to be considered<br />

that the scatter band of the structural<br />

analyses is covered. That is, analogue<br />

to the structural analyses, the regard<br />

of variation of soil stiffness (G min ,<br />

G max ) and loading conditions (M min ,<br />

M max ). Since the time histories taken<br />

from the structural analyses do not<br />

contain the conservatives from the<br />

RRS, this approach allows the application<br />

of lower excitations to the FEM<br />

model, but results in an increase of<br />

computational cost, see Figure 3.<br />

FEM model<br />

Modelling and computations are<br />

per<strong>for</strong>med with the commercial<br />

FEM program ANSYS including the<br />

­LS-DYNA solver. In order to represent<br />

relative displacements of the cask/<br />

cask stack on the base plate, including<br />

tilting, trundling and sliding, besides<br />

the cask/cask stack, the numerical<br />

model also contains a section of the<br />

base plate. Between the (bottom) cask<br />

and the base plate and between the<br />

casks in a stack contact definitions are<br />

applied to the FEM model. For the<br />

contact pairs the friction coefficient<br />

are varied between a lower and an upper<br />

bound depending on the material<br />

(µ cask-base, min = 0.2 und µ cask-base, max =<br />

0.6, µ cask-cask/stacking aid, min = 0.1 and<br />

µ cask-cask/stacking aid, max = 0.3). The<br />

lower bound allows a conservative<br />

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Decommissioning and Waste Management<br />

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 412<br />

a b c<br />

| Fig. 5.<br />

Displacements of a cask of type CASTOR® V/52 relative to the base plate applying various friction coefficients <strong>for</strong> the contact pair cask/base plate; excitation using time histories<br />

compatible with the RRS in Figure 2; horizontal resultant (a) and vertical (b) displacements applying a friction coefficient of µ = 0.4 and horizontal resultant displacements (c) applying<br />

a friction coefficient of µ = 0,2.<br />

a<br />

| Fig. 6.<br />

Cask stack of four casks of type MOSAIK®; horizontally resultant displacements of center of cover plate of the bottom cask (cask no. 1) and top cask (cask no. 4), respectively; excitation using<br />

two time history combinations compatible with the RRS; friction coefficient of µ = 0.6 <strong>for</strong> the contact pair cask/base plate and µ = 0.2 <strong>for</strong> the contact pair cask/cask (a); friction coefficient of<br />

µ = 0.2 <strong>for</strong> the contact pair cask/base plate and µ = 0.1 <strong>for</strong> the contact pair cask/cask (b).<br />

b<br />

approach concerning sliding, while<br />

the upper bound allows a conser vative<br />

approach concerning tilting and<br />

trundling. The excitation within the<br />

nonlinear computations is applied to<br />

the base plate section. Figure 4 contains<br />

FEM models of various casks and<br />

cask stacks of the types CASTOR®<br />

V/52, MTR3 and MOSAIK®, with and<br />

without stacking aids.<br />

Nonlinear time history analyses<br />

For nonlinear time history analyses,<br />

loading conditions of the cask and<br />

within the cask stack are investigated.<br />

Furthermore, <strong>for</strong> cask stacks without<br />

stacking aids imperfections due to<br />

handling accuracy need to be considered.<br />

Depending on the computed<br />

­displacements of the cask/cask stack<br />

relative to the base plate section,<br />

minimum distances can be necessary<br />

in order to avoid collisions of neighbouring<br />

casks <strong>for</strong> the load case SSE.<br />

Results – time histories<br />

compatible with design spectra<br />

Figure 5 contains displacements of a<br />

single cask of type CASTOR® V/52<br />

relative to the base plate section with<br />

an average (µ = 0.4) and a minimum<br />

(µ = 0.2) friction coefficient <strong>for</strong><br />

the contact pair cask/base plate.<br />

­Applying an average friction coefficient<br />

leads to trundling movements<br />

with a maximum horizontal displacement<br />

of ­approx. 70 mm <strong>for</strong> the cover<br />

plate (Figure 5 (a)) and a maximum<br />

vertical displacement of approx.<br />

25 mm <strong>for</strong> the bottom plate with a<br />

phase shift visible by the displacements<br />

of two ortho gonal result<br />

nodes on the bottom ( Figure 5 (b)).<br />

Applying the lower bound of the<br />

­friction ­coefficient leads to a sliding<br />

movement with a com paratively little<br />

maximum horizontal displacement of<br />

approx. 6 mm.<br />

One can see clearly larger horizontal<br />

displacements in Figure 6,<br />

whereat these displacements come<br />

from a cask stack of four casks of type<br />

MOSAIK® without stacking aids.<br />

Exemplarily, two different time<br />

history combinations are contrasted,<br />

respectively, <strong>for</strong> each limit value<br />

investigation of the friction coefficients.<br />

While the influence of the<br />

time history combination on the<br />

horizontal stack displacement is<br />

quiet low concerning the lower<br />

bounds of the friction coefficients (TH<br />

comb. 1 approx. 33 mm vs TH comb. 2<br />

approx. 30 mm), concerning the ­upper<br />

bound it is significant (TH comb. 1<br />

­approx. 250 mm vs TH comb. 2 ­approx.<br />

430 mm). Like the single cask<br />

shown in Figure 5, the whole stack<br />

trundles. As one can see in ­Figure<br />

7 (c), the relative displacements<br />

<strong>for</strong> the horizontally resultant direction<br />

of two ortho gonal nodes, respectively,<br />

are in-phase. The cask no. 4 (top<br />

cask) slides less than 1 mm on the<br />

cask no. 3.<br />

Decommissioning and Waste Management<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

a<br />

b<br />

| Fig. 7.<br />

Cask stack of four casks of type MOSAIK®; excitation using time histories compatible with the RRS; friction coefficient of µ = 0.6 <strong>for</strong> the contact pair cask/base plate and friction coefficient<br />

of µ = 0.3 <strong>for</strong> the contact pair cask/cask; horizontal (a) and vertical (b) displacement of the bottom cask (no. 1) relative to the base plate; horizontal displacements of the top cask (no. 4) and<br />

relative to the cask no. 3 (c).<br />

c<br />

DECOMMISSIONING AND WASTE MANAGEMENT 413<br />

a<br />

b<br />

c<br />

| Fig. 8.<br />

Cask stack of four casks of type MOSAIK®; excitation using time histories taken directly from the structural analyses; friction coefficient of µ = 0.6 <strong>for</strong> the contact pair cask/base plate and<br />

friction coefficient of µ = 0.3 <strong>for</strong> the contact pair cask/cask; horizontal (a) and vertical (b) displacement of the bottom cask (no. 1) relative to the base plate; horizontal displacements of the top<br />

cask (no. 4) and relative to the cask no. 3 (c).<br />

Results – approach of times<br />

histories taken directly<br />

from structural analyses<br />

Since the displacements of the<br />

cask stack of four casks of type<br />

­MOSAIK® computed applying time<br />

histories compatible with the DRS<br />

and the upper bounds of the friction<br />

coefficients are very large, in further<br />

computations time histories taken<br />

directly from the structural analyses<br />

of the building are applied to the<br />

FEM model complying with the<br />

­requirements of KTA 2201.1 [3] and<br />

2201.4 [6] as well. Figure 8 gives a<br />

similar image like Figure 7, the cask<br />

stack trundles, but the displacements<br />

are con siderably lower. For the top<br />

of the cask stack, the maximum<br />

horizontal displacement is approx.<br />

220 mm. Even though the computational<br />

cost increases, the maximum<br />

displacements and thereby<br />

the necessary minimum distance<br />

of two neigh bouring casks/cask<br />

stacks can be reduced in order to<br />

avoid a collision.<br />

Conclusions<br />

In order to proof stability of a cask/<br />

cask stack in interim storage facilities<br />

<strong>for</strong> the load case “safety shutdown<br />

earthquake” tilting, trundling and<br />

sliding of the cask/cask stack need<br />

to be investigated. With the help of<br />

FEM, various computations on a<br />

suitable numerical model containing<br />

the cask, stacking aids if utilised<br />

and a section of the base plate can<br />

be carried out applying friction<br />

­coefficients <strong>for</strong> both contact partners,<br />

cask/base plate and cask/cask, within<br />

a certain range and varying the mass<br />

distribution of the cask stack. In<br />

doing so, the numerical results show<br />

that not the sliding of the cask/<br />

cask stack is significant but the<br />

­trundling leading to a predefinition<br />

of a minimum distance between<br />

neighbouring casks/cask stacks.<br />

Literature<br />

[1] ESK – <strong>Nuclear</strong> Waste Management Commission:<br />

Recommendation of the <strong>Nuclear</strong> Waste Management<br />

Commission: Guidelines <strong>for</strong> dry cask storage of spent fuel and<br />

heat- generating waste; Revised version of 10.06.2013<br />

[2] ESK – <strong>Nuclear</strong> Waste Management Commission:<br />

Recommendation of the <strong>Nuclear</strong> Waste Management<br />

Commission: Guidelines <strong>for</strong> the storage of radioactive waste<br />

with negligible heat generation; Revised version of<br />

10.06.2013<br />

[3] KTA Rule 2201.1 (11-2011): Design of <strong>Nuclear</strong> <strong>Power</strong> Plants<br />

against Seismic Events. Part 1: Principles; Version November<br />

2011<br />

[4] KTA Rule 2201.2 (11-2012): Design of <strong>Nuclear</strong> <strong>Power</strong> Plants<br />

against Seismic Events. Part 2: Subsoil; Version November<br />

2012<br />

[5] KTA Rule 2201.3 (11-2013): Design of <strong>Nuclear</strong> <strong>Power</strong> Plants<br />

against Seismic Events. Part 3: Design of structural<br />

components (civil structures); Version November 2013<br />

[6] KTA Rule 2201.4 (11-2012): Design of <strong>Nuclear</strong> <strong>Power</strong> Plants<br />

against Seismic Events. Part 4: Components; Version<br />

November 2012<br />

Author<br />

Dr.-Ing. Nina Wieczorek<br />

n.wieczorek@gmx.de<br />

Technical Advisor<br />

“Structural Dynamics and<br />

Earthquake Engineering”<br />

TÜV NORD EnSys GmbH & Co. KG<br />

Bahnstr. 31<br />

22525 Hamburg, Germany<br />

Decommissioning and Waste Management<br />

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


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DECOMMISSIONING AND WASTE MANAGEMENT 414<br />

Radioactivity Calculation of the<br />

Concrete Shielding of the Petten LFR<br />

and the Dodewaard BWR<br />

Lino Salamon, Perry Young and Lojze Gačnik<br />

Planned entry <strong>for</strong><br />

1 Introduction After nearly 30 years of operation the <strong>Nuclear</strong> <strong>Power</strong> Plant Dodewaard<br />

(KCD) was ­permanently shut-down in March 1997. The KCD was a General Electric design boiling<br />

water reactor (BWR) with thermal capacity 183 MWt, and an electrical capacity of 54 MWe. After<br />

shutdown data was collected on the radioactive inventory in preparation <strong>for</strong> Safe Enclosure and future<br />

dismantlement. One of the larger reactor components undergoing significant neutron ­activation<br />

is the concrete biological shield (BS). Gamma spectroscopy measurements were made with samples of the concrete BS to<br />

know the mass specific activity of the nuclides at different positions in the BS and to estimate the amount of radioactive<br />

waste that would need to be sent to the Dutch disposal company COVRA at different dismantlement dates [1].<br />

The Low Flux Reactor (LFR) at the<br />

NRG Petten site was permanently<br />

shut-down in November 2010 and<br />

decommissioning was completed<br />

­February 2019. The LFR is a JASON<br />

variant of the Argonaut class reactor<br />

supplied by Hawker Siddeley, with<br />

a thermal capacity of 30 kWt.<br />

Similarly to KCD, gamma spectroscopy<br />

measure ments were made to<br />

assess the radioactivity of the LFR’s<br />

concrete shielding [2].<br />

The aim of this work was to predict<br />

the specific BS activities of KCD and<br />

LFR via activation calculations and to<br />

determine the delimitation between<br />

regions of free-release material and<br />

radioactive material that must be sent<br />

to COVRA. The amount of radioactive<br />

waste estimated from the measurements<br />

was compared to the calculated<br />

values. The ability to predict the<br />

extent of radioactivity in the concrete<br />

via calculation should permit better<br />

planning of a nuclear facility’s decommissioning<br />

and could possibly reduce<br />

costs.<br />

2 Calculation Methodology<br />

The calculation of the activation<br />

inventory was done in two steps. First,<br />

the neutron fluxes and their spectra<br />

were calculated in the BS using the<br />

Monte Carlo radiation-transport code<br />

MCNP6.2 [3]. In the second step, the<br />

flux tallies were used in a FISPACT-II<br />

[4] activation calculation. Based on<br />

the free release activity limits [5] the<br />

concretes were then categorized as<br />

radioactive or free-release and<br />

hence the amount of radioactive<br />

waste were determined <strong>for</strong> different<br />

dismantlement dates. Below in section<br />

2.1 is detailed the methodology<br />

<strong>for</strong> Dodewaard (KCD). The methodology<br />

<strong>for</strong> the LFR can be found in Ref.<br />

[2]; they are broadly similar, although<br />

the MCNP geometric model of the LFR<br />

is of far greater detail, explicitly modelling<br />

almost all of the features in the<br />

reactor and shielding.<br />

2.1 KCD Geometric Model<br />

The MCNP geometry of Dodewaard<br />

from [1] was used to calculate the<br />

neutron fluxes in BS, which is a<br />

circular wall around the containment<br />

wall. Two separated geometrical<br />

models were available: a model of<br />

the lower part extending between the<br />

bottom of the concrete floor up to the<br />

top of the core and a model of the<br />

upper part extending from slightly<br />

below the reactor core up to the top of<br />

the biological shield above the reactor<br />

vessel. The model of the Dodewaard’s<br />

lower part is only a quarter of<br />

the whole geometry, with reflective<br />

surfaces at 0 and 90. The lower<br />

and upper model are visualized in<br />

Figure 1. Both models were updated<br />

to ENDF/BVIII.0 nuclear data libraries<br />

<strong>for</strong> the neutron transport using<br />

temperatures broadly equivalent to<br />

those during operation. These models<br />

are very simple and lack much definition.<br />

To wit, the nuclear fuel region is<br />

a smeared homogenous zone, mixing:<br />

fuel, cladding and water. For the<br />

indicative purposes of this work, they<br />

should be adequate.<br />

| Fig. 1.<br />

The MCNP models of Dodewaard’s lower (left) and upper part (right).<br />

2.1 Material Composition<br />

There are two types of concrete in the<br />

BS of Dodewaard: heavy concrete with<br />

density ρ ≈ 3.5 g/cm 3 (most of the inner<br />

shielding and some of the BS) and<br />

light concrete with density ρ ≈ 2.3 g/<br />

cm 3 (most of the outer part of BS).<br />

The material composition of the<br />

heavy and light concrete was updated<br />

Decommissioning and Waste Management<br />

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with the composition obtained from<br />

the X-Ray Fluorescence (XRF),<br />

­ICP-AES, ICP-MS, Loss of Ignition<br />

(LOI), Total Carbon (TC) and Total Organic<br />

Carbon (TOC) measurements<br />

carried out by TCKI and TNO-Utrecht.<br />

Measurements were per<strong>for</strong>med using<br />

10 free-release heavy concrete and<br />

light concrete samples obtained from<br />

Dodewaard. These had been part of<br />

the previous core extractions [1], and<br />

were being held on site within a<br />

cabinet.<br />

In the calculations, the XRF<br />

measurements were used <strong>for</strong> the mass<br />

fractions of the major constituent<br />

elements and ICP <strong>for</strong> the trace elements<br />

(Eu, Mn and Co). The LOI, TC and<br />

TOC were used together to deduce the<br />

moderating media (H, C and O). Based<br />

on the experience comparing the LFR<br />

gamma measurements to calculations<br />

in [2], two cases were defined: a Best<br />

Estimate case (BE) and a Conservative<br />

case (CON). In these cases, per<br />

concrete type, the mass fractions used<br />

<strong>for</strong> the elements were as follows:<br />

BE:<br />

p MCNP: maximum values of the<br />

measurements <strong>for</strong> hydrogen and<br />

carbon, mean values of the<br />

measure ments <strong>for</strong> the rest<br />

p FISPACT: median values of the<br />

measurements<br />

CON:<br />

p MCNP: mean values of the<br />

measurements<br />

p FISPACT: maximum values of the<br />

measurements<br />

2.2 Variance Reduction<br />

Due to the deep penetration nature of<br />

the problem, some kind of variance<br />

reduction technique had to be applied<br />

to efficiently propagate neutrons in<br />

the concretes. For this reason, the<br />

­ADVANTG code [6] was used to create<br />

space and energy dependent mesh<br />

based weight windows using threedimensional<br />

(3D) discrete ordinates<br />

(SN) solutions of the direct and<br />

adjoint deterministic transport equations<br />

that are calculated by the<br />

­Denovo package. An MCNP geometry<br />

of the upper part overlaid with the<br />

weight-window mesh is presented in<br />

Figure 2.<br />

2.3 MCNP Flux Calculations<br />

MCNP calculations were done in<br />

multiple steps. First, a k-code criticality<br />

calculation was per<strong>for</strong>med to<br />

generate a volumetric fission source<br />

(SSW) in the nuclear fuel. This source,<br />

together with the weight-windows<br />

produced by ADVANTG, was subsequently<br />

used in a fixed source<br />

| Fig. 2.<br />

Upper part of the Dodewaard geometry model<br />

overlaid with a weight-windows spatial mesh.<br />

calculation. The per-source-particle<br />

neutron flux (neutrons/cm 2 ) in 172<br />

energy groups was tallied in small<br />

voxels over the concrete regions using<br />

Cartesian and cylindrical mesh geometry.<br />

2.4 FISPACT Activation<br />

Calculations<br />

Using the obtained flux tallies, the specific<br />

activities (in Bq/g) of concrete<br />

were estimated with FISPACT. In ­order<br />

to get the actual neutron fluxes, the<br />

calculated tallies first need to be scaled<br />

with a scaling factor PNF as ­defined in<br />

[7]. For this scaling, Q= 193 MeV, ¯v f =<br />

2.565 and k eff = 0.98902 (the last two<br />

calculated in MCNP), were used.<br />

­Finally, the scaled fluxes were multiplied<br />

in each time period by the power<br />

history of KCD [8] in that time<br />

period. Specific ­activities in FISPACT<br />

were ­calculated with the 7<strong>09</strong>-energy<br />

group activation library ENDF/BVIII.0<br />

( except <strong>for</strong> Eu-151 and Eu-153 that<br />

used the ­EAF-2010 library, following<br />

Eu C/M results <strong>for</strong> the different<br />

datasets found in [2]).<br />

Release<br />

date<br />

Ref.<br />

[1]<br />

2.5 Radioactive and<br />

Free-Release Zones<br />

Radioactivity of the material is<br />

­defined according to standards<br />

­defined in [5]. Each nuclide i has its<br />

own specific activity limit, I i , that it is<br />

not allowed to exceed in order to<br />

qualify as free release. If a sample<br />

has multiple active nuclides, then an<br />

I 0 ratio is defined as:<br />

where a i is the specific activity of<br />

nuclide i. For I 0 < 1 the sample is<br />

considered to be free release and <strong>for</strong><br />

I 0 ≥ 1 the sample is considered as<br />

radioactive. The I 0 criteria was used to<br />

map the radioactive and free-release<br />

regions of Dodewaard KCD. Finally,<br />

the total mass of concrete voxels<br />

with I 0 ≥ 1 was used to estimate the<br />

radioactive waste at different dismantlement<br />

dates.<br />

3 Results<br />

3.1 Dodewaard<br />

Figure 3 presents the radioactive<br />

(in red) and free-release (in green)<br />

regions at different axial positions of<br />

the Dodewaard model (indicated in<br />

Figure 1), approximately 43 years<br />

after the KCD shutdown. Colored<br />

areas represent the heavy and light<br />

concrete of the biological shield, but<br />

also concretes inside the containment<br />

wall. The largest radial spread of<br />

radioactivity in BS is at the heights<br />

surrounding the active core (at<br />

z ∼ 2300 cm) and it (at z ∼ 2800 cm),<br />

where the BS is mainly made of light<br />

concrete.<br />

Figure 4 compares the radioactivity<br />

of the BS at z = 2750 cm <strong>for</strong><br />

two time steps: a reactor shut-down<br />

(26/03/1997) and the planned end of<br />

Safe Enclosure (01/01/2040). The<br />

radioactivity in the radial direction is<br />

significantly decreased after 43 years<br />

of decay under Safe-Enclosure. However,<br />

according to calculations the<br />

concretes inside the containment wall<br />

and the inner parts of BS are still above<br />

the specific activity limit (I 0 ≥ 1).<br />

Radioactive waste [Mg]<br />

CON<br />

A2 Limits [5]<br />

BE<br />

A2 Limits [5]<br />

BE<br />

KCD Limits [1]<br />

26/03/1997 270 12<strong>08</strong> 929 939<br />

01/01/2010 230 510 322 467<br />

01/01/<strong>2020</strong> 200 386 223 341<br />

01/01/2030 180 272 157 244<br />

01/01/2040 140 189 123 193<br />

| Tab. 1.<br />

Estimated radioactive waste <strong>for</strong> KCD at different dismantlement dates.<br />

DECOMMISSIONING AND WASTE MANAGEMENT 415<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 416<br />

| Fig. 3.<br />

Radioactive (in red) and free-release (in green) regions of BS at different axial positions of the Dodewaard model.<br />

| Fig. 4.<br />

Radioactivity of BS at z = 2750 cm <strong>for</strong> two time steps. Left: reactor shut-down (26/03/1997).<br />

Right: the end of Safe Enclosure (01/01/2040).<br />

Table 1 gives an overview about<br />

the estimated radioactive waste<br />

masses in accordance with the release<br />

date. The second column in Table 1<br />

reports the values from [1], which are<br />

based on the gamma spectrometric<br />

measurements of samples taken from<br />

the BS at different heights and depths.<br />

Radially symmetric distribution was<br />

assumed, meaning that one sample<br />

was representative <strong>for</strong> specific activity<br />

at a certain height and depth in the BS.<br />

The calculated quantities are 4x<br />

higher at the end of life of the reactor,<br />

but at future dates (2010, <strong>2020</strong>,..)<br />

they are between 2x-1x the values in<br />

[1]. The best agreement between the<br />

calculated values and estimations in<br />

[1] are at year 2040, with differences<br />

of 10 % – 40 %, depending on the<br />

case.<br />

One of the reasons <strong>for</strong> the difference<br />

between [1] and the calculated<br />

values could be due to a number<br />

of short-lived nuclides, e.g. Mn-54,<br />

Zn-65 and Cs-134, which were not<br />

considered in Ref [1]. In [1] it is also<br />

not clear what clearance levels are<br />

used to determine the rad-waste<br />

quantities. However, there are several<br />

limits mentioned that could be inferred<br />

to be the clearance levels<br />

but several of these vary slightly<br />

from [5]. The principal results given<br />

here used the ADR A2 limits [5], but<br />

additional BE results used the inferred<br />

KCD limits (Depository Clearance<br />

Levels) [1].<br />

Furthermore, several assumptions<br />

and simplifications on the layout of<br />

facility were made in the original<br />

MCNP model. For some of the regions<br />

in BS there is not enough in<strong>for</strong>mation<br />

in [1] to assign the right concrete type<br />

(heavy or light) in the MCNP model.<br />

Ref. [1] also assumes radial distribution<br />

of activities in BS, but there are<br />

regions where at the same radius and<br />

height the concrete type varies. Even<br />

with the same type of concrete at<br />

certain height and radius the activity<br />

can vary. This is demonstrated in<br />

Figure 5, where the area just above<br />

the chimney (at z ∼ 2600 cm, look at<br />

Figure 1) is more radioactive than<br />

other parts at the same radius. In<br />

addition to that, the results on LFR<br />

reported in Ref. [2], which used the<br />

same calculation procedure as this<br />

work, show that calculations can<br />

overestimate the measured activities<br />

up to a factor of 4.<br />

3.2 LFR<br />

A similar calculation procedure <strong>for</strong> the<br />

BS specific activity estimation was<br />

­applied on LFR. Specific activities in<br />

FISPACT were calculated with the<br />

172-energy group activation library<br />

EAF-2010. Details of the procedure<br />

can be found in [2]. The 7<strong>09</strong>-energy<br />

group ENDF/B VIII.0 was not used<br />

here due to calculation time constraints<br />

(i.e. large amount of voxels<br />

was required to capture the asymmetrical<br />

layout of the facility).<br />

However, the test calculations on<br />

Dodewaard case indicate that the<br />

waste estimates with ENDF/B VIII.0<br />

would be somewhat lower.<br />

Figure 6 compares the radioactivity<br />

map at z = 81 cm (about<br />

50 cm above the active core) <strong>for</strong> the<br />

CON and BE cases. Purple and blue<br />

areas designate the regions with<br />

poor statistics, i.e. the regions with<br />

flux ­uncertainty more than 10 %<br />

and zero flux regions, respectively.<br />

In CON case the radioactivity is<br />

spread over a larger region around<br />

the reactor core.<br />

Decommissioning and Waste Management<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 5.<br />

Radioactivity map (Cartesian mesh)<br />

just above the chimney.<br />

disposed<br />

at COVRA<br />

In Table 2 the mass of radioactive<br />

concretes of different type disposed at<br />

COVRA between years 2017-2019 is<br />

given along with those BE and CON<br />

calculated as radioactive as of January<br />

2014. The radioactive foundation<br />

­concrete disposed at COVRA seems to<br />

lie between the two calculated values.<br />

The disposed barite concrete is 2x-4x<br />

higher than the calculated values. The<br />

barite blocks are large cubes with<br />

sides of about 1 m. One possible<br />

reason <strong>for</strong> the discrepancy is due to<br />

the voxel nature of the calculation<br />

which would cut up the barite cubes.<br />

It is possible that during decommissioning<br />

if one portion of barite was<br />

found to be radioactive the whole<br />

cube was considered radioactive,<br />

which would lead to greater amounts.<br />

The amount of radioactive scrap<br />

concrete is 5x higher comparing to<br />

calculations. One possible reason <strong>for</strong><br />

difference here is that the COVRA<br />

quantity includes the ‘irradiation<br />

wagon’, i.e. moveable shielding<br />

blocks, whereas the calculation does<br />

not consider this structure. Similarly<br />

to the barite, the scrap concrete is<br />

located in 4 large top-shield<br />

structures. These would likely not<br />

be cut up, but rather disposed of<br />

as a whole.<br />

Conclusions<br />

In this work the radioactivity of<br />

BS in two nuclear facilities (i.e. the<br />

KCD and LFR) was mapped out.<br />

The spread of calculated radioactivity<br />

can give us insight of more important<br />

Radioactive waste [Mg]<br />

CON<br />

calc Jan. 2014<br />

regions when planning the Safe<br />

Enclosure and dismantling, and can<br />

save cost and time <strong>for</strong> the activation<br />

measurements. Based on the<br />

mapped radioactive regions, the<br />

quantities of radioactive waste were<br />

estimated.<br />

Overall, the results from BE calculations<br />

match better the radioactive<br />

waste estimated from the activity<br />

measurements. However, even the<br />

CON case at the planned dismantlement<br />

date differs by less than a factor<br />

of 2 and 3 <strong>for</strong> Dodewaard and LFR,<br />

respectively. These reasonable results<br />

demonstrate that conservative quantities<br />

of radioactive waste at different<br />

dates could be estimated be<strong>for</strong>ehand,<br />

using the calculation procedure from<br />

this work.<br />

References<br />

[1] GKN report 99-001/PID/R: “Activation measurements of the<br />

Biological Shield”, Ministerie van Infrastructuur en Milieu.<br />

(2017). Europese overeenkomst voor het internationale<br />

vervoer van gevaarlijke goederen over de weg (ADR)<br />

[2] P. Young, et al. “Characterizing the Radioactivity of the<br />

Concrete Shielding during Decommissioning of the LFR”,<br />

Proc. Int. Conf. <strong>Nuclear</strong> Energy <strong>for</strong> New Europe 2019,<br />

Portorož, Slovenia<br />

[3] C.J. Werner, et al., “MCNP6.2 Release Notes”,<br />

LA-UR-18-2<strong>08</strong><strong>08</strong>, 2018<br />

BE<br />

calc Jan. 2014<br />

Barite 22.7 9.3 4.78<br />

Scrap 9.9 1.94 1.73<br />

Foundation 2.6 6.77 0<br />

| Tab. 2.<br />

Estimated radioactive waste <strong>for</strong> LFR.<br />

| Fig. 6.<br />

Comparison of the radioactivity spread in BS (in XY direction) 50 cm above the active core <strong>for</strong> CON (left) and<br />

BE (right) case of LFR.<br />

[4] M. Fleming, et al., The FISPACT-II User Manual,<br />

UKAEA-R(18)001, January 2018<br />

[5] Europese overeenkomst voor het internationale vervoer van<br />

gevaarlijke goederen over de weg (ADR) Bijlagen, 2017<br />

[6] S. W. Mosher, et al., “ADVANTG – An Automated Variance<br />

Reduction Parameter Generator”, ORNL/TM-2013/416 Rev. 1,<br />

August 2015<br />

[7] L. Snoj, M. Ravnik, “Calculation of <strong>Power</strong> Density with MCNP in<br />

TRIGA reactor”, Proc. Int. Conf. <strong>Nuclear</strong> Energy <strong>for</strong> New Europe<br />

2006, Portorož, Slovenia<br />

[8] https://pris.iaea.org/PRIS/CountryStatistics/<br />

ReactorDetails.aspx?current=422, accessed 18/12/2019<br />

Authors<br />

Lino Salamon<br />

salamon@nrg.eu<br />

Perry Young<br />

Lojze Gačnik<br />

<strong>Nuclear</strong> Research and Consultancy<br />

Group NRG<br />

Westerduinweg 3<br />

1755 Le Petten, Netherlands<br />

DECOMMISSIONING AND WASTE MANAGEMENT 417<br />

Decommissioning and Waste Management<br />

Radioactivity Calculation of the Concrete Shielding of the Petten LFR and the Dodewaard BWR ı Lino Salamon, Perry Young and Lojze Gačnik


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 418<br />

Quality Assurance and Data Analysis<br />

in Automated Radiological<br />

Characterization of Large Soil Volumes<br />

Christoph Klein, Marina Sokcic-Kostic and Felix Langer<br />

Planned entry <strong>for</strong><br />

At the dismantling of FBFC <strong>International</strong>’s fuel fabrication plant in Dessel/Belgium, more than<br />

34.000 tons of soil have already been successfully characterized and sorted by NUKEM’s system<br />

FREMES until end of July <strong>2020</strong>.<br />

The measurement and sorting process is executed fully automated and can process up to 13 tons<br />

per hour, having involved recording and evaluation of up to 16.500 gamma spectra per day.<br />

To ensure the constant quality of the results, a clear-cut algorithm as well as regular consistency checks and<br />

supervision are essential.<br />

This article describes the principal data evaluation of the system from recording to sorting, and gives an overview about<br />

the taken measures of data analysis and quality assurance. Practical representations and typical examples<br />

from operational experience illustrate, how the system reliably per<strong>for</strong>ms highly numerous and frequent gamma<br />

measurements, without necessary expertise of the site operators.<br />

| Fig. 1.<br />

Overview of soil characterization and sorting with FREMES: Material is filled in by truck (right),<br />

then buffered and transported by conveyors to the measurement/sorting belts (blue containers),<br />

and afterwards sorted to piles and containers (left side).<br />

1 Measurement operation<br />

The system FREMES <strong>for</strong> characterization<br />

and sorting of bulk material can<br />

process large quantities of soil, building<br />

rubble and other, in order to release<br />

or dispose of it properly. It provides<br />

measurement with a high throughput<br />

(up to 13 tons per hour <strong>for</strong> the soil<br />

measured in Dessel) by uninterrupted<br />

gamma-spectrometric measurements<br />

of a continuous material stream on the<br />

belt, which is then directly evaluated,<br />

characterized and sorted to the resulting<br />

destination. (See Figure 1) By this<br />

direct examination and sorting of<br />

100 % of the material, conservative<br />

assumptions about the activity (and<br />

there<strong>for</strong>e the amount of radioactive<br />

waste) are minimized.<br />

The central measurement process is<br />

per<strong>for</strong>med according to the following<br />

basic steps:<br />

p The material is buffered and a<br />

­continuous stream with a defined<br />

geometry suitable <strong>for</strong> measurement<br />

is created. (See Figure 2)<br />

p The material is virtually divided<br />

into cells, <strong>for</strong> which gamma spectra<br />

are subsequently taken by HPGe<br />

detectors and a scale takes the<br />

weight.<br />

p Immediately after recording,<br />

the data is evaluated and <strong>for</strong><br />

each separable portion the contained<br />

specific U-235 activity is determined<br />

automatically, in cluding<br />

uncertainty values ac cording to<br />

ISO 11929.<br />

p The result is compared to legal<br />

limits (including a safety margin)<br />

and classified into the categories:<br />

p Free Release (FR)<br />

p Conditional Release (CR)<br />

p (Low Level) Radioactive Waste<br />

(RW)<br />

p Material suspicious of a hot spot<br />

(HS) or other inconclusive<br />

| Fig. 2.<br />

Material is moved in continuously below the<br />

detectors in a defined geometry suited <strong>for</strong><br />

measurement.<br />

activity pattern, which is<br />

collected and re-measured later<br />

p According to the obtained characterization,<br />

the system directs the<br />

material to its storage destination<br />

via sorting belts.<br />

p All results and raw data are stored<br />

on the PC, from which documentation<br />

is automatically provided at<br />

material export.<br />

To guarantee high throughput, all<br />

these steps – including the gamma<br />

spectrum evaluation – run fully automated<br />

in a clear-cut and well-proven<br />

process and there<strong>for</strong>e can be supervised<br />

by an operator, who does not<br />

need to be an expert <strong>for</strong> radiation<br />

measurement.<br />

Considering the large amount<br />

of material, this is an important<br />

advantage, especially <strong>for</strong> costefficiency.<br />

In the application here,<br />

already around 34,000 tons have been<br />

examined over more than 2 years,<br />

­corresponding to around 1.5 million<br />

single gamma measurements and<br />

their evaluation.<br />

However, many aspects of the<br />

system (especially of the gamma<br />

measurement) still need to be supervised<br />

regularly by an expert, to make<br />

sure that the automated evaluation<br />

works reliably and without unintended<br />

effects. To ensure this, NUKEM<br />

employs numerous measures <strong>for</strong><br />

prevention of issues, regular maintenance<br />

and quality assurance, both<br />

remotely and together with the personnel<br />

operating the system on-site.<br />

2 Quality assurance<br />

of the results<br />

Regular activities and provisions of<br />

various types make sure that the<br />

results of the system are reliable and<br />

Decommissioning and Waste Management<br />

Quality Assurance and Data Analysis in Automated Radiological Characterization of Large Soil Volumes ı Christoph Klein, Marina Sokcic-Kostic and Felix Langer


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 3.<br />

The operators ensure reliable results and high throughput by super vising<br />

the process constantly over cameras and software displays. The system<br />

also can be controlled outside of the working station by a tablet PC.<br />

that possible issues are prevented or<br />

quickly fixed:<br />

p Operational / Mechanical:<br />

p Constant observation of the<br />

­operators over material flow<br />

and automated operation<br />

p Direct communication of<br />

operators to NUKEM experts<br />

<strong>for</strong> remediation of issues (telephone,<br />

remote PC connection,<br />

cameras)<br />

p Regular maintenance visits and<br />

checks of the system hardware<br />

p Radiological:<br />

p Daily measurements of radiation<br />

background and check<br />

<strong>for</strong> changes (per<strong>for</strong>med by the<br />

operators)<br />

p Regular reference measurements<br />

(1-2 days) to ensure<br />

correct function of gamma<br />

detectors and weighing system<br />

p Taking of representative<br />

samples from the material<br />

stream and comparison with<br />

an independent measurement<br />

p Regular consistency checks and<br />

deeper analyses of the recorded<br />

data by a radiological measurement<br />

expert<br />

p Automated synchronization<br />

and transfer of the data to<br />

ensure safe storage<br />

Since the system mechanically processes<br />

a large volume of material at all<br />

seasons (temperatures, weather conditions)<br />

issues of the mechanical systems<br />

and irregularities in the material<br />

flows are natural with these high<br />

volumes of soil. The operators there<strong>for</strong>e<br />

ensure that the throughput is as<br />

high as possible and the measurement<br />

conditions are always fulfilled, by<br />

super vising the process with the help<br />

of several technical means (see Figure<br />

3) and intervening when necessary.<br />

NUKEM experts are available on<br />

short notice if the solution of more<br />

complicated problems is demanded<br />

(e.g. with the belt movement<br />

auto matisation systems, or <strong>for</strong> radiological<br />

questions). Via ­telephone and<br />

remote connection to the measurement<br />

PC, problems can be solved in an<br />

easy and efficient way, which has<br />

proven to be very valuably over the<br />

operation period. This allows the quick<br />

solution of most issues and a fast<br />

resumption of operation, without<br />

requiring the presence of an expert<br />

on site in many of the cases, saving<br />

valuable cost and time.<br />

| Fig. 4.<br />

Results of regular QA measurements with a point source (blue), compared to the reference source activity<br />

(orange, with uncertainty area). Remark: Since smaller activities can regularly occur in practice by errors in<br />

positioning of the source on the belt, those values are expected and clear cases have been excluded (red marks).<br />

The reliability of the radiological<br />

results is regularly monitored by an<br />

expert <strong>for</strong> gamma spectrometry. The<br />

automated spectrum evaluation has<br />

always proven to give reliable results<br />

during tests and operation, so that<br />

there is no need <strong>for</strong> a sophisticated<br />

on-site supervision of the measurements.<br />

Instead, the results are<br />

regularly reviewed by experts to check<br />

the correctness and consistency of<br />

evaluation, and to identify possible<br />

| Fig. 5.<br />

Typical signature during review of mass measurements over three days (separated red). The material<br />

portions (blue) show stable masses with certain spread and expected inter ruptions (start/stop, pauses,<br />

etc.). A per<strong>for</strong>med change in the material transport on the third day is clearly visible.<br />

| Fig. 6.<br />

Typical example <strong>for</strong> measurement results from material with specific activity in the region 1-10 Bq/g<br />

U-tot. (Blue: activity results, red: limit of detection.) [Scale shown here in corresponding units of U235.]<br />

DECOMMISSIONING AND WASTE MANAGEMENT 419<br />

Decommissioning and Waste Management<br />

Quality Assurance and Data Analysis in Automated Radiological Characterization of Large Soil Volumes ı Christoph Klein, Marina Sokcic-Kostic and Felix Langer


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 420<br />

| Fig. 7.<br />

Exemplary display of the measurement results from one working shift during review: A set of<br />

uncontaminated material (activity below detection limit) is followed by material with different<br />

magnitudes of activity, which are sorted above or below the legal limit value, including a safety margin.<br />

| Fig. 8.<br />

Specific activities of a material set, which has been sorted to Free Release. (Data description as in<br />

Figure 7.) The specific activity can be seen reliably below the legal limit, including the safety margin.<br />

changes in the measurement conditions<br />

(e.g. changes of background,<br />

hardware issues of the detectors,<br />

unexpected effects, etc.).<br />

These checks typically consist of:<br />

p Remotely connected inspection of<br />

the measurement hardware<br />

p Check of measurement software<br />

and its history files<br />

p Examination of the regular<br />

measurements (background and<br />

reference source) <strong>for</strong> unexpected<br />

changes (also per<strong>for</strong>med qualitatively<br />

by the operators)<br />

p Review of the recorded measurement<br />

results<br />

p Manual inspection of raw<br />

measurement data and evaluation<br />

files<br />

p Regular synchronization to<br />

independent data storages, also<br />

­allowing efficient access to the<br />

data <strong>for</strong> analysis<br />

In addition to these regular checks,<br />

the correct functionality of the system<br />

is checked on site at least every 2 days<br />

by measurement of a reference point<br />

source. These measurements are<br />

required to correctly identify and<br />

provide the activity value of the<br />

source, by which correct energy and<br />

efficiency calibration as well as the<br />

stability of the system can easily be<br />

­assessed by the operators. An example<br />

of these measurements is shown in<br />

Figure 4, which clearly demonstrates<br />

a stable behavior over the measurement<br />

period.<br />

3 Regular review of<br />

measurement results<br />

An important part of the data review<br />

is the graphical display of the<br />

measurement results, by which the<br />

correct system behavior can be<br />

observed and unintended effects can<br />

be noticed. Examples <strong>for</strong> the typical<br />

signatures of mass and activity<br />

measurement and are shown in<br />

Figure 5 and Figure 6. In the latter<br />

display the quality of the results can<br />

be easily verified by check that the<br />

limits of detection remain at stable<br />

values and that the found activities<br />

results show a steady and plausible<br />

behavior (except deviation from<br />

known irregularities during measurement<br />

operation).<br />

Figure 7 shows a typical example<br />

of change from material with no<br />

measurable U235 activity to such with<br />

activity in the vicinity of the legal limit<br />

value <strong>for</strong> U-235, which corresponds<br />

to 1 Bq/g Utot. The latter can be<br />

­identified by FREMES to be below<br />

or above the limit and is sorted accordingly.<br />

An example <strong>for</strong> sorted free<br />

release material in Figure 8, shows<br />

that it has a specific activity<br />

­significantly below the limit and can<br />

be <strong>for</strong>warded to release.<br />

4 Summary<br />

The FREMES system now has operated<br />

reliably <strong>for</strong> over 2 years on the site<br />

in Dessel. It has successfully characterized,<br />

sorted and released over<br />

34,000 tons of material, corresponding<br />

to over one million single<br />

gamma measurements. The radiological<br />

measurement itself is per<strong>for</strong>med<br />

with high throughputs up to<br />

13 tons/hour fully automated under<br />

supervision of operators on site, who<br />

not need to be measurement experts.<br />

The employed numerous quality<br />

assurance procedures thereby guarantee<br />

the reliability with a minimum<br />

ef<strong>for</strong>t by experts themselves. The<br />

remediation of different issues (e.g.<br />

irregularities or hardware problems<br />

with material transport equipment)<br />

as well as radiological review of the<br />

data can be per<strong>for</strong>med efficiently by<br />

remote access in many cases.<br />

This allows the system to<br />

accomplish 100 % radiological<br />

characterization of material in a<br />

cost- effective way also in large<br />

amounts and with high throughput<br />

demands, and thereby minimizes the<br />

material to be classified as non-free<br />

release material as far as possible.<br />

Authors<br />

Dr. Christoph Klein<br />

christoph.klein@<br />

nukemtechnologies.de<br />

Dr. Marina Sokcic-Kostic<br />

Felix Langer<br />

NUKEM Technologies Engineering<br />

Services GmbH<br />

Industriestr. 13<br />

63755 Alzenau, Germany<br />

Decommissioning and Waste Management<br />

Quality Assurance and Data Analysis in Automated Radiological Characterization of Large Soil Volumes ı Christoph Klein, Marina Sokcic-Kostic and Felix Langer


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Errichtung einer Zerlegehalle<br />

für Groß komponenten am Standort des<br />

Entsorgungswerkes für Nuklearanlagen<br />

in Lubmin<br />

Andreas Fuchs und Bernhard Olm<br />

1 Einleitung Für die Konditionierung von radioaktiven Reststoffen und Abfällen, welche<br />

derzeit in den Lagerhallen des Zwischenlagers Nord (ZLN) der EWN GmbH zwischengelagert<br />

werden, beabsichtigt die EWN die Errichtung und den Betrieb einer Zerlegehalle (ZLH) für<br />

Großkomponenten (Reaktordruckgefäße (RDG) und Dampferzeuger (DE), RDG-Einbauten, etc.)<br />

am Standort Lubmin/Rubenow. Aufbau und Auslegungsmerkmale sowie in der ZLH geplante<br />

Tätigkeiten werden vorgestellt.<br />

2 Aufbau der Zerlegehalle<br />

Am Standort Lubmin/Rubenow der<br />

EWN Entsorgungswerk für Nuklearanlagen<br />

GmbH (EWN) befinden sich<br />

im östlichen Bereich des Betriebsgeländes<br />

die Gebäude Zentrale Aktive<br />

Werkstatt (ZAW) und Zentrale Dekontaminations-<br />

und Wasseraufbereitungsanlage<br />

(ZDW).<br />

Für den Betrieb der Einrichtungen<br />

der ZAW und ZDW ist der Umgang mit<br />

radioaktiven Stoffen, insbesondere<br />

die Bearbeitung und Konditionierung<br />

von radioaktiven Reststoffen und<br />

­Abwässern aus den Kernkraftwerken<br />

Greifswald (KGR) und Rheinsberg<br />

(KKR), sowie aus anderen kerntechnischen<br />

Anlagen mit Leichtwasserreaktoren<br />

genehmigt.<br />

Die Möglichkeiten der Bearbeitung<br />

und Konditionierung sind nicht für<br />

alle radioaktiven Reststoffe und<br />

­Abfälle, welche derzeit überwiegend<br />

in den Lagerhallen des Zwischen lagers<br />

Nord – Abfalllager (ZLN-AL) der EWN<br />

GmbH zwischengelagert werden, ausreichend.<br />

Dies betrifft insbesondere<br />

Großkomponenten, z. B. Reaktodruckgefäße<br />

(RDG) und Dampf er zeuger<br />

(DE), sowie Komponenten mit hohem<br />

radiologischem Gefährdungspotential<br />

z. B. RDG-Einbauten in Abschirm- und<br />

Transportvorrichtungen oder in nicht<br />

endlagerfähigen Behältern.<br />

Für die Konditionierung dieser<br />

­radioaktiven Reststoffe und Abfälle,<br />

inklusive endlagerfähiger Verpackung<br />

der Abfälle, beabsichtigt die EWN<br />

die Errichtung und den Betrieb<br />

einer Zerlegehalle (ZLH) für Großkomponenten<br />

am Standort Lubmin/<br />

Rubenow.<br />

Die ZLH soll als Anbau an die<br />

vorhandenen Gebäude südlich der<br />

ZAW und östlich der ZDW errichtet<br />

und betrieben werden. Die ZLH wird<br />

baulich und technologisch mit der<br />

ZDW und der ZAW zu einem Reststoffbearbeitungs-<br />

und Abfallbehandlungszentrum<br />

verbunden und bildet<br />

damit einen zusammenhängenden<br />

Gebäudekomplex.<br />

In der Abbildung 1 ist die Lage der<br />

ZLH zu den angrenzenden Gebäuden<br />

ZAW (Zentrale Aktive Werkstatt) und<br />

ZDW (Zentrale Dekontaminationsund<br />

Wasseraufbereitungsanlage) ersichtlich.<br />

Abbildung 2 ist eine<br />

­Visu­alisierung der Zerlegehalle mit<br />

den angrenzenden Gebäuden ZAW<br />

und ZDW.<br />

Die ZLH besteht im Wesentlichen<br />

aus einem Zerlegebereich mit vorgelagerter<br />

Schleuse, einer Bereit stellungshalle<br />

mit vorgelagerter Schleuse,<br />

der Personenschleuse (Umkleidebereiche<br />

inaktiv und aktiv) mit Kontrollbereichszugang,<br />

Sanitär räumen, einem<br />

Aufenthaltsraum, ­Büros und den<br />

Lüftungszentralen sowie sonstigen<br />

Infra­strukturräumen und Verkehrswegen.<br />

In die Bereit stellungshalle ist<br />

ein Raum zur ­Aufbewahrung von Ölen<br />

und Fetten integriert. Die Abbildung 3<br />

| Abb. 1.<br />

Lageplan der Zerlegehalle.<br />

Planned entry <strong>for</strong><br />

DECOMMISSIONING AND WASTE MANAGEMENT 421<br />

| Abb. 2.<br />

3D-Ansicht der Zerlegehalle mit angrenzender ZAW/ZDW.<br />

| Abb. 3.<br />

Schnitt durch den Zerlegebereich mit vorgelagerter Schleuse.<br />

Decommissioning and Waste Management<br />

Construction of a Dismantling Hall <strong>for</strong> Large Components at Entsorgungswerk für Nuklearanlagen in Lubmin ı Andreas Fuchs and Bernhard Olm


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 422<br />

| Abb. 4.<br />

Schnitt durch die Bereitstellungshalle und die darüber befindliche Personenschleuse.<br />

zeigt einen Schnitt durch die Zerlegehalle<br />

im Bereich der Schleuse sowie<br />

den Zerlegebereich und die bestehende<br />

ZAW.<br />

Abbildung 4 zeigt einen Schnitt<br />

durch die Bereitstellungshalle mit<br />

der darüber befindlichen Personenschleuse.<br />

Die Errichtung und der Betrieb<br />

der ZLH wurde als Änderungs­genehmigung<br />

nach § 7 Strahlenschutzverordnung<br />

(StrlSchV) für den bestehenden<br />

Komplex ZAW/ZDW<br />

beantragt.<br />

3 Ausbaustufen und<br />

geplante Tätigkeiten<br />

In der ZLH werden die im ZLN-AL<br />

lagernden Dampferzeuger und die aktivierten<br />

Komponenten, die beim<br />

Rückbau der Kernkraftwerke Greifswald<br />

(KGR) und Rheinsberg (KKR)<br />

ausgebaut wurden, bearbeitet und<br />

konditioniert.<br />

Bei der Bearbeitung radioaktiver<br />

Reststoffe wird das Ziel verfolgt, den<br />

sicheren Umgang mit radioaktiven<br />

Stoffen zu gewährleisten und weiterhin<br />

die größtmöglichen Massen unter<br />

Beachtung des ALARA-Prinzips und<br />

Abwägung des Aufwand-/Nutzenprinzips<br />

der uneingeschränkten Freigabe<br />

oder der zweckgerichteten Freigabe<br />

nach StrlSchV zuzuführen.<br />

Die bei der Bearbeitung anfallenden<br />

radioaktiven Abfälle werden in endlagerfähige<br />

Behälter verpackt.<br />

Folgende Großkomponenten und<br />

aktivierte Komponenten sind zu bearbeiten:<br />

p Großkomponenten KGR und KKR<br />

p Dampferzeuger (DE) der Blöcke<br />

1, 2, 3 und 4<br />

p Reaktordruckgefäße (RDG) der<br />

Blöcke 1, 2 und 5 des KGR<br />

p Reaktordruckbehälter (RDB) des<br />

KKR<br />

p RDG der Blöcke 3 und 4<br />

des KGR, jeweils mit den<br />

RDG-Einbauten Reaktorschacht<br />

(RS) und Schachtboden (SB)<br />

p RDG-Einbauten Schutzrohrblock<br />

(SRB), Kassettenkorb<br />

(KK), RS und SB des Blockes 5<br />

des KGR<br />

p RDG-Einbauten SRB und KK<br />

der Blöcke 3, 4 und 5 des KGR<br />

p Aktivierte Komponententeile KGR<br />

und KKR<br />

p Bodenplatten der RDG-Einbauten<br />

SRB und KK der Blöcke 1 und<br />

2 des KGR<br />

p Obere Platten der RDG-Einbauten<br />

SB der Blöcke 1 und 2<br />

des KGR<br />

p Segmente der Ringwasserbehälter<br />

(RWB) der Blöcke 1, 2,<br />

3 und 4 des KGR<br />

p Corebauteile (CBT) des KGR<br />

p CBT des KKR<br />

Des Weiteren müssen nachfolgende,<br />

bereits vorzerlegte und in Stahl- und<br />

Betoncontainern verpackte, aktivierte<br />

Abfälle ggf. nach vorheriger Nachzerlegung<br />

in endlagerfähige Behälter<br />

umverpackt werden:<br />

p die RDB-Einbauten SRB, KK, RS<br />

und SB der Blöcke 1 und 2<br />

p die RDB-Einbauten Druckgestell<br />

(DST), Brennelementekorb (BEK),<br />

RS und SB des KKR<br />

p CBT aus der Entsorgung der<br />

Schachtläger für aktivierte Betriebsabfälle<br />

des KGR<br />

p sowie sonstige aktivierte Abfälle<br />

Die Reststoffbehandlung und Abfallkonditionierung<br />

soll in Arbeits­paketen<br />

erfolgen. Die verschiedenen Arbeitspakete<br />

werden – je nach Arbeits<strong>for</strong>tschritt<br />

und Planungsstand –<br />

in Ausbaustufen zusammengefasst.<br />

Zwischen der Abarbeitung der verschiedenen<br />

Arbeitspakete müssen<br />

i. d. R. die Gerätschaften zur Reststoffbehandlung<br />

und Abfallkonditionierung<br />

den zur Zerlegung vorgesehenen<br />

Großkomponenten und<br />

Teilen angepasst werden.<br />

Für die ZLH sind aus technologischer<br />

und radiologischer Sicht<br />

vier Ausbaustufen vorgesehen. Die<br />

erste Ausbaustufe – die Dekontamination,<br />

Zerlegung und Verpackung<br />

der Dampferzeuger – wird detailliert<br />

geplant und ist Antragsgegenstand.<br />

Zum jetzigen Zeitpunkt werden<br />

auch die späteren Ausbaustufen 2<br />

bis 4 in einem Detaillierungsgrad<br />

beschrieben, der es erlaubt, wesentliche,<br />

für die Planung des Gebäudes er<strong>for</strong>derliche<br />

Auslegungsdaten abzuleiten<br />

und darüber hinaus die gene relle<br />

Machbarkeit der zukünf tigen Hantierungsvorgänge<br />

sowie der Zerlege- und<br />

Verpackungsprozesse zu bewerten.<br />

Die detaillierte Planung der Ausbaustufen<br />

2 bis 4 erfolgt jeweils rechtzeitig<br />

vor dem geplanten Beginn der<br />

Tätigkeiten der jeweiligen Ausbaustufe<br />

anhand der dann möglicherweise<br />

neu zur Verfügung stehenden<br />

Hantierungs- und Zerlegetechnik. Die<br />

Ausbaustufen 2 bis 4 werden über<br />

das atomrechtliche Aufsichtsverfahren<br />

unter Berücksichtigung der<br />

zum Zeitpunkt der Antragsstellung<br />

gültigen Gesetzeslage bei der Aufsichtsbehörde<br />

beantragt.<br />

Nachfolgend sind die vorgesehenen<br />

Ausbaustufen aufgelistet:<br />

p Ausbaustufe 1:<br />

Trockenzerlegung (DE)<br />

p Ausbaustufe 2/1:<br />

Trockenzerlegung – Erweiterung<br />

(RWB)<br />

p Ausbaustufe 2/2:<br />

Nass-/Trockenzerlegung (RDG/<br />

RDB, teilweise mit Einbauten)<br />

p Ausbaustufe 3:<br />

Umverpackung der aktivierten<br />

­Abfälle aus Zwischenlagerbehältern<br />

in Endlagercontainer<br />

p Ausbaustufe 4/1:<br />

Trockenzerlegung (RDG-<br />

Einbauten Block 5 KGR)<br />

p Ausbaustufe 4/2: Nasszerlegung<br />

(RDG-Einbauten KGR und CBT<br />

KGR und KKR)<br />

4 Dekontamination<br />

und Zerlegung der<br />

Dampferzeuger<br />

in der 1. Ausbaustufe<br />

In der ersten Ausbaustufe der ZLH<br />

ist geplant, dass die 21 im ZLN-AL<br />

lagernden Dampferzeuger (DE) dekontaminiert,<br />

zerlegt und die entstandenen<br />

Reststoffe dem Stoffkreislauf<br />

(Rezyklierung) zugeführt<br />

werden. Die entstehenden bzw.<br />

­verbleibenden radioaktiven Abfälle<br />

werden in endlagerfähige Behälter<br />

verpackt.<br />

Die DE wurden unter Demontagebedingungen<br />

aus den Blöcken 1-4<br />

Decommissioning and Waste Management<br />

Construction of a Dismantling Hall <strong>for</strong> Large Components at Entsorgungswerk für Nuklearanlagen in Lubmin ı Andreas Fuchs and Bernhard Olm


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

­entnommen und befinden sich als<br />

Großkomponenten in der Halle 7 des<br />

ZLN-AL. Die DE sind entleert und alle<br />

Öffnungen sind mit Stahlblechen verschlossen.<br />

Weiterhin sind die DE an<br />

der berührbaren Oberfläche durch<br />

eine Oberflächenbeschichtung bzw.<br />

Dekontamination frei von abwischbarer<br />

Kontamination.<br />

Das Gesamtgewicht jedes DE<br />

­beträgt 157 Mg. Die max. Gesamtaktivität<br />

pro DE liegt bei 1,1 E+11 Bq<br />

(Bezugszeitpunkt 01/2021).<br />

Vor der Zerlegung der DE werden<br />

die Nadelrohre im Zerlegebereich<br />

primärseitig dekontaminiert (siehe<br />

Abbildung 5), um das oben genannte<br />

Ziel der Rezyklierung der unverschlossenen<br />

Nadelrohre zu erreichen.<br />

Nach erfolgreicher Dekontamination<br />

wird der DE auf den Zerlegeplatz<br />

verbracht, um diesen in liegender<br />

Position zu zerlegen. Der DE wird<br />

eingerüstet, um die Zugänglichkeit zu<br />

gewährleisten (siehe Abbildung 6).<br />

Die Zerlegung kann in einem separaten,<br />

lufttechnisch abgeschlossenen<br />

Zerlegebereich erfolgen. Als Zerlegeverfahren<br />

kommen u. a. Seilsäge,<br />

mechanische Trennwerkzeuge wie<br />

Winkelschleifer, Abbruchhammer,<br />

Stichsäge und thermische Trennwerkzeuge<br />

wie Autogenbrenner zur Anwendung.<br />

| Abb. 5.<br />

Dampferzeuger im Zerlegebereich während der Dekontamination.<br />

| Abb. 6.<br />

Dampferzeuger auf dem Zerlegeplatz.<br />

DECOMMISSIONING AND WASTE MANAGEMENT 423<br />

5 Wesentliche Merkmale<br />

der weiteren<br />

Ausbaustufen<br />

Ab Ausbaustufe 2 wird eine Verpackungsstation<br />

eingerichtet, die es<br />

ermöglicht, die zerlegten Komponenten<br />

endlagergerecht in Konrad-<br />

Container zu verpacken.<br />

Dabei werden zunächst die Ringwasserbehältersegmente<br />

nachzerlegt<br />

und verpackt.<br />

Diese befinden sich in 20’-Containern<br />

(52 Stück). Die RWB-Segmente<br />

befinden sich in liegender Position<br />

auf einem Tragrahmen im Container.<br />

Die Masse der einzelnen RWB-<br />

Segmente variiert zwischen ca. 10<br />

und 22 Mg.<br />

Die Anlieferung der 20’-Container<br />

in die ZLH erfolgt mit einer Transporteinheit<br />

bestehend z. B. aus Zugmaschine<br />

und Plattenwagen (siehe<br />

Abbildung 7). Die Transporteinheit<br />

fährt über die vorgelagerte Schleuse<br />

in den Zerlegebereich, wo der<br />

20’-Container mit dem Brückenkran<br />

abgeladen wird. Nachdem die<br />

Transporteinheit den Zerlegebereich<br />

verlassen hat, wird der Containerdeckel<br />

abgenommen. Das RWB-Segment<br />

wird mit dem Brückenkran und einer<br />

Traverse entladen und auf dem<br />

| Abb. 7.<br />

Antransport des Containers.<br />

| Abb. 8.<br />

Zerlegung RDG Block 5.<br />

Decommissioning and Waste Management<br />

Construction of a Dismantling Hall <strong>for</strong> Large Components at Entsorgungswerk für Nuklearanlagen in Lubmin ı Andreas Fuchs and Bernhard Olm


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

DECOMMISSIONING AND WASTE MANAGEMENT 424<br />

| Abb. 9.<br />

Nasszerlegung: geöffneter RDG mit Peripherie.<br />

vorbereiteten Zerlegeplatz abgelegt.<br />

Anschließend wird der 20’-Container<br />

verschlossen und aus dem Zerlegebereich<br />

herausgefahren.<br />

Die Zerlegung der RWB-Segmente<br />

soll aus jetziger Sicht in liegender<br />

Position erfolgen. Die Zerlegung wird<br />

an der Luft in einem separaten lufttechnisch<br />

abgeschlossenen Zerlegebereich<br />

(Einhausung) mit lokalen<br />

­Abschirmungen und Interventionsmöglichkeiten<br />

sowie Absaugung erfolgen.<br />

Als Zerlegeverfahren kommen<br />

u. a. die Seilsäge, mechanische Trennwerkzeuge<br />

wie Winkelschleifer, Abbruchhammer<br />

und thermische Trennwerkzeuge<br />

wie Autogenbrenner zur<br />

Anwendung. Zur Minimierung der<br />

endzulagernden Abfallmenge werden<br />

ggf. zerlegte Teile in der ZAW nachzerlegt<br />

und dekontaminiert, damit<br />

eine Freimessung nach den gültigen<br />

Freigabekriterien erfolgen kann.<br />

Die Teile, die nicht freigemessen<br />

werden können, werden in endlagerfähige<br />

Behälter verpackt und<br />

ggf. entsprechend den An<strong>for</strong>derungen<br />

für das Endlager Konrad in der<br />

Bereit stellungshalle mit Beton vergossen.<br />

Darauffolgend werden – ebenfalls<br />

im Rahmen der 2. Ausbaustufe – sämtliche<br />

im ZLN gelagerte RDG/RDB<br />

zerlegt.<br />

Zunächst ist beabsichtigt, das<br />

Reaktordruckgefäß Block 5 in liegender<br />

Position trocken zu zerlegen<br />

(Ab bil dung 8). Die Zerlegung wird an<br />

der Luft in einem separaten lufttechnisch<br />

abgeschlossenen Zerlege bereich<br />

(Einhausung) mit lokalen Abschirmungen<br />

und Interventionsmöglichkeiten<br />

erfolgen. Als Zerlegeverfahren<br />

kommen u. a. die Bandsäge, die Seilsäge,<br />

mechanische Trennwerkzeuge<br />

(fernhantiert oder fernbedient) und<br />

thermische Trennwerkzeuge wie<br />

Auto­genbrenner (fernhantiert oder<br />

fern­bedient) zur Anwendung.<br />

Die RDG der Blöcke 3 und 4 mit<br />

ihren Einbauten RS und SB sind aufgrund<br />

ihrer Masse und ihrer geometrischen<br />

Gegebenheiten auslegungsbestimmend<br />

für die Abmessungen des<br />

Zerlegebereiches, für die Kranhakenbzw.<br />

Gebäudehöhe sowie für die<br />

statische Belastung. Sie sind somit<br />

abdeckend für die Betrachtung der<br />

Großkomponenten bezüglich der<br />

maximal zu handhabenden Massen<br />

und der maximal zu handhabenden<br />

Geometrien.<br />

Die RDG der Blöcke 3 und 4<br />

wurden jeweils im Ganzen aus ihrer<br />

Einbaulage entfernt und in liegender<br />

Position ins ZLN-AL verbracht.<br />

Zuvor wurde ein ca. 5,5 m langer<br />

| Abb. 10.<br />

Umverpackung mit Abschirmglocke.<br />

Abschirmzylinder im Core-Bereich<br />

außen am RDG montiert. Anschließend<br />

wurden die RDG-Einbauten RS<br />

und SB in das RDG eingesetzt. Abschließend<br />

wurde jeweils eine Abschirmplatte<br />

mit Lastanschlagpunkt<br />

(Lastaufnahme­öse) unter Verwendung<br />

der vorhandenen Gewindesacklochbohrungen<br />

am RDG-Flansch verschraubt.<br />

Für diese RDG wurde ein Konzept<br />

entwickelt, bei dem vorgesehen ist,<br />

die Behälter in der Grube des Zerlegebereiches<br />

aufzurichten, mit Wasser zu<br />

füllen und die enthaltenen Einbauten<br />

fernbedient unter Wasser zu zerlegen<br />

(siehe Abbildung 9)<br />

In Ausbaustufe 3 ist im Wesentlichen<br />

ein Umverpacken von zerlegten<br />

und in Stahl- oder Betoncontainern<br />

verpackten Teilen in endlagerfähige<br />

Behälter vorgesehen.<br />

Es befinden sich Einsatzkörbe/<br />

Sammelbehälter mit aktivierten bzw.<br />

kontaminierten Abfällen in Stahloder<br />

Betoncontainern im Abfalllager<br />

des Zwischenlagers Nord. Bei den<br />

­Abfällen handelt es sich u. a. um<br />

­Reaktoreinbauten der Blöcke 1 und 2<br />

des KGR bzw. des Rheinsberger<br />

Reaktors.<br />

Diese Abfälle befinden sich in ­<br />

Stahl- oder Betoncontainern, welche<br />

keine Endlager-Zulassung besitzen.<br />

Deshalb müssen diese in endlagerfähige<br />

Behälter umverpackt werden.<br />

Auf Grund der radiologischen Randbedingungen<br />

erfolgt die Um verpackung<br />

fernbedient. Hierfür wird die im Übergang<br />

zwischen Zerlege bereich und<br />

­Bereitstellungshalle eingebaute Verpackungsstation<br />

aus der Ausbaustufe 2<br />

weitergenutzt. Zusätzlich wird im<br />

Zerlegebereich eine Umverpackungsstation<br />

errichtet. In Abbildung 10<br />

sind die Umverpackungs station und<br />

die Verpackungsstation konzeptionell<br />

dargestellt.<br />

Decommissioning and Waste Management<br />

Construction of a Dismantling Hall <strong>for</strong> Large Components at Entsorgungswerk für Nuklearanlagen in Lubmin ı Andreas Fuchs and Bernhard Olm


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Abb. 11.<br />

Nasszerlegebecken – Einrichtung als Trockenzerlegeplatz.<br />

| Abb. 12.<br />

Absetzen eines Kassettenkorbes im Nasszerlegeplatz.<br />

vorgesehen, die Inbetriebnahme<br />

2023.<br />

Die derzeitige Terminplanung<br />

sieht die Verarbeitung der im ZLN<br />

gelagerten Großkomponenten in der<br />

Zerlegehalle bis 2059 vor.<br />

Mit der Errichtung der Zerlegehalle<br />

für Großkomponenten am<br />

Standort des Entsorgungswerkes für<br />

Nuklearanlagen in Lubmin wird eine<br />

wesentliche Voraussetzung dafür<br />

geschaffen, dass sämtliche im<br />

­Zwischenlager Nord (ZLN) befindlichen<br />

Reststoffe behandelt und in<br />

endlagerfähige Verpackungen überführt<br />

werden können.<br />

Autoren<br />

Andreas Fuchs<br />

andreas.fuchs@steag.com<br />

Steag Energy Services GmbH<br />

Rüttenscheider Straße 1–3<br />

45128 Essen, Germany<br />

Bernhard Olm<br />

bernhard.olm@ewn-gmbh.de<br />

EWN Entsorgungswerk<br />

für Nuklearanlagen GmbH<br />

Latzower Straße 1,<br />

175<strong>09</strong> Rubenow, Germany<br />

DECOMMISSIONING AND WASTE MANAGEMENT 425<br />

| Abb. 13.<br />

Baufeld nach Fertigstellung der Sauberkeitsschicht im Mai <strong>2020</strong>.<br />

In Ausbaustufe 4 wird das für die<br />

Grube im Zerlegebereich geplante<br />

Becken eingebaut und im ersten<br />

Schritt als Trockenzerlegeplatz genutzt<br />

(siehe Abbildung 11).<br />

Zunächst ist die Zerlegung von in<br />

sogenannten Abschirm- und Transportvorrichtungen<br />

(ATV) verpackten<br />

Reaktoreinbauten vorgesehen.<br />

Nach Einbau der Nasszerlegetechnik<br />

als Nasszerlegeplatz erfolgt die<br />

Unterwasserzerlegung von hochaktivierten<br />

RDB-Einbauten und CBT,<br />

wie in Abbildung 12 dargestellt.<br />

6 Aktueller Stand<br />

des Projektes<br />

Die Errichtung der Zerlegehalle hat<br />

mit der Herstellung der Tiefgründung<br />

auf ca. 330 Großbohrpfählen auf<br />

Basis einer vorgezogenen Teilbaugenehmigung<br />

im November 2018<br />

begonnen (siehe Abbildung 13). Die<br />

letzten Pfähle wurden im Januar <strong>2020</strong><br />

fertig gestellt.<br />

Der Hochbau beginnt mit der<br />

Herstellung der Bodenplatte im<br />

Sommer <strong>2020</strong>, die Fertigstellung<br />

des Gebäudes ist bis Ende 2022<br />

Decommissioning and Waste Management<br />

Construction of a Dismantling Hall <strong>for</strong> Large Components at Entsorgungswerk für Nuklearanlagen in Lubmin ı Andreas Fuchs and Bernhard Olm


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

426<br />

ENVIRONMENT AND SAFETY<br />

Planned entry <strong>for</strong><br />

Current Procedure <strong>for</strong> Determining<br />

Release Parameters <strong>for</strong> a Plane Crash<br />

on a <strong>Nuclear</strong> Facility in the Context<br />

of Accident Analyses<br />

Steffen Böhlke and Henrik Niegoth<br />

The possible accidents in nuclear<br />

facilities are examined in an accident<br />

analysis according to §13 (1) (6a)<br />

StrlSchG /L-2/. The determination of<br />

the range of assumed incidents is<br />

per<strong>for</strong>med on a case-by-case basis<br />

taking into account location- and<br />

­facility-specific conditions and using<br />

the definitions contained in the<br />

Incident Guidelines /L-3/, in the<br />

various <strong>Nuclear</strong> Waste Management<br />

Commission (= Entsorgungskommission<br />

(ESK)) guidelines /L-4/, /L-5/<br />

and /L-6/, and in the licensing constraints<br />

and conditions. A distinction<br />

must be made between mal functions,<br />

design basis accidents, and beyonddesign-basis<br />

events. The design basis<br />

accidents are differen tiated into the<br />

categories “internal impacts” and<br />

“ external impacts”. The external<br />

­impacts in turn are classified as<br />

natural or man-made impacts.<br />

Due to the low probability of<br />

occurrence, the incident “plane crash”<br />

normally is included among the<br />

beyond-design-basis events <strong>for</strong> which<br />

measures to reduce damage have to be<br />

considered. Such measures then are<br />

adequate if the radiological effects do<br />

not necessitate any relevant disaster<br />

protection measures in accordance<br />

with /L-7/.<br />

In recent years, STEAG Energy<br />

Services has established a licensable<br />

standard procedure <strong>for</strong> accident<br />

analyses <strong>for</strong> new-build nuclear<br />

facilities or facilities undergoing<br />

modernization and refurbishment,<br />

inside and outside Germany, which is<br />

generally approved by experts and<br />

licensing authorities. For this procedure<br />

the “plane crash” incident is<br />

always a special case since the release<br />

parameters cannot be directly taken<br />

from generally applicable bodies of<br />

regulations, standards or codes.<br />

1 Introduction For protection against major safety-related incidents in nuclear facilities,<br />

<strong>for</strong> example a plane crash on a radioactive waste storage facility or a conditioning plant, provision<br />

has to be made in the planning <strong>for</strong> structural, or other technical safety measures against potential<br />

accidents in order to limit the release of radioactive substances in the environment<br />

of the facility. The planning limits contained in § 104 of the “Radiation Protection Ordinance”<br />

(= Strahlenschutzverordnung (StrlSchV) /L-1/), “Limitation of exposure due to accidents”,<br />

in conjunction with § 194 StrlSchV/L-1/, are to be taken as basis <strong>for</strong> this.<br />

In the following, the general procedure<br />

<strong>for</strong> determining radiological<br />

impacts is explained taking the<br />

example of a plane crash on a storage<br />

building <strong>for</strong> radioactive waste. In<br />

addition to the determination of the<br />

load impacts occurring in case of a<br />

plane crash, it is necessary to<br />

­determine the nuclide-­specific activity<br />

inventory of the waste packages and<br />

the respective release parameters<br />

and to per<strong>for</strong>m the cor responding<br />

dispersion and dose cal culations in<br />

order to evaluate the facility’s radiological<br />

impact on its environment.<br />

The method can also be adapted to<br />

conditioning plants <strong>for</strong> radioactive<br />

waste, i.e., facilities <strong>for</strong> treatment of<br />

residual materials.<br />

2 Plane crash scenario<br />

The hypothetical scenario refers to a<br />

fully fueled military aircraft crashing<br />

at high speed into a storage building<br />

containing waste packages, causing a<br />

fire within the facility because of the<br />

kerosene carried by the aircraft. The<br />

appropriate load assumptions <strong>for</strong> a<br />

plane crash, e.g. details of the impact<br />

load-time diagram or the impact<br />

area and impact angle, are described<br />

in the Reactor Safety Commission<br />

(= Reaktor-Sicherheitskommission<br />

(RSK)) guidelines /L-8/.<br />

The sequence of a worst-case plane<br />

crash into a storage building can be<br />

described as follows: As conservative<br />

assumption, the military aircraft<br />

crashes vertically into the roof of the<br />

storage building. Upon penetrating<br />

the roof of the building, it punches a<br />

piece of concrete debris out of the<br />

roof. At the same time, as parts break<br />

off and kerosene leaks out, the body of<br />

the aircraft loses mass, the kinetic<br />

energy of which, however, does not<br />

contribute to the pattern of damage.<br />

The rest of the aircraft is decelerated<br />

and, together with the concrete<br />

debris, falls with reduced velocity<br />

onto the waste packages stored below,<br />

inside the storage building. If a<br />

­statistical verification is not available,<br />

additionally the collapse of the<br />

collapse of further parts of the roof is<br />

assumed. The beams of the roof<br />

structure next to the area of impact<br />

are also affected by the plane crash. A<br />

conservative assumption is that thereby<br />

the roof structure will collapse<br />

across the entire width of the building<br />

damaging the waste packages below.<br />

In the affected stacks of packages,<br />

the kinetic energy is consumed proportionately<br />

and, depending on the<br />

mass of the impacting pieces of debris<br />

and the masses of the packages in the<br />

stack, is distributed among the waste<br />

packages through impact analysis.<br />

Mechanical loading can occur in the<br />

process, acting, on the one hand, on<br />

waste packages stored directly under<br />

the area of impact and, on the other<br />

hand, on neighboring waste packages,<br />

as a result of falling parts of the roof.<br />

Waste packages located only partially<br />

within one of the two areas subjected<br />

to loading are considered to be fully<br />

impacted. This covers the action of<br />

­additionally flying debris.<br />

Furthermore, a fire has to be<br />

ex pected to occur subsequent to<br />

the crash due to the kerosene carried<br />

on board the military aircraft.<br />

Making allowance <strong>for</strong> the debris<br />

­covering the floor after the impact,<br />

the depth of the resulting kerosene<br />

lake is ascertained through geometric<br />

analysis in order to determine the<br />

maximum duration of a fire. Within<br />

the scenario the fire is assumed to<br />

burn <strong>for</strong> at least 30 minutes. As a<br />

­consequence of the kerosene fire<br />

on the floor of the entire storage<br />

Environment and Safety<br />

Current Procedure <strong>for</strong> Determining Release Parameters <strong>for</strong> a Plane Crash on a <strong>Nuclear</strong> Facility in the Context of Accident Analyses<br />

ı Steffen Böhlke and Henrik Niegoth


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

facility, a part of the waste packages<br />

additionally are subjected to thermal<br />

loading.<br />

3 Nuclide vector and<br />

activity inventory<br />

In addition to the properties of the<br />

waste packages themselves, par ticularly<br />

the radiological characteristics<br />

of the radioactive wastes, <strong>for</strong> example<br />

the nuclide composition or the activity<br />

inventory, are important input data<br />

<strong>for</strong> determining the radiological<br />

effects subsequent to an accident.<br />

A nuclide composition that covers<br />

the occurring radioactive wastes can<br />

be estimated on the basis of analyses<br />

of the operation and post-operation of<br />

the nuclear power plant at the site of<br />

the planned storage building, or on<br />

the basis of experience from the<br />

dismantling of parts of other nuclear<br />

power plants. For calculation of the<br />

release fractions, the nuclides are<br />

­classified particularly on the basis of<br />

their volatility under the assumed<br />

load impacts.<br />

To determine the activity inventory<br />

one can rely on empirical values or<br />

concrete in<strong>for</strong>mation from databases<br />

of existent nuclear facilities. In the<br />

case of new-build facilities, conservative<br />

assumptions are made with the<br />

aid of radiological planning guides.<br />

The activity inventory of the waste<br />

packages also can be deduced from<br />

dose rate criteria.<br />

4 Release fractions<br />

Extensive investigations of release<br />

parameters and concrete in<strong>for</strong>mation<br />

about release fractions can be found in<br />

the Konrad Transport Study /L-9/,<br />

together with the corrections in the<br />

corresponding more detailed study<br />

/L-10/ and verification /L11/, as well<br />

as in the ESK statement on the stress<br />

test <strong>for</strong> storage facilities <strong>for</strong> low- and<br />

intermediate-level radioactive waste,<br />

stationary facilities <strong>for</strong> the con ditioning<br />

of low- and intermediate-level<br />

radioactive waste, disposal facilities<br />

<strong>for</strong> radioactive waste in Germany<br />

/L-12/. These studies are suitable in<br />

themselves <strong>for</strong> determining release<br />

Impact speed<br />

fractions of radioactive residues<br />

and wastes resulting from the load<br />

impacts of a plane crash.<br />

In the following, the term Konrad<br />

Transport Study is understood to<br />

include both the more detailed study<br />

and the verification. The calculation<br />

options of the Konrad Transport<br />

Study /L-9/ and the ESK statement<br />

on the stress test /L-12/ can be<br />

applied independently of each other<br />

and also lead to different release<br />

fractions.<br />

4.1 Konrad Transport Study<br />

In the Konrad Transport Study /L-9/,<br />

release fractions are determined<br />

experimentally <strong>for</strong> certain package<br />

and waste types and different load<br />

cases. In addition, release fractions<br />

are extended or generalized <strong>for</strong><br />

further package types and load cases<br />

on the basis of empirical and physical<br />

dependencies.<br />

In con<strong>for</strong>mity with the conditioning<br />

and packaging requirements<br />

<strong>for</strong> the Konrad repository in Germany,<br />

in /L-9/ accident loads of transport<br />

vehicles and transport casks are<br />

­classified in nine severity categories<br />

(= Belastungsklassen (BK)) and possible<br />

waste package combinations in<br />

eight waste package groups (= Abfallgebindegruppen<br />

(AGG)), adequately<br />

covering the range of possible package<br />

and load case combinations. Solely<br />

containers approved <strong>for</strong> transport and<br />

storage in the Konrad final repository<br />

are taken into account, and solely load<br />

cases which occur on the basis of the<br />

specified modes of transport are<br />

considered (transport by road and<br />

rail). In the severity categories, both<br />

mechanical and thermal loads are<br />

taken into account <strong>for</strong> the release.<br />

The severity categories are listed in<br />

the following Table 1.<br />

The classification of the mechanical<br />

loads according to the Konrad<br />

Transport Study /L-9/ goes to the<br />

maximum of a load equivalent of an<br />

impact of the package at a speed of<br />

110 km/h. The debris resulting from a<br />

plane crash will have substantially<br />

higher speeds, well in excess of<br />

Duration and temperature of fire<br />

No fire 30 min at 800 °C 60 min at 800 °C<br />

0 – 35 km/h BK 1 BK 2 BK 3<br />

36 – 80 km/h BK 4 BK 5 BK 6<br />

80 – 110 km/h BK 7 BK 8 BK 9<br />

| Tab. 1.<br />

Severity categories used in the Konrad Transport Study.<br />

110 km/h, and larger masses. The<br />

­severity category definitions initially<br />

are inadequate <strong>for</strong> analyzing the load<br />

impacts resulting from a plane crash.<br />

The high energy input in the packages<br />

as a result of the plane crash there<strong>for</strong>e<br />

makes it necessary to extrapolate<br />

the mechanically induced release<br />

fractions <strong>for</strong> the non-respirable particles<br />

with an aerodynamic equivalent<br />

diameter (AED) of > 10 µm. Extrapolation<br />

of the release fractions is<br />

­per<strong>for</strong>med using the specific energy<br />

input into the packages during impact.<br />

The extrapolation <strong>for</strong>mula <strong>for</strong><br />

the release fractions is:<br />

Here f 110 and f 80 represent the release<br />

fractions of the corresponding waste<br />

package groups at 80 km/h and<br />

110 km/h, respectively, and v the<br />

velocity of the impacting piece of<br />

debris.<br />

Since the release fractions in the<br />

Konrad Transport Study /L-9/ always<br />

refer to packages with a mass of<br />

11 tons and a volume of 7.4 m 3 , the<br />

release fractions <strong>for</strong> package types<br />

with a different mass and volume<br />

must be adjusted. The following scales<br />

are to be applied:<br />

For AGG 5 and 7:<br />

For AGG 1, 2, 3, 4 and 6:<br />

where m package mass<br />

V package volume<br />

f release fraction of AGG<br />

<strong>for</strong> the severity class (BK)<br />

f korr corrected release fraction<br />

For smaller volumes or masses the<br />

release fractions increase. The release<br />

fractions of the particles with AEDs<br />

from 0 to 10 µm and 10 to 100 µm,<br />

­derived from the two studies /L-9/<br />

and /L-10/, scaled, and extrapolated<br />

<strong>for</strong> the plane crash, are conservatively<br />

added to determine the releases.<br />

Hence, from the individual release<br />

fractions, <strong>for</strong> every waste package<br />

a total release fraction can be put<br />

together from mechanical and<br />

thermal loads.<br />

ENVIRONMENT AND SAFETY 427<br />

Current Procedure <strong>for</strong> Determining Release Parameters <strong>for</strong> a Plane Crash on a <strong>Nuclear</strong> Facility in the Context of Accident Analyses<br />

Environment and Safety<br />

ı Steffen Böhlke and Henrik Niegoth


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

ENVIRONMENT AND SAFETY 428<br />

Package type Thermal impact Large-area mechanical impact Puncti<strong>for</strong>m mechanical impact<br />

Combustible<br />

raw waste<br />

| Tab. 2.<br />

Release fractions in accordance with the ESK stress test.<br />

Combustible<br />

conditioned<br />

waste<br />

Noncombustible<br />

waste<br />

Raw waste,<br />

of which 50%<br />

respirable<br />

Conditioned<br />

waste, of which<br />

10% respirable<br />

Raw waste,<br />

of which 50%<br />

respirable<br />

Conditioned<br />

waste, of which<br />

10% respirable<br />

Cast iron cask 2,00E-05 2,00E-05 2,00E-05 0 0 0 0<br />

Concrete cask 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03<br />

Konrad IV container 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03<br />

20-ft. container 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03<br />

200-l drums 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03<br />

4.2 ESK stress test<br />

To establish the release fractions <strong>for</strong> a<br />

plane crash, it is also possible to use<br />

the release fractions cited in the ESK<br />

statement on the stress test /L-12/.<br />

Typical damage patterns were defined<br />

by ESK <strong>for</strong> this purpose, covering<br />

all types of serious impacts on the<br />

facilities under investigation, including<br />

the waste packages they<br />

contain. In /L-12/ these typical<br />

damage patterns are assumed and the<br />

resistance of the facilities against<br />

these damage patterns is evaluated.<br />

In regard to the typical damage<br />

patterns of a pattern crash, the following<br />

types of impacts have to be considered:<br />

p Thermal impacts due to a longer<br />

lasting fire (escape of fuel and<br />

burnup in the area of the radioactive<br />

wastes)<br />

p Mechanical impacts on waste<br />

packages, taking into account the<br />

difference in energy input:<br />

p Puncti<strong>for</strong>m mechanical impact<br />

(impact of an engine shaft on<br />

packages)<br />

p Large-area mechanical impact<br />

(roof truss falls on packages)<br />

In /L-12/, <strong>for</strong> the scenarios of the<br />

possible impacts, release fractions are<br />

compiled <strong>for</strong> cast iron and concrete<br />

casks as well as Konrad containers,<br />

20-foot containers and 200-liter steel<br />

drums as functions of the degree of<br />

conditioning (raw waste, conditioned<br />

waste) and the combustibility of the<br />

inventory of the waste packages. The<br />

basis <strong>for</strong> /L-12/ is, inter alia, the<br />

­Konrad Transport Study /L-9/, which<br />

was already used <strong>for</strong> the mechanical<br />

impacts.<br />

The relevant release fractions are<br />

shown in the following Table 2:<br />

5 Determination<br />

of source term<br />

The activity released into the environment<br />

of the facility (= source term)<br />

is determined as the product of the<br />

­nuclide-specific activities of the<br />

affected waste packages and the<br />

respective release fractions of the<br />

mechanical-thermal impacts. Due<br />

account has to be taken of the<br />

number of affected waste packages<br />

which have been subjected to<br />

mechanical and thermal loading as<br />

a result of the plane crash.<br />

Scenario-specific restraining ­effects<br />

possibly also need to be considered.<br />

This can be, <strong>for</strong> example, debris of<br />

the aircraft or the concrete roof, which<br />

contribute in a certain way to retain<br />

the radioactive inventory in the<br />

environment of the facility.<br />

6 Dispersion and<br />

dose calculations<br />

To determine the exposure in the<br />

environment of the facility, dispersion<br />

and dose calculations are made.<br />

The definitive basis <strong>for</strong> calculations<br />

is the incident calculation bases<br />

(= Störfall­berechnungsgrundlagen<br />

(SBG)) /L-13/. The dose rate coefficients<br />

from the compilation of dose<br />

coefficients <strong>for</strong> external and internal<br />

radiation exposure – Part I and Part II<br />

– published in the Bundesanzeiger<br />

(German Federal Government’s<br />

­Bulletin) of 23 July 2001 /L-14/, are<br />

used <strong>for</strong> the calculations.<br />

| Fig. 1.<br />

Sample-dose-distributions in the environment of a potential local release.<br />

Environment and Safety<br />

Current Procedure <strong>for</strong> Determining Release Parameters <strong>for</strong> a Plane Crash on a <strong>Nuclear</strong> Facility in the Context of Accident Analyses<br />

ı Steffen Böhlke and Henrik Niegoth


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

As result, the effective doses in a<br />

calendar year <strong>for</strong> the different age<br />

groups and different exposure pathways<br />

are determined at the most<br />

unfavorable point of exposure in<br />

the environment of the facility. The<br />

extent to which the intervention<br />

reference level of 100 mSv assessed<br />

<strong>for</strong> the plane crash is exhausted is<br />

established and evaluated according<br />

to /L-7/.<br />

A graphic description of a sample<br />

dose distribution in the environment<br />

of a potential local release is provided<br />

by the following illustrations (see<br />

Figure 1).<br />

7 Summary<br />

By means of the<br />

p scenario definition of the beyond-­<br />

design-basis event “plane crash”,<br />

p the determination of nuclide<br />

vectors and activity inventories,<br />

and<br />

p the determination of release<br />

fractions and source terms,<br />

described in this article, a standard<br />

procedure approved by authorities<br />

and the consulted experts has been<br />

established which enables committed<br />

doses following a plane crash on<br />

existing facilities <strong>for</strong> the dismantling<br />

of nuclear facilities, or such facilities<br />

still to be built, to be determined<br />

in a comprehensible and verifiable<br />

manner, without incurring approval<br />

risks.<br />

The described methods also can<br />

be adapted to determine committed<br />

doses of other less intensive scenarios<br />

involving smaller load impacts<br />

(e.g. the falling of packages, or<br />

earthquakes).<br />

References<br />

/L-1/<br />

/L-2/<br />

/L-3/<br />

/L-4/<br />

/L-5/<br />

/L-6/<br />

/L-7/<br />

/L-8/<br />

/L-9/<br />

Verordnung zum Schutz vor der schädlichen Wirkung<br />

ionisierender Strahlung (Strahlenschutzverordnung –<br />

StrlSchV) vom 29. November 2018 (BGBl. I S. 2034, 2036),<br />

die durch Artikel 1 der Verordnung vom 27. März <strong>2020</strong><br />

(BGBl. I S. 748) geändert worden ist.<br />

Gesetz zum Schutz vor der schädlichen Wirkung<br />

ionisierender Strahlung (Strahlenschutzgesetz – StrlSchG)<br />

vom 27. Juni 2017 (BGBl. I S. 1966), das zuletzt durch<br />

Artikel 248 der Verordnung vom 19. Juni <strong>2020</strong> (BGBl. I S.<br />

1328) geändert worden ist<br />

Leitlinien zur Beurteilung der Auslegung von Kernkraftwerken<br />

mit Druckwasserreaktoren gegen Störfälle im<br />

Sinne des § 28 Abs. 3 der Strahlenschutzverordnung –<br />

Störfall-Leitlinien - vom 18. Oktober 1983<br />

ESK-Empfehlung vom 10.06.2013: „ESK-Leitlinien für die<br />

Zwischenlagerung von radioaktiven Abfällen mit<br />

vernachlässigbarer Wärmeentwicklung“<br />

ESK-Empfehlung vom 10.06.2013: „Leitlinien für die<br />

trockene Zwischenlagerung bestrahlter Brennelemente<br />

und Wärme entwickelnder radioaktiver Abfälle in<br />

Behältern“<br />

ESK-Empfehlung vom 16.03.2015: „Leitlinien zur<br />

Stilllegung kerntechnischer Anlagen“<br />

Verordnung zur Festlegung von Dosiswerten für frühe<br />

Notfallschutzmaßnahmen (Notfall-Dosiswerte-Verordnung<br />

– NDWV) (SSK) vom 29. November 2018 (BGBl. I S. 2034,<br />

2172)<br />

RSK-Leitlinien für Druckwasserreaktoren,<br />

Ursprungsfassung (3. Ausgabe vom 14. Oktober 1981)<br />

mit Änderungen vom 15.11.1996<br />

Transportstudie Konrad, Sicherheitsanalyse zur<br />

Beförderung radioaktiver Abfälle zum Endlager Konrad,<br />

Gesellschaft für Anlagen- und Reaktorsicherheit mbH<br />

(GRS), 20<strong>09</strong>, mit Corrigendum vom April 2010<br />

/L-10/ GRS: Vertiefung und Ergänzung ausgewählter Aspekte der<br />

Abfalltransportrisiko-analyse für die Standortregion der<br />

Schachtanlage Konrad, Abschlussbericht zum Vorhaben<br />

3607R02600, Arbeitspaket 1, Teilaufgaben 11-14, Bericht<br />

GRS-A-3684 vom Februar 2013<br />

/L-11/ Richter, C., Forell, B., Sentuc, F.-N.: Überprüfung des<br />

unfallbedingten Freisetzungsverhaltens bei der<br />

Beförderung radioaktiver Stoffe, Abschlussbericht,<br />

Arbeitspaket 3, GRS - 482, Oktober 2017<br />

/L-12/ ESK-Stellungnahme vom 11.07.2013: „ESK-Stresstest<br />

für Anlagen und Einrichtungen der Ver- und Entsorgung<br />

in Deutschland, Teil 2: Lager für schwach- und<br />

mittelradioaktive Abfälle, stationäre Einrichtungen zur<br />

Konditionierung schwach- und mittelradioaktiver Abfälle,<br />

Endlager für radioaktive Abfälle“<br />

/L-13/ BMU: Störfallberechnungsgrundlagen, Stand Oktober<br />

1983, mit Kapitel 4 “Berechnung der Strahlenexposition”<br />

gemäß § 49 StrlSchV, Stand September 2003<br />

/L-14/ Bekanntmachung der Dosiskoeffizienten zur Berechnung<br />

der Strahlenexposition vom 23. Juli 2001, Bundesanzeiger<br />

Verlag<br />

Authors<br />

Dr. Steffen Böhlke<br />

Head of Licensing & <strong>Nuclear</strong><br />

Calculation<br />

steffen.boehlke@steag.com<br />

ENVIRONMENT AND SAFETY 429<br />

Henrik Niegoth<br />

Licensing & <strong>Nuclear</strong> Calculations<br />

STEAG Energy Services GmbH<br />

Rüttenscheider Str. 1-3<br />

45128 Essen, Germany<br />

Current Procedure <strong>for</strong> Determining Release Parameters <strong>for</strong> a Plane Crash on a <strong>Nuclear</strong> Facility in the Context of Accident Analyses<br />

Environment and Safety<br />

ı Steffen Böhlke and Henrik Niegoth


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

430<br />

RESEARCH AND INNOVATION<br />

Evaluation of MACST Strategies <strong>for</strong><br />

Extended Loss of AC Electric <strong>Power</strong><br />

Event in OPR1000 <strong>Nuclear</strong> <strong>Power</strong> Plants<br />

Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim<br />

1 Introduction In the nuclear power plant (NPP), station black out (SBO) is initiated by simultaneous loss<br />

of offsite power (LOOP) and operational failure of both emergency diesel generators (EDGs). In such cases, the primary<br />

operator action is required to recover alternative alternating current (AAC) power by manually operating an AAC diesel<br />

generator (AAC DG). If the AAC DG is also unavailable, the plant remains inoperable from all AC power recovery over<br />

the long term. After all the extended loss of AC power (ELAP) event occurs.<br />

The ELAP event has been considered<br />

as one of the consequent accidents<br />

initiated by natural disasters. In this<br />

event, only active systems powered by<br />

direct current (DC) from batteries and<br />

the passive accident coping measures<br />

such as turbine driven auxiliary feed<br />

water pumps (TD-AFWPs) play an<br />

essential role. So, the functions,<br />

operational range and capacity of the<br />

passive systems could be significant<br />

factors during the event [1]. The<br />

­response strategy of the ELAP event<br />

can thus be very limited.<br />

The component cooling water<br />

system (CCWS) and charging pumps<br />

providing seal injection water to the<br />

reactor coolant pumps (RCPs) are also<br />

not available during the ELAP event.<br />

Stoppage of seal injection flow can<br />

cause the inflow of coolant into<br />

the seal cartridge, which results in<br />

exposing the RCP seals at a high<br />

temperature. Maintaining the RCP<br />

seals at a high temperature can<br />

degrade seal materials and increase<br />

leak rates. In such conditions, the<br />

continuous loss of the inventory could<br />

occur in the reactor coolant system<br />

(RCS) through a RCP seal leakage and<br />

it could bring about core uncovery<br />

and damage.<br />

Early cooldown and depressurization<br />

of the RCS could minimize the<br />

inventory loss through the RCP seals<br />

and provide rapid RCS make-up<br />

through the injection of borated water<br />

from the safety injection tanks (SITs).<br />

According to the coping strategy of<br />

the ELAP event, RCS cooldown and<br />

depressurization are conducted via<br />

main steam release to the atmosphere<br />

through the main steam safety valves<br />

(MSSVs) and/or atmospheric dump<br />

valves (ADVs) of the main steam<br />

supply system. The RCS conditions<br />

will decrease to the actuating pressure<br />

of the SITs and then reach the entry<br />

| Fig. 1.<br />

OPR1000 NPP nodalization.<br />

Research and Innovation<br />

Evaluation of MACST Strategies <strong>for</strong> Extended Loss of AC Electric <strong>Power</strong> Event in OPR1000 <strong>Nuclear</strong> <strong>Power</strong> Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Event no. Event Set value<br />

10<br />

1 Initiating event SBO<br />

2 Reactor trip<br />

3 RCP trip<br />

4 Turbine trip<br />

5 MSSVs open SG pressure: 8.618 MPa<br />

6 Fail to operate AAC DG ELAP<br />

7 Start of a TD-AFWPs operation SG level: 23.6 %<br />

8 Completing load shedding<br />

9 ADVs open Cooling rate: 50 °F/hr<br />

| Tab. 1.<br />

Sequence of ELAP event.<br />

Completing connection<br />

of a 1 MW mobile generator<br />

condition of the shutdown cooling<br />

system (SCS). Available AFWPs will<br />

continuously supply water flow to<br />

steam generators (SGs) to make-up<br />

<strong>for</strong> steam release.<br />

Operators will regulate ADVs to<br />

control the amount of the steam<br />

release and the RCS cooldown rate as<br />

necessary. Because all AC power is not<br />

available, ADVs should be operated<br />

by local manual control during the<br />

ELAP event. If the ADV local control is<br />

not available, the RCS will not be<br />

depressurized and reach the injection<br />

pressure of SITs. In such cases, the<br />

back-up strategy should be necessary<br />

to compensate <strong>for</strong> RCS inventory loss<br />

to avoid core uncovery and damage,<br />

because installed equipment has<br />

­certain limitations under the ELAP<br />

conditions.<br />

In this study, we present effects of<br />

the RCP seal leakage on the safety<br />

capabilities of the NPP during the<br />

ELAP event. A comparative study is<br />

conducted according to whether the<br />

RCP seal leakage occurs. We also<br />

examine the feasibility of the back-up<br />

strategy using mobile facilities by<br />

­assuming the worst in the ELAP event.<br />

The target plants are the OPR1000<br />

NPP and other conditions <strong>for</strong> the<br />

target scenario are adopted from<br />

coping strategies presented in the<br />

Stress Test Report [2]. To analyze the<br />

thermal hydrodynamic behavior of<br />

the plant, RELAP5 Mod 3.3 is used<br />

[3]. The nodalization diagram of the<br />

OPR1000 NPP is shown in Figure 1.<br />

2 Effects of RCS seal<br />

leakage<br />

This section aims to provide effects of<br />

the RCS seal leakage on the NPP<br />

­systems at an early stage in the ELAP<br />

event. The comparative case study is<br />

per<strong>for</strong>med <strong>for</strong> two different cases<br />

whether the RCP seal leakage occurs.<br />

Table 1 shows a sequence of the ELAP<br />

event until completion of connecting a<br />

1 MW mobile generator to cope with<br />

total loss of AC power according to the<br />

stress test guideline of <strong>Nuclear</strong> Safety<br />

and Security Commission (NSSC) [4].<br />

A RCP seal leakage coincides with<br />

loss of all AC power, leading to<br />

­stoppage of seal injection flow to<br />

RCPs from charging pumps. The initial<br />

seal leak rate is assumed to be 25 gpm<br />

(1.58 l/s) per one RCP [5]. The<br />

RCP seal leaks will be completely<br />

stopped by activating a charging<br />

pump supplying seal injection flow<br />

­after recovering AC power by connecting<br />

of the 1 MW mobile generator.<br />

All conditions <strong>for</strong> the analysis are<br />

| Fig. 2.<br />

Pressurizer pressures.<br />

adopted from coping strategies<br />

presented in the Ref. [2].<br />

Two transient cases are analyzed in<br />

this section. For Case 1, we consider<br />

an initial seal leakage rate of 25 gpm<br />

(1.58 l/s) per one RCP. Case 2 assumes<br />

that a RCP seal leakage does<br />

not occur during the ELAP event.<br />

Figures 2 and 3 compare pressure and<br />

water level changes of the pressurizer,<br />

respectively. Transient behavior of<br />

two cases shows obvious differences<br />

depending on the RCP seal leakage. In<br />

Case 1, the pressurizer pressure and<br />

water level are rapidly reduced from<br />

the beginning of the event due to the<br />

continuous loss of the RCS inventory<br />

by the seal leakage. The water level of<br />

the pressurizer in Case 1 is completely<br />

empty within one hour. On the other<br />

hand, the pressure and water level of<br />

the pressurizer in Case 2 are initially<br />

maintained and start to decrease after<br />

ADVs open. The water level of the<br />

pressurizer is totally exhausted after<br />

three hours.<br />

Figure 4 shows temperature<br />

changes of hot legs <strong>for</strong> both cases. The<br />

results of Case 1 and Case 2 are similar<br />

in overall trend. For the first two<br />

hours, hot legs temperature is maintained,<br />

but starts to decrease after the<br />

plant cooling begins. The cooling rate<br />

RESEARCH AND INNOVATION 431<br />

| Fig. 3.<br />

Pressurizer water levels.<br />

| Fig. 4.<br />

Hot legs temperatures.<br />

Research and Innovation<br />

Evaluation of MACST Strategies <strong>for</strong> Extended Loss of AC Electric <strong>Power</strong> Event in OPR1000 <strong>Nuclear</strong> <strong>Power</strong> Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

RESEARCH AND INNOVATION 432<br />

| Fig. 5.<br />

RCP seal leak rates.<br />

| Fig. 6.<br />

Core collapsed water levels.<br />

| Fig. 7.<br />

Steam generator pressures.<br />

| Fig. 8.<br />

Pressurizer pressure (Phase 1).<br />

is 50 oF/hr (27.8 oC/hr) as described<br />

in Table 1. Variations of the RCP seal<br />

leak rate <strong>for</strong> one RCP are shown in<br />

Figure 5. In Case 1, the seal leak rate<br />

continues to decrease as pressure of<br />

the primary system is reduced. On the<br />

contrary, the seal leakage does not<br />

occur in Case 2.<br />

The changes of core collapsed<br />

water level <strong>for</strong> both cases are depicted<br />

in Figure 6. In Case 1, the abrupt drop<br />

of the core collapsed water level is<br />

observed at the early stage due to the<br />

continuous loss of the inventory<br />

through RCP seals. On the other hand,<br />

the core collapsed water level of<br />

Case 2 starts to decrease three hours<br />

later than that of Case 1. In both cases,<br />

however, the core collapsed water<br />

level does not reach the top of the<br />

active core. Thus, core degradation<br />

and melting due to the core uncovery<br />

do not occur during the event. When<br />

the pressurizer pressure decreases to<br />

the set point of the SIT actuation, the<br />

core collapsed water level starts to<br />

recover as shown in Figure 6.<br />

The pressure behavior of SGs<br />

during the event period shows no<br />

meaningful differences between two<br />

cases as shown in Figure 7. In both<br />

cases, the pressure of SGs is maintained<br />

during the first two hours by<br />

opening MSSVs and starts to decrease<br />

after ADVs open.<br />

3 MACST strategy<br />

In previous section, we paid attention<br />

to an effect of the RCP seal leakage.<br />

The RCP seal leakage had a great<br />

effect on the primary pressure and<br />

Phase no.<br />

Phase 1<br />

(0 ~ 8 hr)<br />

Phase 2<br />

(8 ~ 72 hr)<br />

Phase 3<br />

(72 hr ~ )<br />

| Tab. 2.<br />

MACST Strategy.<br />

Coping strategy<br />

Installed equipment<br />

MACST facilities<br />

supplementing installed<br />

equipment<br />

All on/off-site<br />

equipment<br />

inventory conditions. The abrupt drop<br />

of the core collapsed water level was<br />

observed from the early stage due to<br />

rapid depletion of the pressurizer<br />

­water level. We identified that depressurization<br />

of the RCS can reduce the<br />

inventory loss through the RCP seals<br />

and provide the injection of borated<br />

water from the SITs at an early stage.<br />

If the operation of the ADVs is failed,<br />

the RCS will not be depressurized and<br />

reach the injection pressure of SITs.<br />

The U.S. <strong>Nuclear</strong> Regulatory<br />

Committee (NRC) has developed<br />

­flexible mitigation (FLEX) strategy<br />

<strong>for</strong> increasing defense-in-depth <strong>for</strong><br />

beyond-design-basis events such as<br />

the ELAP and loss of ultimate heat<br />

sink (LUHS) events [6-10]. The<br />

OPR1000 NPPs have also been introduced<br />

the multi-barrier accident<br />

­coping strategy (MACST) to prevent<br />

the severe consequences from<br />

such events. Coping strategies are<br />

established to keep the pressure<br />

boundaries of the RCS, appropriately<br />

cool down the reactor core <strong>for</strong><br />

avoiding fuel damage, and maintain<br />

the integrity of the containment<br />

building in the transient conditions<br />

with the events of loss of safety<br />

functions. There are three phases<br />

depending on coping strategies as<br />

shown in Table 2.<br />

In Phase 1, the plant copes with<br />

only the installed equipment, such as<br />

DC batteries and natural circulation<br />

cooldown through SGs by the available<br />

AFWPs up to 8 hours. And then<br />

the cooldown and depressurization of<br />

the RCS are per<strong>for</strong>med by MSSVs<br />

and/or ADVs of the main steam supply<br />

system. In Phase 2, the MACST<br />

facilities are completely connected<br />

and supplement functions of installed<br />

Research and Innovation<br />

Evaluation of MACST Strategies <strong>for</strong> Extended Loss of AC Electric <strong>Power</strong> Event in OPR1000 <strong>Nuclear</strong> <strong>Power</strong> Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Phase Time Event<br />

Phase 1<br />

Phase 2<br />

0 sec Initiating event<br />

0 sec Reactor trip<br />

0 sec RCP trip<br />

0 sec Turbine trip<br />

0 sec Start of RCP seal leakages<br />

8 sec MSSVs open<br />

10 min Fail to operate AAC DG<br />

17 min Start of AFWPs operation<br />

30 min Completing load shedding<br />

2 hr Failure of RCS cooldown using ADVs<br />

8 hr Completing connection of HP mobile pump<br />

8 hr 15 min Activating HP mobile pump<br />

8 hr 30 min Preparing connection of 3.2 MW mobile generator<br />

72 hr Completing connection of 3.2 MW mobile generator<br />

| Tab. 3.<br />

Sequence of ELAP event during Phase 1 and Phase 2.<br />

equipment <strong>for</strong> supplying AC power to<br />

some safety systems or alternative<br />

RCS make-up to avoid core uncovery<br />

and damage. From 72 hours after the<br />

event initiation, the plant enters the<br />

final phase with maintaining Phase 2<br />

strategy and all available on/offsite<br />

facilities are in service.<br />

As stated above, one of the main<br />

concerns of the coping strategy of<br />

the ELAP event is the RCS make-up<br />

strategy due to the continuous inventory<br />

loss by the RCP seal leakage. In<br />

the MACST <strong>for</strong> the ELAP event, early<br />

cooldown and depressurization of the<br />

RCS should be conducted in Phase 1<br />

<strong>for</strong> reducing the inventory loss<br />

and providing rapid borated water<br />

injection from the SITs as described in<br />

Section 2. Operators locally control<br />

ADVs <strong>for</strong> depressurizing the primary<br />

system by releasing the steam.<br />

If the ADV local control and connecting<br />

the mobile generator are<br />

failed in both Phase 1 and 2, the RCS<br />

inventory will continuously decrease<br />

and the RCS make-up by SITs and<br />

CPs will not be available. There<strong>for</strong>e,<br />

the back-up strategy <strong>for</strong> the RCS<br />

make-up should be necessary to avoid<br />

core uncovery and damage. The highpressure<br />

(HP) mobile pump, with a<br />

nominal flow rate of 40 gpm at<br />

1500 psig, will be equipped <strong>for</strong> primary<br />

inventory make-up and boration<br />

during Phase 2 in case of a failure<br />

of the RCS depressurization, SIT<br />

injection, and CP operation.<br />

4 Feasibility study<br />

on high-pressure<br />

mobile pump <strong>for</strong> MACST<br />

strategy<br />

In this section, we assume a failure of<br />

ADV local control and connection<br />

of the mobile generator to create a<br />

situation using the HP mobile pump<br />

<strong>for</strong> RCS make-up during Phase 2. It is<br />

assumed that the HP mobile pump is<br />

successfully connected at 8 hours<br />

after initiating the event and starts<br />

operation in 15 minutes after the<br />

connection. Table 3 shows a major<br />

­sequence of the ELAP event in Phase 1<br />

and Phase 2 to comply with the<br />

MACST strategy.<br />

Transient behavior of pressurizer<br />

pressure during Phase 1 is shown in<br />

Figure 8. The pressurizer pressure decreases<br />

from the beginning of the<br />

event due to the continuous leakage of<br />

the RCS inventory through the RCP<br />

seals, and then is maintained approximately<br />

9.0 MPa (1290 psig) without<br />

opening ADVs at the end of Phase 1.<br />

The pressure of the primary system<br />

does not sufficiently decrease to the<br />

set point of the SIT actuation, so<br />

borated water is not injected to the<br />

RCS.<br />

The RCP seal leak rate <strong>for</strong> four<br />

RCPs during Phase 1 are shown in<br />

Figure 9. A tendency of the leak rate<br />

coincides with transient behavior of<br />

pressurizer pressure. The leak rate<br />

starts to oscillate between 6 and 9<br />

gpm after 7 hours because the upstream<br />

of the RCP seal becomes two<br />

phases and the core level is reduced to<br />

the cold leg elevation.<br />

The level in the reactor core is<br />

shown in Figure 10. The core<br />

collapsed water level starts to decrease<br />

from 2.5 hours after initiating the<br />

event due to the continuous loss of the<br />

RCS inventory. Although the core<br />

uncovery does not occur during<br />

Phase 1, the continuous loss of the<br />

inventory occur without any RCS<br />

make-up action. At the end of Phase 1<br />

which is the time of connecting and<br />

activating the HP mobile pump, RCS<br />

pressure is kept at 9.0 MPa (1290 psig)<br />

and RCP seal leak rates show an<br />

oscillatory behavior with an average<br />

of 7 gpm. These conditions are consistent<br />

with the design specifications<br />

of the HP mobile pump.<br />

To control reactivity and make up<br />

RCS inventory, Phase 2 strategy using<br />

MACST facilities should be per<strong>for</strong>med<br />

within 8 hours after the event. Figure<br />

11 shows transient behavior of<br />

pressurizer pressure during Phase 1<br />

and Phase 2. In Phase 2, pressurizer<br />

pressure is kept at approximately<br />

RESEARCH AND INNOVATION 433<br />

| Fig. 9.<br />

RCP seal leak rate (Phase 1).<br />

| Fig. 10.<br />

Core collapsed water level (Phase 1).<br />

Research and Innovation<br />

Evaluation of MACST Strategies <strong>for</strong> Extended Loss of AC Electric <strong>Power</strong> Event in OPR1000 <strong>Nuclear</strong> <strong>Power</strong> Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

RESEARCH AND INNOVATION 434<br />

| Fig. 11.<br />

Pressurizer pressure (Phase 1 + Phase 2).<br />

9.2 MPa (1320 psig) and the HP mobile<br />

pump continues to function. The<br />

core collapsed water level are shown<br />

in Figure 12. Since the charging rate<br />

(approximately 2.51 kg/s) <strong>for</strong> RCS<br />

inventory make-up is larger than the<br />

rate of inventory loss (approximately<br />

2.40 kg/s) through RCP seals,<br />

the inventory of RCS is gradually<br />

re covered. When taking into consideration<br />

that the initial flow rate of<br />

the RCP seal leak is conservatively<br />

estimated at 25 gpm, the HP mobile<br />

pump can give the entire satisfaction<br />

as the back-up strategy <strong>for</strong> RCS<br />

make-up.<br />

5 Conclusion<br />

In this study, we particularly paid<br />

attention to an effect of the RCP seal<br />

leakage causing the continuous loss<br />

of the RCS inventory on the safety capabilities<br />

of the NPP. Two different<br />

cases depending on consideration of<br />

the RCP seal leakage were compared.<br />

The RCP seal leakage had a great<br />

effect on the primary pressure and<br />

inventory conditions. The primary<br />

pressure rapidly decreased at the<br />

beginning of the event when considering<br />

the RCP seal leakage. The<br />

abrupt drop of the core collapsed<br />

water level was also observed from<br />

the early stage due to rapid depletion<br />

of the pressurizer water level. However,<br />

the core uncovery did not occur<br />

and the core collapsed water level<br />

starts to recover after reaching the<br />

actuating pressure of the SITs by early<br />

cooldown and depressurization of the<br />

RCS. The secondary side conditions<br />

also presented no meaningful differences<br />

between two cases when<br />

assuming the leak rate of 25 gpm<br />

(1.58 l/s) per one RCP.<br />

This study also examined the<br />

­feasibility of the MACST strategy <strong>for</strong><br />

inventory make-up using the HP<br />

mobile pump having a capacity of<br />

40 gpm at 1500 psig in Phase 2 of the<br />

ELAP event. To verify the strategy, we<br />

assumed failure of RCS cooldown and<br />

depressurization by local control of<br />

ADVs in Phase 1 and a subsequent<br />

failure of the 1 MW mobile generator<br />

in Phase 2. It is confirmed that<br />

the RCS pressure could be kept at<br />

maximum 9.2 MPa (1320 psig) and<br />

the seal leak rate <strong>for</strong> one RCP was<br />

­approximately 7 gpm at the end of<br />

Phase 1 through the simulation with<br />

RELAP5 code. It is concluded that<br />

these RCS conditions were sufficient<br />

to provide RCS make-up and boration<br />

using the HP mobile pump in Phase 2<br />

and the core water level could be fully<br />

recovered be<strong>for</strong>e entering Phase 3.<br />

Acknowledgments<br />

This work was supported by the Korea<br />

Institute of Energy Technology Evaluation<br />

and Planning (KETEP) and the<br />

Ministry of Trade, Industry & Energy<br />

(MOTIE) of the Republic of Korea<br />

(No. 20161510101840).<br />

References<br />

1. S. W. Lee et. al, “Extended Station Black Out Coping<br />

Capabilities of APR1400”, Science and Technology of <strong>Nuclear</strong><br />

Installations, Vol. 2014, p. 1-10, 2014.<br />

2. Korea Hydro and <strong>Nuclear</strong> <strong>Power</strong> Co. Ltd., “Stress Test Report<br />

<strong>for</strong> Hanul Unit 3&4”, 2017.<br />

3. In<strong>for</strong>mation Systems Laboratories, Inc., “RELAP5/MOD3.3<br />

Code Manual”, 2016.<br />

4. <strong>Nuclear</strong> Safety and Security Commission, “Stress Test <strong>for</strong><br />

<strong>Nuclear</strong> <strong>Power</strong> Plants in Long Term Operation”, 2013.<br />

5. J. Hartz et. al, “WCAP-17601-P: Reactor Coolant System<br />

Response to the Extended Loss of AC <strong>Power</strong> Event <strong>for</strong><br />

Westinghouse, Combustion Engineering and Babcock &<br />

Wilcox NSSS Designs, 2012.<br />

6. <strong>Nuclear</strong> Energy Institute, “Diverse and Flexible Coping<br />

Strategies (FLEX) Implementation Guide”, 2012.<br />

7. D. H. Kim et. Al, “Development of Mitigation Strategy <strong>for</strong><br />

Beyond Design Basis External Events <strong>for</strong> NRC Design<br />

Certification”, Transactions of the Korean <strong>Nuclear</strong> Society<br />

Autumn Meeting, 2013.<br />

8. United States <strong>Nuclear</strong> Regulatory Commission, “Byron Station,<br />

Units 1 and 2 – Safety Evaluation Regarding Implementation<br />

of Mitigating Strategies and Reliable Spent Fuel Pool<br />

Instrumentation Related to Orders EA-12-049 and EA-12-051”,<br />

2016.<br />

9. M. M. Rahman and M. B, Shohag, “FLEX Strategy to Cope<br />

with Extended SBO <strong>for</strong> APR1400”, <strong>International</strong> <strong>Journal</strong> of<br />

Engineering Research and Technology (IJERT), Vol. 5, Issue 10,<br />

2016.<br />

10. C. F. Huang et. al, “Analysis of a Postulated ELAP Event in<br />

Maanshan NPP using Trace Code”, TopSafe 2017, 2017.<br />

| Fig. 12.<br />

Core collapsed water level (Phase 1 + Phase 2).<br />

Authors<br />

Bongsik Chu<br />

bongsik.choo@khnp.co.kr<br />

Seyun Kim<br />

Junkyu Song<br />

Minjeong Kim<br />

Chang Hyun Kim<br />

Korea Hydro and <strong>Nuclear</strong> <strong>Power</strong><br />

Co., Ltd.,<br />

Central Research Institute<br />

70, Yuseong-daero<br />

1312-gil, Yuseong-gu<br />

Daejeon, 34101, Korea<br />

Research and Innovation<br />

Evaluation of MACST Strategies <strong>for</strong> Extended Loss of AC Electric <strong>Power</strong> Event in OPR1000 <strong>Nuclear</strong> <strong>Power</strong> Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Neutronic Study of CAREM-25<br />

Advanced Small Modular Reactor<br />

Using Monte Carlo Simulation<br />

Saeed Zare Ganjaroodi and Ali Pazirandeh<br />

1 Introduction The trend in development has been towards design certification of small modular reactors,<br />

which are defined as advanced reactors that produce electricity up to 300 MW(e), designed to be built in factories and<br />

shipped to utilities <strong>for</strong> installation as demand arises. These new factory-built designs aim to reduce lengthy con struction<br />

times while simultaneously increasing quality, thereby minimizing the financing costs associated with nowadays design<br />

projects that span 5–8 years. Small Modular Reactors designs include water-cooled reactors, high-temperature<br />

gas-cooled reactors, as well as liquid metal cooled reactors with fast neutron spectrum. Some of Small Modular Reactors<br />

are to be deployed as multiple-module power plants. Several countries are also pioneering the development and<br />

­application of transportable nuclear power plants, including floating and seabed-based Small Modular Reactors. The<br />

distinct concepts of operations, staffing and security requirements, size of emergency planning zones (EPZs), licensing<br />

process, legal and regulatory framework are the main issues <strong>for</strong> the Small Modular Reactors deployment. The projected<br />

timelines of readiness <strong>for</strong> deployment of SMRs generally range from the present to 2025–2030 [1-2].<br />

Central Argentina de Elementos<br />

­Modulares (CAREM-25) is a national<br />

SMR development project based on<br />

LWR technology coordinated by the<br />

Argentina National Atomic Energy<br />

Commission (CNEA) in collaboration<br />

with leading nuclear companies in<br />

­Argentina with the purpose to<br />

develop, design and construct innovative<br />

small nuclear power plants<br />

with high economic competitiveness<br />

and high level of safety. CAREM-25 is<br />

deployed as a prototype to validate<br />

the innovations <strong>for</strong> the future commercial<br />

version of CAREM that<br />

will generate an electric output of<br />

150-300 MW(e). CAREM-25 is an<br />

integral type PWR based on indirect<br />

steam cycle with distinctive features<br />

that simplify the design and support<br />

the objective of achieving a higher<br />

level of safety. Some of the design<br />

characteristics of CAREM-25 are<br />

integrated primary cooling system,<br />

in-vessel hydraulic control rod drive<br />

mechanisms and safety systems<br />

relying on passive features. Coolant<br />

flow in the primary reactor system<br />

is done by natural circulation.<br />

­CAREM-25 reactor was developed<br />

using domestic technology with at<br />

least 70 % of the components and<br />

­related services required by Argentine<br />

companies [1-3-4].<br />

Although several reports and<br />

essays on various technical aspects of<br />

CAREM-25 small modular reactor<br />

have been studied in recent years but<br />

their Neutronic behavior of the core<br />

in critical situation has not been<br />

discussed in details [3-4-5-6-7-8-9-10-<br />

11-12]. Also, due to the importance of<br />

calculation of the critical parameters<br />

in critical condition, some parameters<br />

such as neutron spectrum, power<br />

peaking factor associated with each<br />

fuel assemblies and the worth of<br />

­control rods are evaluated <strong>for</strong> the first<br />

time in this paper.<br />

In this work, MCNPX code is used<br />

<strong>for</strong> the neutronic simulation of the<br />

CAREM-25 reactor core. The reasons<br />

<strong>for</strong> using this code are different<br />

capabilities of the code to analyze the<br />

Neutronic calculation and to benchmark<br />

the results with references to<br />

show an appropriate consistency.<br />

MCNPX is a general purpose<br />

Monte Carlo radiation transport code<br />

designed to track many particle types<br />

over broad ranges of energies. It can<br />

be used in several transport modes:<br />

neutron only, photon only, electron<br />

only, combined neutron/photon<br />

transport where the photons are<br />

produced by neutron interactions,<br />

neutron/ photon/electron, photon/<br />

electron, or electron/photon [13].<br />

The MCNPX code is using Monte Carlo<br />

method to simulate the geometry and<br />

solve transport equation by tracing<br />

individual particles and recording<br />

some aspects (tallies) of their average<br />

behavior. It does not solve an explicit<br />

transport equation, but rather obtains<br />

answers by simulating individual<br />

particles and recording some tallies of<br />

individual particle average behavior.<br />

Then, the average behavior of the<br />

particles in the physical system is<br />

interpreted as the average behavior of<br />

the simulated particles [14].<br />

This code was used to create some<br />

inputs <strong>for</strong> the core of CAREM-25<br />

reactor core to calculate and analyze<br />

the Neutronic parameters such as<br />

effective multiplication factor, neutron<br />

flux distribution, axial power<br />

distribution, power peaking factor<br />

related to each fuel assemblies and<br />

the worth of control rods.<br />

2 Materials and Methods<br />

2.1 CAREM-25 Reactor<br />

CAREM-25 has been chosen as the<br />

reference small modular reactor in<br />

this study <strong>for</strong> Neutronic simulation.<br />

CAREM-25 is an integrated and<br />

self-pressurized reactor that has<br />

primary cooling by natural circulation.<br />

This reactor has some features<br />

that make the reactor extremely<br />

simple and also contribute to a higher<br />

level of safety. CAREM-25 core design<br />

data are shown in Table 1 [1].<br />

The CAREM-25 reactor pressure<br />

vessel (RPV) contains the core, the<br />

steam generators (SG), the whole<br />

primary coolant, and the absorber rod<br />

drive mechanisms. The RPV diameter<br />

is about 3.2 m and the overall length<br />

is about 11 m. CAREM-25 Reactor<br />

­Pressure Vessel is shown in Figure 1<br />

[1-4-5-12].<br />

The core consists of 61 hexagonal<br />

fuel assemblies having 1.4 active<br />

lengths. Each fuel assembly has 1<strong>08</strong><br />

fuel rods, 18 guide thimbles and 1<br />

instrumentation thimble (Figure 2).<br />

The fuel is enriched Uranium Oxide<br />

[1-15].<br />

CAREM-25 core do not use any<br />

chemical controller such as boric acid.<br />

Core reactivity is controlled by the use<br />

of 8 % gadolinium oxide (Gd 2 O 3 )<br />

mixed with 92 % uranium oxide as<br />

burnable poison in specific fuel rods<br />

in 42 fuel assemblies. Absorbing<br />

­Materials including Silver (80 %),<br />

Cadmium (5 %) and Indium (15 %)<br />

are used into the adjusting and safety<br />

control rods [15-16].<br />

RESEARCH AND INNOVATION 435<br />

Research and Innovation<br />

Neutronic Study of CAREM-25 Advanced Small Modular Reactor Using Monte Carlo Simulation ı Saeed Zare Ganjaroodi and Ali Pazirandeh


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

RESEARCH AND INNOVATION 436<br />

Parameter<br />

Technology developer<br />

Country of origin<br />

Reactor type<br />

Electrical capacity (MW) 27<br />

Value<br />

CNEA<br />

Thermal capacity (MW) 100<br />

Design life (year) 60<br />

Coolant/moderator<br />

Primary circulation<br />

Argentina<br />

System pressure (MPa) (Primary Cycle) 12.25<br />

System pressure (MPa) (Secondary Cycle) 4.7<br />

Main reactivity control mechanism<br />

RPV height (m) 11<br />

RPV diameter (m) 3.2<br />

Coolant temperature, core inlet (°C) 284<br />

Coolant temperature, core outlet (°C) 326<br />

<strong>Power</strong> conversion process<br />

Passive safety features<br />

Active safety features<br />

Fuel type/assembly array<br />

Fuel rod cladding material<br />

Absorbent pellet<br />

Control rod cladding material<br />

Fuel active length (m) 1.4<br />

Number of fuel assembly 61<br />

Fuel enrichment (%)<br />

Fuel burn-up (GWd/ton)<br />

Fuel cycle (month)<br />

Number of safety trains 2<br />

Emergency safety and Residual heat removal systems<br />

Modules per plant 1<br />

| Tab. 1.<br />

CAREM-25 design parameters.<br />

Integral PWR<br />

Light water<br />

Natural circulation<br />

Only by control rods<br />

Indirect Rankine cycle<br />

Yes<br />

Yes<br />

UO2 pellets/hexagonal<br />

Zry-4<br />

Ag-In-Cd<br />

AISI 316 L<br />

3.1 (prototype)<br />

24 (prototype)<br />

14 (prototype)<br />

Passive<br />

2.2 MCNPX Neutronic<br />

Simulation<br />

MCNPX2.6.0 code is used to model<br />

CAREM-25 core and calculate neutronic<br />

parameters such as effective<br />

multiplication factor, thermal, epithermal<br />

and fast neutron flux distribution,<br />

axial power distribution,<br />

power peaking factor related to each<br />

fuel assembly and the worth of<br />

adjusting control rods in steady state<br />

<strong>for</strong> fresh fuel. In this code, the KCODE<br />

card is used <strong>for</strong> critical source calculations,<br />

and 1 million particles with<br />

200 cycles were considered. The axial<br />

coolant temperature changes from<br />

284 (°C) in the core inlet to approximately<br />

326(°C) in the core outlet.<br />

In the calculation, the temperature<br />

of water is considered 305 (°C) as<br />

average temperature via using the<br />

ENDF/B-VI library and 42.C, 51.C,<br />

52.C and 70.C identification databases<br />

codes in MCNPX input file.<br />

3 Results and Discussion<br />

In this work, the neutronic evaluation<br />

of the CAREM- 25 advance small<br />

modular reactor is discussed using<br />

MCNPX code simulation. According to<br />

the calculation of Neutronic parameters<br />

in critical condition, in addition<br />

to analyzing the core in critical<br />

situation, some parameters including,<br />

neutron flux distributions, neutron<br />

spectrum, power distributions, power<br />

peaking factor related to each fuel<br />

assemblies and the worth of control<br />

rods have been discussed <strong>for</strong> the first<br />

time in this paper. It should be noted<br />

that simulation is done according to<br />

the design criteria in the latest reports<br />

from Argentina.<br />

| Fig. 1.<br />

CAREM-25 reactor pressure vessel and fuel assembly diagram.<br />

Control rods are classified into<br />

several groups as adjusting and safety<br />

control rods. Adjusting control rods are<br />

applied in 19 fuel assembly in ­specific<br />

rods. Safety rods are applied just into<br />

6 fuel assembly. Control ­systems are<br />

used to reactivity control during normal<br />

operation and to produce a sudden<br />

interruption of the nuclear chain reaction<br />

when required [11-12-15-16].<br />

3.1 Criticality calculations<br />

The effective multiplication factor<br />

(k eff ) is calculated 1.04576 with<br />

0.00028 percent error (excess reactivity<br />

is equal to 43.75 (mk)) in<br />

­MCNPX code when 10 % of the adjust<br />

control rods are into the core. Comparison<br />

of results shows an appropriate<br />

consistency with reference. The<br />

small difference in the results is due<br />

to the difference in the use of the<br />

percentage of gadolinium oxide in<br />

the fuel mixture in the simulation by<br />

MCNPX code.<br />

Given that the CAREM-25 reactor<br />

has not been constructed yet, Reactor<br />

design parameters according to the<br />

latest reports indicate that use of 8 %<br />

gadolinium oxide mixed with 92 %<br />

uranium oxide as burnable poison<br />

in specific fuel rods in 42 fuel<br />

assemblies. However, in some reports<br />

there has been a ratio of 7.5 % to<br />

Research and Innovation<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 2.<br />

The configuration of different types of fuel assemblies into the core.<br />

92.5 % <strong>for</strong> the amount of gadolinium<br />

oxide and fuel [17]. There<strong>for</strong>e, due to<br />

the high neutron absorption cross<br />

section of gadolinium, the greater<br />

amount used in this simulation<br />

­reduces the neutron flux and the<br />

effective multiplication factor consequently.<br />

Due to the lack of poisons ( 135 Xe<br />

and 149 Sm) in the beginning of the<br />

reactor operation <strong>for</strong> archiving<br />

criticality state, both adjust and safety<br />

control rods should be inserted into<br />

the reactor core. By inserting 50 % of<br />

adjusting control rods and 10 % of<br />

safety control rods into the core the<br />

reactor will achieve to critical condition.<br />

3.2 Axial and radial flux<br />

distributions<br />

The axial flux distribution in the<br />

­CAREM-25 small modular reactor<br />

core has been shown in Figure 3.<br />

It should be noted that the energy<br />

intervals are selected according to the<br />

reference from Argentina [19].<br />

The maximum thermal flux is<br />

­related about 57 (cm) from the height<br />

of the core. The inserting of the<br />

control rods into the core can move<br />

the maximum flux height of about<br />

40 % to the end. In Figures 4, 5 and 6<br />

the radial thermal, epithermal and<br />

fast flux distributions are shown in<br />

critical condition using cubic mesh in<br />

MCNPX code.<br />

As the figures show the neutron<br />

flux distribution in this reactor has a<br />

one-thirds symmetry. The reason of<br />

flux drop in some zones is the ­entrance<br />

of control rods and gadolinium oxide<br />

with the high absorption crosssection.<br />

Calculations show flux drop is in<br />

some zones that the control rods and<br />

gadolinium oxide with the high<br />

absorption cross-section are inserted<br />

into the core. Also, the figures<br />

illustrate that the thermal and total<br />

flux decrease due to the insertion of<br />

adjusting control rods (in 19 fuel<br />

assemblies) into the core during the<br />

| Fig. 4.<br />

The radial thermal flux distribution per one neutron in critical condition.<br />

| Fig. 5.<br />

The radial epithermal flux distribution per one neutron in critical condition.<br />

| Fig. 6.<br />

The radial fast flux distribution per one neutron in critical condition.<br />

| Fig. 3.<br />

Axial flux distribution in critical condition.<br />

expected operating transient. The<br />

­remarkable decreases of thermal flux<br />

have occurred in fuel assemblies,<br />

which includes control rods and<br />

RESEARCH AND INNOVATION 437<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

RESEARCH AND INNOVATION 438<br />

| Fig. 7.<br />

Neutron spectrum in the CAREM-25 reactor core.<br />

gadolinium oxide. The insertion of the<br />

adjusting control rod clusters into the<br />

core changes thermal flux peaks with<br />

respect to both of the place and the<br />

magnitude. By inserting 50 and<br />

100 percent of control rods into the<br />

core, the central maximum thermal<br />

flux decrease remarkably.<br />

The neutron spectrum in the<br />

­CAREM-25 small modular reactor<br />

core by MCNPX code has been shown<br />

in Figure 7.<br />

Figures 8 and 9 illustrate the effect<br />

of gadolinium oxide mixed with<br />

­uranium oxide as fuel on fission and<br />

absorption cross-section in the core.<br />

According to Figures 8 and 9<br />

gadolinium oxide with high absorption<br />

cross-section can effect on<br />

the fission rate in the core as a burnable<br />

poison. Gadolinium oxide mixed<br />

with uranium oxide as fuel can reduce<br />

the fission cross-section in order of ten<br />

times considerably.<br />

3.3 Worth of adjust control<br />

rods<br />

From the theoretical point of view, the<br />

integral worth of control rod is the<br />

total reactivity along the control rod.<br />

On the other hand, the differential<br />

worth is the reactivity in each unit of<br />

the length of the control rod. In Figure<br />

10, the integral and differential worth<br />

of adjusting control rods are plotted in<br />

of the core. According to the MCNPX<br />

code calculation, the total reactivity<br />

value of the adjusting control rods is<br />

almost 130.74 mk.<br />

| Fig. 8.<br />

UO 2 (3.1%) neutron cross-section.<br />

3.4 <strong>Power</strong> Peaking Factor<br />

Calculation<br />

<strong>Power</strong> Peaking Factor of each fuel<br />

assembly indicates a factor that determines<br />

the amount of power produced<br />

in each fuel assembly. This coefficient<br />

is defined as follows:<br />

| Fig. 9.<br />

UO 2 (3.1%) + Gd 2 O 3 neutron cross-section.<br />

| Fig. 10.<br />

Integral and differential worth of adjust control rods.<br />

The axial power and power peaking<br />

factor (PPF) distribution in the core by<br />

MCNPX code has been shown in<br />

Figures 11.<br />

The maximum power distribution<br />

in the core is related about 57 (cm)<br />

from the height of the core. The<br />

inserting of the control rods into the<br />

core can move the maximum axial<br />

power height of about 40 % to the<br />

end. It should be noted that axial flux<br />

and power distribution will reach the<br />

maximum value at the same axial<br />

height.<br />

Calculated <strong>Power</strong> Peaking Factor<br />

(PPF) of each fuel assembly by<br />

MCNPX code has been demonstrated<br />

in Figure 12.<br />

The power peaking factors of the<br />

FAs are calculated using the MCNPX<br />

code. Since the coolant temperature<br />

and density considerably affect the<br />

neutron moderation and fission<br />

cross-section of materials, the height<br />

of the core is divided into 10 equally<br />

spaced zones to calculate the power in<br />

each zone.<br />

The maximum calculated power<br />

peaking factor <strong>for</strong> fuel assembly is<br />

Research and Innovation<br />

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<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

| Fig. 11.<br />

Axial power distribution in the core.<br />

| Fig. 12.<br />

<strong>Power</strong> Peaking Factor (PPF) of fuel assemblies.<br />

calculated 1.526 in the hottest fuel<br />

assembly. The maximum power<br />

peaking factors are related to central<br />

fuel assemblies and some fuel assemblies<br />

consist of 3.1% enrichment<br />

fuel without any gadolinium oxide.<br />

4 Conclusions<br />

The purpose of this study is the<br />

Neutronic evaluation of the generation<br />

IV reactors CAREM-25 small<br />

modular reactor using Mote carlo<br />

simulation. Regarding some unique<br />

feathers, CAREM-25 is chosen as the<br />

reference small modular reactor in<br />

this study <strong>for</strong> simulation. CAREM-25<br />

is an integrated and self-pressurized<br />

reactor that has primary cooling by<br />

natural circulation. This reactor has<br />

some features that greatly make the<br />

reactor simply the reactor and also<br />

contribute to a higher level of safety.<br />

Considering the importance of calculation<br />

of the parameters in critical<br />

condition, and the reactor has not<br />

constructed yet, Neutronic simulation<br />

is done according to the latest design<br />

parameters from Argentina. Then, in<br />

addition to the Neutronic evaluation,<br />

some parameters such as neutron<br />

spectrum, power peaking factor<br />

related to each fuel assemblies and<br />

the worth of control rods discussed<br />

<strong>for</strong> the first time in this study. Results<br />

show that the reactor core has<br />

­approximately 43.75 (mk) excess<br />

­reactivity by inserting 10 % of adjusting<br />

control rods into the core in<br />

MCNPX code. The reason <strong>for</strong> this high<br />

excess reactivity of the core is the long<br />

cycle of this reactor over 14 months.<br />

The excess reactivity of fresh fuel<br />

is appropriately compensated using<br />

gadolinium oxide as a burnable poison<br />

mixed with uranium dioxide in some<br />

fuel rods in 24 fuel assemblies. Due to<br />

the lack of poisons ( 135 Xe and 149 Sm)<br />

in the beginning of the reactor<br />

operation <strong>for</strong> archiving criticality<br />

state, both adjust and safety control<br />

rods should be inserted into the<br />

­reactor core. By inserting 50 % of<br />

­adjusting control rods and 10 % of<br />

safety control rods into the core the<br />

CAREM-25 reactor core will reach to<br />

critical condition. The inserting of the<br />

control rods into the core can move<br />

the maximum flux height of about<br />

40% to the end. Also, near control<br />

rods and burnable poisons regions<br />

due to the absorbers materials, the<br />

neutron flux dropped sharply. It<br />

should be noted that the total<br />

reactivity value of the adjust control<br />

rods is almost 130.74 (mk). The<br />

maximum calculated axial power<br />

peaking factor <strong>for</strong> fuel assemblies is<br />

1.526 in the hottest fuel assembly. The<br />

maximum axial power peaking factor<br />

is related to central fuel assemblies<br />

and some fuel assemblies consist<br />

of 3.1 % enrichment fuel without<br />

gadolinium oxide.<br />

References<br />

[1] <strong>International</strong> Atomic Energy Agency, 2014. Advances in<br />

Small Modular Reactor Technology Developments, A<br />

Supplement to: IAEA Advanced Reactors In<strong>for</strong>mation System<br />

(ARIS). IAEA, Vienna.<br />

[2] <strong>International</strong> Atomic Energy Agency, 2006. Status of<br />

Innovative Small and Medium Sized Reactor Designs 2005<br />

Reactors with Conventional Refueling Schemes.<br />

IAEATECDOC-1485. IAEA, Vienna.<br />

[3] Ishida, M., 2000. Development of New <strong>Nuclear</strong> <strong>Power</strong> Plant<br />

in Argentina, Advisory Group Meeting on Optimizing<br />

Technology, Safety and Economics of Water Cooled Reactors<br />

(Vienna, Austria).<br />

[4] Ishida, M., et al., 2001. CAREM Project Development<br />

Activities”. <strong>International</strong> Seminar on Status and Prospects <strong>for</strong><br />

Small and Medium Size Reactors (Cairo, Egypt).<br />

[5] CNEA & INVAP, 2000. CAREM-25-in<strong>for</strong>me Consolidado.<br />

[6] Gomez, S., 2000. Development Activities on Advanced LWR<br />

Designs in Argentina, Technical Committee Meeting on<br />

Per<strong>for</strong>mance of Operating and Advanced Light Water Reactor<br />

Designs (Munich, Germany).<br />

[7] Delmastro, D., 2000. Thermal-hydraulic Aspects of CAREM<br />

Reactor, IAEA TCM on Natural Circulation Data and Methods<br />

<strong>for</strong> Innovative <strong>Nuclear</strong> <strong>Power</strong> Plant Design (Vienna, Austria).<br />

[8] Delmastro, D., Mazzi, R., Santecchia, A., Ishida, V., Gomez, S.,<br />

Gomez de Soler, S., Ramilo, L., 2002. CAREM: An Advanced<br />

Integrated PWR’. In IAEA, Small and Medium Sized Reactors:<br />

Status and Prospects. IAEA-CSP-14/P, pp. 224e231.<br />

[9] Mazzi, R., Santecchia, A., Ishida, V., Delmastro, D., Gomez, S.,<br />

Gomez de Soler, S., Ramilo, L., 2002. CAREM Project<br />

Development. In IAEA, Small and Medium Sized Reactors:<br />

Status and Prospects. IAEA-CSP-14/P, pp. 232e243.<br />

[10] Reyes, J., 2005. Integral System Experiment Scaling<br />

Methodology. Annex 11, Natural Circulation in Water Cooled<br />

<strong>Nuclear</strong> <strong>Power</strong> Plants Phenomena, Models, and<br />

Methodology <strong>for</strong> System Reliability Assessments.<br />

IAEA TECDOC 1474.<br />

[11] Boado Magan, H., Delmastro, D.F., Markiewicz, M., Lopasso,<br />

E., Diez, F., Gimenez, M., Rauschert, A., Halpert, S., Chocron,<br />

M., Dezzutti, J.C., Pirani, H., Balbi, C., Fittipaldi, A., Schlamp,<br />

M., Murmis, G.M., Lis, H., 2011. CAREM Project Status,<br />

Science and Technology of <strong>Nuclear</strong> Installations. Article ID<br />

140373.<br />

[12] Boado Magana, H., Delmastrob, D.F., Markiewiczb, M.,<br />

Lopassob, E., Diez, F., Gim_enezb, M., Rauschertb, A.,<br />

Halperta, S., Chocr_onc, M., Dezzuttic, J.C., Pirani, H.,<br />

Balbi, C., 2012. CAREM Prototype Construction and Licensing<br />

Status. IAEA-CN-164e5S01.<br />

[13] Pelowitz, D.B., 20<strong>08</strong>. MCNPXTM User’s Manual Version 2.6.0.<br />

LOS ALAMOS NATIONAL LABORATORY.<br />

[14] Liu, B., Lv, X., Zhao, W., Wang, K., Tu, J., Ouyang, X., 2010.<br />

The comparison of MCNP perturbation technique with MCNP<br />

difference method in critical calculation. Nucl. Eng. Des. 240,<br />

2005e2010.<br />

[15] Villarino, E., Hergenreder, D., Matzkin, S., 2012. Neutronic-<br />

Core Per<strong>for</strong>mance of CAREM-25 Reactor. INVAP, Argentina.<br />

[16] Diego Ferraro, 20<strong>09</strong>. Calculo de exposicion de estructuras<br />

interiores recipinte de presion del CAREM-25 mediante<br />

MCNP. Instituto Balseiro Universidad Nacional de Cuyo<br />

Comision Nacional de Energia Atomica. San Carlod de<br />

Bariloche Argentina.<br />

[17] S. Tashakor, E. Zarifi, M. Naminazari. 2017. Neutronic<br />

simulation of CAREM-25 small modular reactor. Progress in<br />

<strong>Nuclear</strong> Energy 99 (2017) 185e195.<br />

Authors<br />

Saeed Zare Ganjaroodi<br />

Ali Pazirandeh<br />

Islamic Azad University<br />

Engineering Science and Research<br />

Branch<br />

Shodada Hesarak blvd,<br />

Daneshgah Square,<br />

Sattari Highway,<br />

Tehran, Iran<br />

RESEARCH AND INNOVATION 439<br />

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Operating results 2019<br />

In 2019 the seven (7) German nuclear power plants generated 75.10<br />

billion kilowatt hours (kWh) of electricity gross. At the end of 2019<br />

the Philippsburg 2 nuclear power plant ceased commercial operation<br />

due to the revision of the German Atomic Energy Act in the political<br />

aftermath of the accidents in Fukushima, Japan, in 2011. Seven<br />

nuclear power plants with an electric gross output of 10,013 MWe<br />

were in operation during the year 2019.<br />

All seven nuclear power plants in operation in 2019 achieved<br />

results with a gross production greater than 10 billion kilowatt hours,<br />

one power plant, The Isar 2 unit even produced more than 12 billion<br />

kilowatt hours.<br />

Additionally the Isar 2 unit achieved one of the world’s ten best<br />

production results in 2019 (“Top Ten”, sixth place). At the end of<br />

2019, 449 reactor units were in operation in 31 countries worldwide<br />

and 54 were under construction in 16 countries. The share of nuclear<br />

power in world electricity production was around 11 %. German<br />

nuclear power plants have been occupying top spots in electricity<br />

production <strong>for</strong> decades thus providing an impressive demonstration<br />

of their efficiency, availability and reliability.<br />

The Taishan-1 nuclear power plant in China (capacity: 1,750 MWe<br />

gross, 1,660 MWe net, reactor type: EPR, the most powerful nuclear<br />

power plant worldwide and the most powerful single power plant<br />

worldwide) achieved the world record in electricity production in<br />

2019 with appr. 13 billion kilowatt hours.<br />

Worldwide, 41 nuclear power plant units achieved production<br />

­results of more than 10 billion kilowatt hours net in the year 2019.<br />

Additionally German nuclear power plants are leading with their<br />

lifetime electricity production. The Brokdorf, Emsland, Grohnde,<br />

Isar 2 and Philipsburg 2 nuclear power plant have produced more<br />

than 350 billion kilowatt hours since their first criticality.<br />

441<br />

REPORT<br />

Operating results of nuclear power plants in Germany 2018 and 2019<br />

<strong>Nuclear</strong> power plant Rated power Gross electricity<br />

generation<br />

in MWh<br />

Availability<br />

factor*<br />

in %<br />

Energy availability<br />

factor**<br />

in %<br />

gross<br />

in MWe<br />

net<br />

in MWe<br />

2018 2019 2018 2019 2018 2019<br />

Brokdorf KBR 1,480 1,410 10,375,751 10,153,213 90.60 87.69 84.72 82.34<br />

Emsland KKE 1,406 1,335 11,495,686 10,781,232 94.78 89.20 94.67 89.12<br />

Grohnde KWG 1,430 1,360 10,946,635 10,700,632 92.82 90.10 91.61 89.80<br />

Gundremmingen KRB C 1,344 1,288 10,361,862 10,381,798 90.41 89.20 89.85 88.50<br />

Isar KKI 2 1,485 1,410 12,127,490 12,036,656 95.46 95.95 95.24 95.68<br />

Neckarwestheim GKN II 1,400 1,310 9,703,700 10,411,410 81.35 94.03 81.00 88.00<br />

Philippsburg KKP 2 1,468 1,402 10,993,639 10,606,307 90.63 89.63 90.47 89.31<br />

Total 10,013 9,515 76,004,763 75,071,247 90.85 90.82 89.60 88.86<br />

* Availability factor (time availability factor) kt = tN/tV: The time availability factor kt<br />

is the quotient of available time of a plant (tV) and the reference period (tN).<br />

The time availability factor is a degree <strong>for</strong> the deployability of a power plant.<br />

** Energy availability factor kW = WV/WN: The energy availability factor kW is the quotient of available<br />

energy of a plant (WV ) and the nominal energy (WN). The nominal energy WN is the product<br />

of nominal capacity and reference period. This variable is used as a reference variable (100 % value)<br />

<strong>for</strong> availability considerations. The available energy WV is the energy which can be generated<br />

in the reference period due to the technical and operational condition of the plant.<br />

Energy availability factors in excess of 100 % are thus impossible, as opposed to energy utilisation.<br />

*** Inclusive of round up/down, rated power in 2019.<br />

**** The Gundremmingen nuclear power plant (KRB B) was permanently shutdown on 31 December 2017<br />

due to the revision of the German Atomic Energy Act in 2011.<br />

All data in this report as of 31 March <strong>2020</strong><br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

442<br />

Brokdorf<br />

REPORT<br />

Operating sequence in 2019<br />

100<br />

90<br />

80<br />

70<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

Electrical output in %<br />

January February March April May June July August September October November December<br />

In 2019, the Brokdorf nuclear power plant (KBR) was connected to p Trans<strong>for</strong>mers<br />

the grid <strong>for</strong> a total of 7,682 operating hours with an availability factor replacement<br />

of 82.3 %. The gross generation <strong>for</strong> the year under review was p Building ZB.9<br />

10,153,212 MWh. In 2019, the thermal reactor output was again<br />

­limited to a maximum of 95 % with a coolant temperature reduced<br />

by 3 K due to the specifications of ME 02/2017 “Increased oxide layer<br />

thickness on fuel rod cladding tubes of fuel elements”.<br />

On 13 April 2019, the plant was disconnected from the grid by<br />

triggering “Manual-TUSA” due to increasing vibrations at the turbine-­<br />

generator set. After completion of the inspection work, the unit was<br />

0<br />

reconnected to the grid on 22 April 2019.<br />

In the period from 17 to 22 November 2019, the plant was shut<br />

down to the “subcritical cold” state due to several findings on the<br />

100<br />

­baffle in the cooling water return structure.<br />

100<br />

80<br />

60<br />

40<br />

20<br />

80<br />

Planned shutdowns<br />

On 7 June 2019, the plant was shut down <strong>for</strong> the 31 st refuelling and<br />

60<br />

annual inspection.<br />

The inspection and maintenance included the following priorities:<br />

40<br />

p Reactor<br />

Full core discharge<br />

Replacement of 60 fresh fuel elements<br />

20<br />

Inspection of fuel elements,<br />

0<br />

control elements, throttle bodies.<br />

p Containment Leak rate test.<br />

p Main coolant Ring exchange electric motor<br />

pump YD30 Inspection of axial bearings<br />

Replacing the mechanical seal<br />

Positionierung: (axial bearing).<br />

p Main Bezug, coolant links, Inspection untenof the axial bearing.<br />

pump YD40<br />

p Steam generator WS test Steam generator 10/20,<br />

VGB: HKS6K additionally 30 % 30/40.<br />

p Main <strong>atw</strong>: steam 100 safety 60 Internal 0 0 inspections.<br />

and relief valves station<br />

p Cooling water Work in the pump antechambers of the<br />

secondary cooling water systems<br />

VE10/20 and 30/40<br />

Work in the main cooling water<br />

channels VA40-60.<br />

p Turbine/<br />

Control Inspection / Generator<br />

Generator ND II + III, run-out measurements,<br />

Displacement measurements<br />

Inspection bearing SB14.<br />

X = 20,475 Y = 95,25 B = 173,5 H = 38,2<br />

Exchange CS32 (emergency power supply),<br />

Exchange CS41 (normal power supply).<br />

Fire protection upgrading overflow flaps.<br />

The grid synchronisation took place on 9 July 2019 at <strong>08</strong>:17 h after<br />

31.1 days.<br />

Compared to planning, the start-up date was delayed by 4.6 days.<br />

The inspection extension is mainly due to the additional inspections<br />

of the heating tube plugs on steam generators 30 and 40.<br />

Unplanned shutdowns and reactor/turbine trip<br />

On 13 April 2019, the turbine was disconnected from the grid by<br />

triggering “manual turbine trip” due to increasing vibrations at the<br />

turbine-generator set. After completion of the inspection work, the<br />

unit was reconnected with the grid on 22 April 2019.<br />

In the period from 17 to 22 November 2019, the plant was shut<br />

down to “subcritical cold” state due to several findings on the baffle<br />

in the cooling water return structure.<br />

<strong>Power</strong> reductions above 10 % and longer than <strong>for</strong> 24 h<br />

Load reductions <strong>for</strong> the implementation of the grid-supporting<br />

power control as well as redispatch on demand of the control centre<br />

were carried out.<br />

WANO Review/Technical Support Mission<br />

The WANO Peer Review 2019 was conducted in the period 15 to<br />

26 July 2019 in the <strong>for</strong>m of an “Optimized Peer Review”, i.e.<br />

­shortened to 2 weeks. In the run-up to the WANO Peer Review 2019,<br />

a Crew Per<strong>for</strong>mance Observation (CPO) was successfully completed<br />

<strong>for</strong> the first time by the KBR shift team on the simulator of a German<br />

nuclear power plant with pressurised water reactor in the period<br />

from 6 to 9 May 2019.<br />

In summary, WANO confirmed a very good result <strong>for</strong> KBR.<br />

Delivery of fuel elements<br />

During the reporting year 28 fuel elements were delivered and<br />

stored.<br />

Waste management status<br />

By the end of the year 2019, 33 loaded CASTOR © cask were located<br />

at the Brokdorf on-site intermediate storage.<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

443<br />

Operating data<br />

Review period 2019<br />

REPORT<br />

Plant operator: PreussenElektra GmbH<br />

Shareholder/Owner: PreussenElektra GmbH (80 %),<br />

Vattenfall Europe <strong>Nuclear</strong> Energy GmbH (20 %)<br />

Plant name: Kernkraftwerk Brokdorf (KBR)<br />

Address: PreussenElektra GmbH, Kernkraftwerk Brokdorf,<br />

25576 Brokdorf, Germany<br />

Phone: +49 4829 752560<br />

Web: www.preussenelektra.de<br />

100<br />

90<br />

80<br />

70<br />

84<br />

Availability factor in %<br />

Capacity factor in %<br />

92<br />

93<br />

93<br />

93<br />

90<br />

78<br />

First synchronisation: 10-14-1986<br />

Date of commercial operation: 12-22-1986<br />

Design electrical rating (gross):<br />

1,480 MW<br />

Design electrical rating (net):<br />

1,410 MW<br />

Reactor type:<br />

PWR<br />

Supplier:<br />

Siemens/KWU<br />

60<br />

50<br />

40<br />

44<br />

The following operating results were achieved:<br />

Operating period, reactor:<br />

7,682 h<br />

Gross electrical energy generated in 2019:<br />

10,153,212 MWh<br />

Net electrical energy generated in 2019:<br />

9,635,834 MWh<br />

Gross electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

360,721,021 MWh<br />

Net electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

342,884,965 MWh<br />

Availability factor in 2019: 87.69 %<br />

Availability factor since<br />

date of commercial operation: 89.78 %<br />

Capacity factor 2019: 82.34 %<br />

Capacity factor since<br />

date of commercial operation: 85.94 %<br />

Downtime<br />

(schedule and <strong>for</strong>ced) in 2019: 12.31 %<br />

Number of reactor scrams 2019: 0<br />

30<br />

20<br />

10<br />

0<br />

10<br />

9<br />

8<br />

84<br />

2012<br />

93<br />

2013<br />

93<br />

2014<br />

93<br />

2015<br />

93<br />

2016<br />

52<br />

2017<br />

Collective radiation dose of own<br />

and outside personnel in Sv<br />

91<br />

2018<br />

88<br />

2019<br />

Licensed annual emission limits in 2019:<br />

Emission of noble gases with plant exhaust air:<br />

Emission of iodine-131 with plant exhaust air:<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

1.0 · 10 15 Bq<br />

6.0 · 10 9 Bq<br />

5.55 · 10 10 Bq<br />

Proportion of licensed annual emission limits<br />

<strong>for</strong> radioactive materials in 2019 <strong>for</strong>:<br />

Emission of noble gases with plant exhaust air: 0.075 %<br />

Emission of iodine-131 with plant exhaust air: 0.<strong>08</strong>7 %<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium): 0.0005 %<br />

Collective dose:<br />

0.158 Sv<br />

7<br />

6<br />

5<br />

4<br />

3<br />

2<br />

1<br />

0<br />

0.13<br />

2012<br />

0.22<br />

2013<br />

0.17<br />

2014<br />

0.14<br />

2015<br />

0.14<br />

2016<br />

0.13<br />

2017<br />

0.14 0.16<br />

2018 2019<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

444<br />

Emsland<br />

REPORT<br />

Operating sequence in 2019<br />

Electrical output in %<br />

100<br />

90<br />

80<br />

70<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

0<br />

January February March April May June July August September October November December<br />

Apart from the 39.4 days refuelling outage the Emsland nuclear<br />

power plant (KKE) had been operating uninterrupted and mainly at<br />

full load during the review period 2019. Producing a gross power<br />

generation of 10,781,232 MWh with a capacity factor of 89.12 % the<br />

power plant achieved a very good operating result.<br />

Planned shutdowns<br />

32 rd refuelling and 31 rd overall maintenance outage.<br />

The annual outage was scheduled <strong>for</strong> the period 17 May to 26 June.<br />

The outage took 39.4 days from breaker to breaker including an<br />

outage prolongation (18 days) due to replacement of the generator.<br />

In addition to the replacement of 40 fuel elements the following<br />

100<br />

major maintenance and inspection activities were carried out:<br />

80<br />

p Inspection of core and reactor pressure vessel internals.<br />

p Inspection of a reactor coolant pump.<br />

60<br />

p Inspection of pressurizer valves.<br />

p Eddy current test on steam generator tubes.<br />

40<br />

p Containment leak rate test.<br />

p Pressure test on different coolers and tanks.<br />

20<br />

p Inspection on main condensate pump.<br />

p0<br />

Maintenance works on different trans<strong>for</strong>mers.<br />

p Different automatic non-destructive examinations.<br />

Unplanned shutdowns and reactor/turbine trip<br />

None.<br />

<strong>Power</strong> reductions above 10 % and longer than <strong>for</strong> 24 h<br />

16 April to 17 May: 17 st Stretch-out operation.<br />

Peer Reviews<br />

From 9 September 2019 to 20 September 2019 a WANO team of<br />

10 experienced nuclear professionals from 7 different countries,<br />

conducted an optimized Peer Review (PR) at KKE NPP.<br />

In summary, the WANO exit report concludes that KKE completed<br />

the 2019 Peer Review with a very good result.<br />

Delivery of fuel elements<br />

28 Uranium-fuel elements were delivered.<br />

Waste management status<br />

No CASTOR © cask loading was carried out during the review period<br />

2019.<br />

At the end of the year 47 loaded casks were stored in the local<br />

interim storage facility.<br />

Positionierung:<br />

Bezug, links, unten<br />

X = 20,475 Y = 95,25 B = 173,5 H = 38,2<br />

VGB: HKS6K 30 %<br />

<strong>atw</strong>: 100 60 0 0<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

445<br />

Operating data<br />

Review period 2019<br />

REPORT<br />

Plant operator: Kernkraftwerke Lippe-Ems GmbH<br />

Shareholder/Owner: RWE <strong>Power</strong> AG (87,5 %),<br />

PreussenElektra GmbH (12,5 %)<br />

Plant name: Kernkraftwerk Emsland (KKE)<br />

Address: Kernkraftwerk Emsland,<br />

Am Hilgenberg, 49811 Lingen, Germany<br />

Phone: +49 591 806-1612<br />

Web: www.rwe.com<br />

100<br />

90<br />

80<br />

95<br />

Availability factor in %<br />

Capacity factor in %<br />

95<br />

95<br />

91<br />

94<br />

93<br />

95<br />

89<br />

70<br />

First synchronisation: 04-19-1988<br />

Date of commercial operation: 06-20-1988<br />

Design electrical rating (gross):<br />

1,406 MW<br />

Design electrical rating (net):<br />

1,335 MW<br />

Reactor type:<br />

PWR<br />

Supplier:<br />

Siemens/KWU<br />

60<br />

50<br />

40<br />

The following operating results were achieved:<br />

Operating period, reactor:<br />

7,821 h<br />

Gross electrical energy generated in 2019:<br />

10,781,232 MWh<br />

Net electrical energy generated in 2019:<br />

10,237,<strong>09</strong>3 MWh<br />

Gross electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

357,600,201 MWh<br />

Net electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

339,066,997 MWh<br />

Availability factor in 2019: 89.20 %<br />

Availability factor since<br />

date of commercial operation: 93.91 %<br />

Capacity factor 2019: 89.12 %<br />

Capacity factor since<br />

date of commercial operation: 93.77 %<br />

Downtime<br />

(schedule and <strong>for</strong>ced) in 2019: 10.80 %<br />

Number of reactor scrams 2019: 0<br />

30<br />

20<br />

10<br />

0<br />

10<br />

9<br />

8<br />

95<br />

2012<br />

95<br />

2013<br />

95<br />

2014<br />

91<br />

2015<br />

94<br />

2016<br />

93<br />

2017<br />

Collective radiation dose of own<br />

and outside personnel in Sv<br />

95<br />

2018<br />

89<br />

2019<br />

Licensed annual emission limits in 2019:<br />

Emission of noble gases with plant exhaust air:<br />

Emission of iodine-131 with plant exhaust air:<br />

(incl. H-3 and C-14)<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

1.0 · 10 15 Bq<br />

5.0 · 10 9 Bq<br />

3.7 · 10 10 Bq<br />

7<br />

6<br />

5<br />

Proportion of licensed annual emission limits<br />

<strong>for</strong> radioactive materials in 2019 <strong>for</strong>:<br />

Emission of noble gases with plant exhaust air: 0.016 %<br />

Emission of iodine-131 with plant exhaust air: 0.0 %<br />

(incl. H-3 and C-14)<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium): 0.00 %<br />

Collective dose:<br />

0.067 Sv<br />

4<br />

3<br />

2<br />

1<br />

0<br />

0.<strong>09</strong><br />

2012<br />

0.<strong>08</strong><br />

2013<br />

0.06<br />

2014<br />

0.10<br />

2015<br />

0.05<br />

2016<br />

0.<strong>09</strong><br />

2017<br />

0.06 0.07<br />

2018 2019<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

446<br />

Grohnde<br />

REPORT<br />

Operating sequence in 2019<br />

100<br />

90<br />

80<br />

70<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

Electrical output in %<br />

January February March April May June July August September October November December<br />

During the 2019 reporting year, the Grohnde nuclear power plant<br />

was scheduled <strong>for</strong> a 36-day overhaul with refueling from grif and<br />

80<br />

achieved a time availability of 90.1 %. Gross generation amounted to<br />

10,700,632 MWh.<br />

60<br />

Compared to the planned 26 days, the inspection was extended<br />

by 250 hours. Delays in the inspection of the drive rods, the<br />

40<br />

­replacement of the LVD lance G10, a conspicuous closing behaviour<br />

of the DH spray valve YP10 S233 and contamination of the turbine oil<br />

20<br />

were the main reasons <strong>for</strong> the revision delay.<br />

Due to a high Weser water temperature, the output was reduced<br />

0<br />

to 447 MW on June 26 and then increased again to full capacity.<br />

According to the specifications of the load distribution system,<br />

26 load reductions were made in 2019 over a total of 228 hours and<br />

100<br />

156 mains and 70 primary controls <strong>for</strong> a total of 4,307 hours.<br />

100<br />

After completion of the work on bracing the RPV head and closing<br />

the primary circuit (RKL hydraulically sealed), a limited availability<br />

of a core instrumentation lance (internal neutron measuring system<br />

<strong>for</strong> the determination and monitoring of the power distribution<br />

density) and 3 Fuel element exit temperature measurements, caused<br />

by a defective plug connection, were determined. The defective core<br />

instrumentation lance was replaced.<br />

Unplanned shutdowns and reactor/turbine trip<br />

None.<br />

<strong>Power</strong> reductions above 10 % and longer than <strong>for</strong> 24 h<br />

In the months January, February, April, October and December load<br />

following operation due to requirements of the load distributor.<br />

80<br />

Planned shutdowns<br />

21 April to 27 May: 36 th refuelling and plant inspection with<br />

60<br />

maintenance.<br />

As planned, the Grohnde nuclear power plant was shut down on<br />

40<br />

April 21 <strong>for</strong> revision and the 36 th refuelling. The main planned work<br />

of the inspection and maintenance were:<br />

20<br />

p Unloading and loading with the insertion<br />

0<br />

of 52 fresh fuel elements.<br />

p Full inspection on 19 fuel elements.<br />

p Eddy current testing on 32 control rods.<br />

p Visual inspection on 15 throttle bodies.<br />

p YD10 D001 Overhaul axial bearing.<br />

p YD30 Positionierung:<br />

D001 Motor conversion.<br />

p YD20 Bezug, D001 Replacing links, unten the mechanical seal.<br />

p YD40 D001 Replacing the mechanical seal &WS test<br />

of the pump shaft.<br />

p Start-up VGB: test HKS6K of the BE 30 centring % pins of the UKG and the OKG.<br />

p VA01 <strong>atw</strong>: + VA02 100 Inspection 60 0 0and cleaning of cooling water sections.<br />

p TF20 B001 Cleaning the nuclear intercooler.<br />

p TF20 S013/S014 Replacing the screws on the quick-closing flaps.<br />

p TH20 Inspection of secondary shut-offs with pipe freezing.<br />

p Work and inspections in the redundancies with the main focus<br />

of activities in the main redundancy 2/6 (maintenance work on<br />

valves and actuators and tests on tanks, batteries and<br />

electrotechnical branches).<br />

X = 20,475 Y = 95,25 B = 173,5 H = 38,2<br />

Delivery of fuel elements<br />

In March 2019 the delivery of 44 U-/U-Gd fuel elements of the<br />

company Westinghouse took place.<br />

Waste management status<br />

No CASTOR © V/19 containers were loaded in 2019.<br />

The interim storage facility with 34 stored CASTOR © V/19 casks<br />

was handed over to Bundesgesellschaft für Zwischenlagerung mbH<br />

(BGZ).<br />

General points/management systems<br />

In September 2019, the surveillance audit of the quality ­management<br />

system (ISO 9001) and the recertification of the environmental<br />

management system (ISO 14001) and the occupational health<br />

and safety management system (OHSAS 18001) were successfully<br />

completed.<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

447<br />

Operating data<br />

Review period 2019<br />

REPORT<br />

Plant operator: Gemeinschaftskernkraftwerk Grohnde GmbH & Co. OHG<br />

Shareholder/Owner: PreussenElektra GmbH (83,3 %),<br />

Stadtwerke Bielefeld (16,7 %)<br />

Plant name: Kernkraftwerk Grohnde (KWG)<br />

Address: Gemeinschaftskernkraftwerk Grohnde GmbH & Co. OHG,<br />

P.O. bx 12 30, 31857 Emmerthal, Germany<br />

Phone: +49 5155 67-1<br />

E-mail: kwg-kraftwerksleitung@preussenelektra.de<br />

Web: www.preussenelektra.de<br />

100<br />

90<br />

80<br />

70<br />

95<br />

Availability factor in %<br />

Capacity factor in %<br />

89<br />

84<br />

89<br />

73<br />

82<br />

92<br />

90<br />

First synchronisation: <strong>09</strong>-05-1984<br />

Date of commercial operation: 02-01-1985<br />

Design electrical rating (gross):<br />

1,430 MW<br />

Design electrical rating (net):<br />

1,360 MW<br />

Reactor type:<br />

PWR<br />

Supplier:<br />

Siemens/KWU<br />

60<br />

50<br />

40<br />

The following operating results were achieved:<br />

Operating period, reactor:<br />

7,889 h<br />

Gross electrical energy generated in 2019:<br />

10,700,632 MWh<br />

Net electrical energy generated in 2019:<br />

10,113,330 MWh<br />

Gross electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

388,274,835 MWh<br />

Net electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

367,<strong>08</strong>2,606 MWh<br />

Availability factor in 2019: 90.10 %<br />

Availability factor since<br />

date of commercial operation: 91.70 %<br />

Capacity factor 2019: 89.80 %<br />

Capacity factor since<br />

date of commercial operation: 91.30 %<br />

Downtime<br />

(schedule and <strong>for</strong>ced) in 2019: 9.90 %<br />

Number of reactor scrams 2019: 0<br />

30<br />

20<br />

10<br />

0<br />

10<br />

9<br />

8<br />

95<br />

2012<br />

90<br />

2013<br />

84<br />

2014<br />

89<br />

2015<br />

75<br />

2016<br />

86<br />

2017<br />

Collective radiation dose of own<br />

and outside personnel in Sv<br />

93<br />

2018<br />

90<br />

2019<br />

Licensed annual emission limits in 2019:<br />

Emission of noble gases with plant exhaust air:<br />

Emission of iodine-131 with plant exhaust air:<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

9.0 · 10 14 Bq<br />

7.5 · 10 9 Bq<br />

5.55 · 10 10 Bq<br />

Proportion of licensed annual emission limits<br />

<strong>for</strong> radioactive materials in 2019 <strong>for</strong>:<br />

Emission of noble gases with plant exhaust air: 0.019 %<br />

Emission of iodine-131 with plant exhaust air: 0.000 %<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium): 0.000 %<br />

Collective dose:<br />

0.261 Sv<br />

7<br />

6<br />

5<br />

4<br />

3<br />

2<br />

1<br />

0<br />

0.27<br />

2012<br />

0.54<br />

2013<br />

0.25<br />

2014<br />

0.31<br />

2015<br />

0.52<br />

2016<br />

0.23<br />

2017<br />

0.12 0.26<br />

2018 2019<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

448<br />

Gundremmingen C<br />

REPORT<br />

Operating sequence in 2019<br />

Electrical output in %<br />

100<br />

90<br />

80<br />

70<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

0<br />

January February March April May June July August September October November December<br />

In the review year 2019, unit C of Gundremmingen (KRB C) nuclear<br />

power plant was operated at full load without any major restrictions<br />

except <strong>for</strong> one planned outage <strong>for</strong> refuelling.<br />

From 4 April to 21 April unit C was in stretch out operation.<br />

During the shutdown a total of 152 fuel elements were unloaded and<br />

replaced with 112 fresh and 40 (2 MOX) partially spent fuel elements.<br />

During the outage all safety relevant workings were monitored<br />

by the relevant nuclear controlling authority, the Bavarian State<br />

­Ministry of the Environment and Consumer Protection (StMUV), and<br />

consulted authorized experts. The inspection of the technical systems<br />

with regard to safety and reliability confirmed the excellent ­condition<br />

of the plant.<br />

A gross total of 10,381,798 MWh of electricity was produced.<br />

100<br />

Planned shutdowns<br />

80<br />

21 April to 29 May 2019: 33 th refuelling outage and 21 th overall<br />

maintenance inspection.<br />

60<br />

The following major activities were carried out:<br />

p Refuelling and sipping of all fuel elements inside the core; result:<br />

40<br />

two defective fuel elements.<br />

p20<br />

Visual inspection and non-destructive testing of reactor pressure<br />

vessel stud bolts and internals.<br />

p0<br />

Inspection of main isolation valves of main steam and safety and<br />

relief valves.<br />

p Emptying of redundancies 1 and 3 <strong>for</strong> preventive measures on<br />

valves and tanks.<br />

p Inner inspection and pressure tests on high pressure preheater<br />

strings Positionierung:<br />

and reactor water clean-up system.<br />

p Extensive Bezug, non-destructive links, unten testing of pipes and tanks.<br />

p Inspection of two emergency diesel generators.<br />

p Precautionary replacement of 10 kV power cables.<br />

X = 20,475 Y = 95,25 B = 173,5 H = 38,2<br />

VGB: HKS6K 30 %<br />

<strong>atw</strong>: 100 60 0 0<br />

Unplanned shutdowns and reactor/turbine trip<br />

29 to 31 May: Manual scram due to safety valve staying in open<br />

position during a period test, subsequently exchange of one of his<br />

pilot valves.<br />

<strong>Power</strong> reductions above 10 % and longer than <strong>for</strong> 24 h<br />

24 to 25 February: Periodic tests.<br />

4 to 21 April: Stretch-out-operation.<br />

4 to 5 August: period tests and maintenance work.<br />

1 to 4 December: period tests, change of the control rod traversing<br />

order , leak detection in turbine condenser and maintenance work.<br />

Delivery of fuel elements<br />

In 2019, no fresh fuel elements were delivered.<br />

Waste management status<br />

In 2019, a total of 9 CASTOR © casks were loaded. Thus, at the end of<br />

2019, 69 CASTOR © casks with each 52 spent fuel elements out of<br />

units B and C are stored in the local interim storage.<br />

General points<br />

In the year 2019, the recertification of the environmental ­ management<br />

system (ISO 14001) and energy management system (ISO<br />

50001) were successfully carried out.<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

449<br />

Operating data<br />

Review period 2019<br />

REPORT<br />

Plant operator: Kernkraftwerk Gundremmingen GmbH<br />

Shareholder/Owner: RWE <strong>Power</strong> AG (75 %),<br />

PreussenElektra GmbH (25 %)<br />

Plant name: Kernkraftwerk Gundremmingen C (KRB C)<br />

Address: Kernkraftwerk Gundremmingen GmbH,<br />

Dr.-August-Weckesser-Straße 1, 89355 Gundremmingen, Germany<br />

Phone: +49 8224 78-1<br />

E-mail: kontakt@kkw-gundremmingen.de<br />

Web: www.kkw-gundremmingen.de<br />

100<br />

90<br />

80<br />

70<br />

91<br />

Availability factor in %<br />

Capacity factor in %<br />

89<br />

90<br />

90<br />

86<br />

86<br />

90<br />

89<br />

First synchronisation: 11-02-1984<br />

Date of commercial operation: 01-18-1985<br />

Design electrical rating (gross):<br />

1,344 MW<br />

Design electrical rating (net):<br />

1,288 MW<br />

Reactor type:<br />

BWR<br />

Supplier:<br />

Siemens/KWU,<br />

Hochtief<br />

The following operating results were achieved:<br />

Operating period, reactor:<br />

7,810 h<br />

Gross electrical energy generated in 2019:<br />

10,381,798 MWh<br />

Net electrical energy generated in 2019:<br />

9,900,234 MWh<br />

Gross electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

341,323,552 MWh<br />

Net electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

325,<strong>08</strong>2,303 MWh<br />

Availability factor in 2019: 89.20 %<br />

Availability factor since<br />

date of commercial operation: 89.20 %<br />

Capacity factor 2019: 88.50 %<br />

Capacity factor since<br />

date of commercial operation: 87.60 %<br />

Downtime<br />

(schedule and <strong>for</strong>ced) in 2019: 10.80 %<br />

Number of reactor scrams 2019: 1<br />

Licensed annual emission limits in 2019<br />

(values added up <strong>for</strong> Units B and C, site emission):<br />

Emission of noble gases with plant exhaust air:<br />

1.85 · 10 15 Bq<br />

Emission of iodine-131 with plant exhaust air:<br />

2.20 · 10 10 Bq<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

1.10 · 10 11 Bq<br />

Proportion of licensed annual emission limits <strong>for</strong> radioactive<br />

materials in 2019 <strong>for</strong> (values added up <strong>for</strong> Units B and C):<br />

Emission of noble gases with plant exhaust air: 0.48 %<br />

Emission of iodine-131 with plant exhaust air: 0.40 %<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium): 0.16 %<br />

Collective dose:<br />

0.79 Sv<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

0<br />

10<br />

9<br />

8<br />

7<br />

6<br />

5<br />

4<br />

3<br />

2<br />

1<br />

92<br />

2012<br />

0.78<br />

90<br />

2013<br />

Collective radiation dose of own<br />

and outside personnel in Sv<br />

1.36<br />

90<br />

2014<br />

1.14<br />

90<br />

2015<br />

1.49<br />

86<br />

2016<br />

0.84<br />

88<br />

2017<br />

0.89<br />

90<br />

2018<br />

0.55<br />

89<br />

2019<br />

0.79<br />

0<br />

2012<br />

2013<br />

2014<br />

2015<br />

2016<br />

2017<br />

2018 2019<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

450<br />

Isar 2<br />

REPORT<br />

Operating sequence in 2019<br />

100<br />

90<br />

80<br />

70<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

Electrical output in %<br />

January February March April May June July August September October November December<br />

With a gross electricity generation of 12,036,656 MWh and a<br />

­availability of 95.68 %, unit 2 of Isar (KKI 2) nuclear power plant<br />

achieved an excellent operating result in 2019. The unit also made an<br />

important contribution to grid stability through increased load<br />

sequence and control operation, which, however, reduced the net<br />

electricity supply that can be generated by 384,950,000 MWh,<br />

corresponding to 11.4 full operation days. The highest generator<br />

­active power was reached on 20.01.2019 and amounted to 1520 MW.<br />

Planned shutdowns<br />

The fuel element replacement with plant inspection and maintenance<br />

took place from 13 July 2019 to 27 July 2019 with a duration of<br />

14.8 days. During the inspection and maintenance, 48 new fuel<br />

100<br />

elements were inserted.<br />

80<br />

Unplanned shutdowns and reactor/turbine trip<br />

None.<br />

60<br />

<strong>Power</strong> reductions above 10 % and longer than <strong>for</strong> 24 h<br />

40<br />

None.<br />

20<br />

Safety Reviews<br />

20<br />

0<br />

and 22 Februar: Management evaluation KKI.<br />

6 March: Review of operations by the management<br />

of PreussenElektra GmbH.<br />

11 to 15 March:­­ ­ ­Re-certification audit by DNV GL<br />

Business Assurance Zertifizierung & Umweltgutachter<br />

GmbH according to DIN EN ISO<br />

Positionierung:<br />

Bezug, links, 9001/14001, untenBS OHSAS 18001 and EMAS.<br />

12 March­­ ­ Inspection in accordance with §16 of the Major<br />

and 30 April:­ ­ ­Accidents Ordinance (Störfall Verordnung) –<br />

VGB: HKS6K Fire 30 Protection % and Immission Control.<br />

3 and <strong>atw</strong>: 4 July: 100 60 Internal 0 0audit<br />

“Measurement and test<br />

equipment monitoring” in KKI.<br />

7 August:­­ ­ Management system status meeting.<br />

8, 18 October Management system audit - Part 1 in the KKI.<br />

and 13 November:<br />

10 October: 2 nd operational review (half-year review) by<br />

the management of PreussenElektra GmbH.<br />

29 October:­­ ­ Audit “Fuel element handling and<br />

and 4/5 November fuel element disposal“.<br />

4/5 December: Plant inspection “Integrated<br />

Management System.<br />

X = 20,475 Y = 95,25 B = 173,5 H = 38,2<br />

Delivery of fuel elements<br />

In the year under review, 40 uranium fuel elements were delivered<br />

from Westinghouse. The dry storage facility contains 16 uranium fuel<br />

elements in stock.<br />

Waste management status<br />

In 2019, no fuel elements were stored in the BELLA on-site interim<br />

storage facility.<br />

Of the storage and transport casks stored in the on-site interim<br />

storage facility, 26 CASTOR © V/19 casks as well as 7 TN24E have to<br />

be assigned to KKI 2.<br />

The interim storage facility was taken over by Bundesgesellschaft<br />

für Zwischenlagerung mbH (BGZ) on 1 January 2019.<br />

General points<br />

Due to increased oil temperatures at the engine of the main coolant<br />

pump JEB30 AP001, a precautionary bearing oil change was carried<br />

out. For this purpose, the plant output was reduced on 10 August<br />

2019 from 07:00 hrs to the minimum load point of 360 MW with a<br />

gradient of 30 MW/min.<br />

After the scheduled completion of the exchange and start of the<br />

main coolant pump, full load operation was reached again around<br />

6:34 pm.<br />

An internal emergency exercise was conducted on 18 November2019.<br />

The exercise started within the normal working hours. The scenario<br />

assumed was an earthquake close to the site, which caused the<br />

failure of the external main electricity supply and the cooling water<br />

supply of the emergency diesel generators as well as leakages in the<br />

storage pool cooling water pipe. In addition, the failure of various<br />

communication systems was assumed in the course of the exercise.<br />

The emergency exercise was completed professionally and<br />

purposefully by a highly motivated team.<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

451<br />

Operating data<br />

Review period 2019<br />

REPORT<br />

Plant operator: PreussenElektra GmbH<br />

Shareholder/Owner: PreussenElektra GmbH (75 %),<br />

Stadtwerke München GmbH (25 %)<br />

Plant name: Kernkraftwerk Isar 2 (KKI 2)<br />

Address: PreussenElektra GmbH, Kernkraftwerk Isar,<br />

Postfach 11 26, 84049 Essenbach, Germany<br />

Phone: +49 8702 38-2465<br />

Web: www.preussenelektra.de<br />

100<br />

90<br />

80<br />

94<br />

Availability factor in %<br />

Capacity factor in %<br />

94<br />

90<br />

89<br />

96<br />

91<br />

95<br />

96<br />

70<br />

First synchronisation: 01-22-1988<br />

Date of commercial operation: 04-<strong>09</strong>-1988<br />

Design electrical rating (gross):<br />

1,485 MW<br />

Design electrical rating (net):<br />

1,410 MW<br />

Reactor type:<br />

PWR<br />

Supplier:<br />

Siemens/KWU<br />

60<br />

50<br />

40<br />

The following operating results were achieved:<br />

Operating period, reactor:<br />

8,405 h<br />

Gross electrical energy generated in 2019:<br />

12,036,656 MWh<br />

Net electrical energy generated in 2019:<br />

11,375,505 MWh<br />

Gross electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

365,762,469 MWh<br />

Net electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

345,352,<strong>09</strong>4 MWh<br />

Availability factor in 2019: 95.95 %<br />

Availability factor since<br />

date of commercial operation: 93.36 %<br />

Capacity factor 2019: 95.68 %<br />

Capacity factor since<br />

date of commercial operation: 92.48 %<br />

Downtime<br />

(schedule and <strong>for</strong>ced) in 2019: 4.05 %<br />

Number of reactor scrams 2019: 0<br />

30<br />

20<br />

10<br />

0<br />

10<br />

9<br />

8<br />

94<br />

2012<br />

96<br />

2013<br />

95<br />

2014<br />

89<br />

2015<br />

96<br />

2016<br />

92<br />

2017<br />

Collective radiation dose of own<br />

and outside personnel in Sv<br />

95<br />

2018<br />

96<br />

2019<br />

Licensed annual emission limits in 2019:<br />

Emission of noble gases with plant exhaust air:<br />

Emission of iodine-131 with plant exhaust air:<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

1.1 · 10 15 Bq<br />

1.1 · 10 10 Bq<br />

5.5 · 10 10 Bq<br />

Proportion of licensed annual emission limits<br />

<strong>for</strong> radioactive materials in 2019 <strong>for</strong>:<br />

Emission of noble gases with plant exhaust air: 0.127 %<br />

Emission of iodine-131 with plant exhaust air:<br />

< limit of detection<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

< limit of detection<br />

Collective dose:<br />

0.047 Sv<br />

7<br />

6<br />

5<br />

4<br />

3<br />

2<br />

1<br />

0<br />

0.14<br />

2012<br />

0.<strong>08</strong><br />

2013<br />

0.<strong>09</strong><br />

2014<br />

0.25<br />

2015<br />

0.06<br />

2016<br />

0.14<br />

2017<br />

0.06 0.05<br />

2018 2019<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

452<br />

Neckarwestheim II<br />

REPORT<br />

Operating sequence in 2019<br />

100<br />

90<br />

80<br />

70<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

Electrical output in %<br />

January February March April May June July August September October November December<br />

In 100 2019, the Neckarwestheim II nuclear power plant (GKN II)<br />

generated a gross output of 10,411,400 MWh. Net electrical<br />

­generation 80 was 9,758,339 MWh, of which 9,371,600 MWh were<br />

­supplied to the public three-phase grid and 1,039,800 MWh to the<br />

static 60 converter system of Deutsche Bahn AG. The plant was<br />

­connected to the grid <strong>for</strong> 7,699.8 hours. This results in a time<br />

­utilization 40 of 87.90 %.<br />

Since<br />

20<br />

the three-phase alternating current machine was<br />

­commissioned, 340,241,584 MWh gross and 318,174,476 MWh net<br />

0<br />

have been generated.<br />

Planned shutdowns<br />

100<br />

9 August to 22 September: 36 th fuel element replacement and annual<br />

inspection with maintenance.<br />

80<br />

The inspection and maintenance included the following priorities:<br />

p Fuel element replacement with the use of 40 new fuel elements.<br />

60<br />

p Eddy current tests of the heating tubes of all 4 steam generators.<br />

p Secondary tube sheet inspection on all 4 steam generators.<br />

40<br />

p Major overhaul of a primary-side safety valve on the<br />

pressuriser JEF10.<br />

20<br />

p Pressure test of the heat exchangers in the volume control<br />

0<br />

system.<br />

p Leak rate check of the containment.<br />

p Partial major overhaul of the main feed water pump LAC30 and<br />

the main condensate pump LCB10.<br />

p Major overhaul of main steam valves at LBA10 and LBA40.<br />

p Maintenance Positionierung:<br />

activities on trans<strong>for</strong>mers and on<br />

both Bezug, grid connections. links, unten<br />

p Maintenance on the switchgear and on mechanical components<br />

in the main redundancy 2/6.<br />

X = 20,475 Y = 95,25 B = 173,5 H = 38,2<br />

VGB: HKS6K 30 %<br />

<strong>atw</strong>: 100 60 0 0<br />

Unplanned shutdowns and reactor/turbine trip<br />

3 to 22 September: Unplanned extension of the maintenance.<br />

<strong>Power</strong> reductions above 10 % and longer than <strong>for</strong> 24 h<br />

13 June to 9 August: Stretch-out operation<br />

January to April and October to December: Load sequence ­operation.<br />

Integrated management system (IMS) EnKK<br />

The Integrated Management System (IMS) of EnBW Kernkraft<br />

GmbH (EnKK) according to KTA 1402 with the partial systems<br />

<strong>for</strong> nuclear safety (SMS), quality management (QMS/QSÜ),<br />

­occupational safety management (AMS) as well as environmental<br />

and energy management (UMS, EnMS) was also continuously<br />

­further developed in 2019. The scope and content of the respective<br />

process descriptions were gradually adapted to the various internal<br />

requirements and the related approval-relevant specifications.<br />

The completeness and effectiveness (con<strong>for</strong>mity) of the processoriented<br />

IMS, including the quality management measures, were<br />

confirmed by appropriate internal audits as well as by a several-day<br />

inspection by the assessor (ESN) and the supervisory authority at the<br />

GKN and KKP sites.<br />

The modular and demand-oriented structure of the IMS according to<br />

KTA 1402 also enables continuous and efficient adaptation to the<br />

site-specific requirements in operation/post-operation in subsequent<br />

years. Another important focus will be the gradual integration of<br />

dismantling aspects into the IMS in order to exploit synergy effects.<br />

Waste management status<br />

In 2019, 5 CASTOR © V/19 casks were loaded with 69 GKN I and 2<br />

GKN II fuel elements and transported to the Neckarwestheim on-site<br />

interim storage facility. At the end of 2019, 716 GKN II fuel elements<br />

(dry storage, wet storage and reactor) and 49 GKN I fuel elements<br />

(wet storage) were thus in the GKN II facility. Since 1 January 2019,<br />

the on-site interim storage facility in Neckarwestheim has been<br />

operated by the federally owned Gesellschaft für Zwischenlagerung<br />

(BGZ). This is the implementation of the “Act on the Reorganisation<br />

of Responsibility in <strong>Nuclear</strong> Waste Management”.<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

453<br />

Operating data<br />

Review period 2019<br />

REPORT<br />

Plant operator: EnBW Kernkraft GmbH (EnKK)<br />

Shareholder/Owner: EnBW Erneuerbare und Konventionelle<br />

Erzeugung AG (98,45 %), ZEAG Energie AG, Deutsche Bahn AG,<br />

Kernkraftwerk Obrigheim GmbH<br />

Plant name: Kernkraftwerk Neckarwestheim II (GKN II)<br />

Address: EnBW Kernkraft GmbH, Kernkraftwerk Neckarwestheim,<br />

Im Steinbruch, 74382 Neckarwestheim, Germany<br />

Phone: +49 7133 13-0<br />

E-mail: poststelle-gkn@kk.enbw.com<br />

Web: www.enbw.com<br />

100<br />

90<br />

80<br />

70<br />

92<br />

Availability factor in %<br />

Capacity factor in %<br />

90<br />

93<br />

93<br />

94<br />

89<br />

81<br />

88<br />

First synchronisation: 01-03-1989<br />

Date of commercial operation: 04-15-1989<br />

Design electrical rating (gross):<br />

1,400 MW<br />

Design electrical rating (net):<br />

1,310 MW<br />

Reactor type:<br />

PWR<br />

Supplier:<br />

Siemens/KWU<br />

60<br />

50<br />

40<br />

The following operating results were achieved:<br />

Operating period, reactor:<br />

7,706 h<br />

Gross electrical energy generated in 2019:<br />

10,411,400 MWh<br />

Net electrical energy generated in 2019:<br />

9,758,339 MWh<br />

Gross electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

340,241,584 MWh<br />

Net electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

318,174,476 MWh<br />

Availability factor in 2019: 94.03 %<br />

Availability factor since<br />

date of commercial operation: 92.92 %<br />

Capacity factor 2019: 88.00 %<br />

Capacity factor since<br />

date of commercial operation: 92.55 %<br />

Downtime<br />

(schedule and <strong>for</strong>ced) in 2019: 5.87 %<br />

Number of reactor scrams 2019: 0<br />

30<br />

20<br />

10<br />

0<br />

10<br />

9<br />

8<br />

92<br />

2012<br />

90<br />

2013<br />

93<br />

2014<br />

93<br />

2015<br />

95<br />

2016<br />

89<br />

2017<br />

Collective radiation dose of own<br />

and outside personnel in Sv<br />

81<br />

2018<br />

94<br />

2019<br />

Licensed annual emission limits in 2019:<br />

Emission of noble gases with plant exhaust air:<br />

Emission of iodine-131 with plant exhaust air:<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

1.0 · 10 15 Bq<br />

1.1 · 10 10 Bq<br />

6.0 · 10 10 Bq<br />

Proportion of licensed annual emission limits<br />

<strong>for</strong> radioactive materials in 2019 <strong>for</strong>:<br />

Emission of noble gases with plant exhaust air: 0.0<strong>09</strong>8 %<br />

Emission of iodine-131 with plant exhaust air:<br />

< limit of detection<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

< limit of detection<br />

Collective dose:<br />

0.<strong>09</strong>6 Sv<br />

7<br />

6<br />

5<br />

4<br />

3<br />

2<br />

1<br />

0<br />

0.13<br />

2012<br />

0.<strong>08</strong><br />

2013<br />

0.10<br />

2014<br />

0.12<br />

2015<br />

0.<strong>08</strong><br />

2016<br />

0.15<br />

2017<br />

0.12 0.10<br />

2018 2019<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

454<br />

Philippsburg 2<br />

REPORT<br />

Operating sequence in 2019<br />

100<br />

90<br />

80<br />

70<br />

60<br />

50<br />

40<br />

30<br />

20<br />

10<br />

Electrical output in %<br />

January February March April May June July August September October November December<br />

In 100 the reporting year 2019 the nuclear power plant block<br />

­Philippsburg 2 (KKP 2) generated a gross output of 10,993,639 MWh.<br />

The<br />

80<br />

net electrical power generation consisted of 10,323,151 MWh.<br />

The plant was 7,939 h on the grid. This corresponds to a availabilty<br />

60<br />

factor of 90.63 %.<br />

Since the commissioning of the plant 366,161,155 MWh gross and<br />

40<br />

347,076,473 MWh net were generated<br />

20<br />

Planned shutdowns<br />

11 May to 15 June: 33 nd refuelling and annual major inspection.<br />

0<br />

Major inspection work carried out:<br />

p Inspection of one of the three main feed pumps.<br />

p<br />

100<br />

Eddy current testing of two of the four steam generators.<br />

p Leak test of reactor containment.<br />

p 80 Inspection of the main cooling water system.<br />

p Engine replacement on two of six main cooling water pumps.<br />

p Maintenance work on individual emergency power generators.<br />

60<br />

Unplanned 40 shutdowns and reactor/turbine trip<br />

18 August: Turbine trip (TUSA) via the criterion “high condenser<br />

pressure”. 20<br />

<strong>Power</strong> 0 reductions above 10 % and longer than <strong>for</strong> 24 h<br />

15 March to 11 May: Stretch-out operation<br />

26 July to 24 August: Reduction of heat input into the Rhine and<br />

compliance with the permissible outlet temperature.<br />

15 October to 2 November: Reduction of heat input into the Rhine<br />

and compliance Positionierung:<br />

with the permissible outlet temperature.<br />

8 November Bezug, to 3 links, December: unten Reduction of heat input into the Rhine<br />

and compliance with the permissible outlet temperature.<br />

X = 20,475 Y = 95,25 B = 173,5 H = 38,2<br />

VGB: HKS6K 30 %<br />

<strong>atw</strong>: 100 60 0 0<br />

Integrated management system (IMS) EnKK<br />

(NPP P, GKN, KWO)<br />

The integrated management system (IMS) of the EnBW Kernkraft<br />

GmbH (EnKK) with its partial system <strong>for</strong> nuclear safety (SMS),<br />

quality management (QMS/QSÜ) as well as environmental<br />

and energy management (UMS, EnMS, Umwelt- und Energiemanagementsystem)<br />

were also in 2019 continuously further<br />

developed. Scope and content of each process descriptions were<br />

gradually adapted to the different internal requirements and related<br />

approval criteria. Besides the confirmation of con<strong>for</strong>mity <strong>for</strong> the<br />

IMS, the recertification of the EnKK energy management system<br />

(EnMS, Energiemanagementsystem) according to DIN EN ISO 50001<br />

took place in 2019 to improve energy efficiency. The certificate<br />

was thus exten ded by three years.<br />

The completeness and effectiveness of the process-oriented IMS,<br />

including the quality management measures, were confirmed by<br />

appropriate internal audits as well as by a several-day inspection by<br />

the expert (ESN) and the supervisory authority at the GKN and KKP<br />

sites.<br />

The modular and demand-oriented structure of the IMS according<br />

to KTA1402 also enables continuous and efficient adaptation to the<br />

site-specific requirements in operation/post-operation in subsequent<br />

years. Another important focus will be the gradual integration of<br />

dismantling aspects into the IMS in order to exploit synergy effects.<br />

Waste management status<br />

During the year 2019 in total 2 transportation and storage casks of<br />

type CASTOR © V/19 were stored in the on-site intermediate storage.<br />

Altogether 33 loaded CASTOR © V/19 and 29 loaded CASTOR ©<br />

V/25 casks were at the on-site intermediate storage.<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

455<br />

Operating data<br />

Review period 2019<br />

REPORT<br />

Plant operator: EnBW Kernkraft GmbH (EnKK)<br />

Shareholder/Owner: EnBW Erneuerbare und Konventionelle<br />

Erzeugung AG (98,45 %), ZEAG Energie AG, Deutsche Bahn AG,<br />

Kernkraftwerk Obrigheim GmbH<br />

Plant name: Kernkraftwerk Philippsburg 2 (KKP 2)<br />

Address: EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg,<br />

P.O. box 11 40, 76652 Philippsburg, Germany<br />

Phone: +49 7256 95-0<br />

E-mail: Poststelle-kkp@kk.enbw.com<br />

Web: www.enbw.com<br />

100<br />

90<br />

80<br />

70<br />

86<br />

Availability factor in %<br />

Capacity factor in %<br />

73<br />

82<br />

90<br />

82<br />

90<br />

89<br />

First synchronisation: 12-17-1984<br />

Date of commercial operation: 04-18-1985<br />

Design electrical rating (gross):<br />

1,468 MW<br />

Design electrical rating (net):<br />

1,402 MW<br />

Reactor type:<br />

PWR<br />

Supplier:<br />

Siemens/KWU<br />

60<br />

50<br />

40<br />

63<br />

The following operating results were achieved:<br />

Operating period, reactor:<br />

7,865 h<br />

Gross electrical energy generated in 2019:<br />

10,606,307 MWh<br />

Net electrical energy generated in 2019:<br />

9,963,117 MWh<br />

Gross electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

376,767,462 MWh<br />

Net electrical energy generated since<br />

first synchronisation until 12-31-2019:<br />

357,039,589 MWh<br />

Availability factor in 2019: 89.63 %<br />

Availability factor since<br />

date of commercial operation: 88.78 %<br />

Capacity factor 2019: 89.31 %<br />

Capacity factor since<br />

date of commercial operation: 88.51 %<br />

Downtime<br />

(schedule and <strong>for</strong>ced) in 2019: 10.37 %<br />

Number of reactor scrams 2019: 0<br />

30<br />

20<br />

10<br />

0<br />

10<br />

9<br />

8<br />

86<br />

2012<br />

73<br />

2013<br />

82<br />

2014<br />

91<br />

2015<br />

82<br />

2016<br />

63<br />

2017<br />

Collective radiation dose of own<br />

and outside personnel in Sv<br />

91<br />

2018<br />

90<br />

2019<br />

Licensed annual emission limits in 2019:<br />

Emission of noble gases with plant exhaust air:<br />

Emission of iodine-131 with plant exhaust air:<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium):<br />

1.1 · 10 15 Bq<br />

1.1 · 10 10 Bq<br />

5.5 · 10 10 Bq<br />

Proportion of licensed annual emission limits<br />

<strong>for</strong> radioactive materials in 2019 <strong>for</strong>:<br />

Emission of noble gases with plant exhaust air: 0.11 %<br />

Emission of iodine-131 with plant exhaust air: 0.0002 %<br />

Emission of nuclear fission and activation products<br />

with plant waste water (excluding tritium): 0.06 %<br />

Collective dose:<br />

0.072 Sv<br />

7<br />

6<br />

5<br />

4<br />

3<br />

2<br />

1<br />

0<br />

0.22<br />

2012<br />

0.16<br />

2013<br />

0.14<br />

2014<br />

0.15<br />

2015<br />

0.18<br />

2016<br />

0.07<br />

2017<br />

0.12 0.07<br />

2018 2019<br />

Report<br />

Operating results 2019


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

456<br />

NEWS<br />

Top<br />

Safe start-up of Unit 1 of<br />

Barakah <strong>Nuclear</strong> Energy Plant<br />

successfully achieved<br />

p Start-up is major step in process<br />

<strong>for</strong> upcoming generation of<br />

emissions-free electricity<br />

p Process undertaken in line with<br />

regulatory requirements and<br />

highest international standards <strong>for</strong><br />

nuclear quality and safety<br />

(enec) On 1 August <strong>2020</strong>, the Emirates<br />

<strong>Nuclear</strong> Energy Corporation (ENEC)<br />

announced that its operating and<br />

maintenance subsidiary, Nawah<br />

Energy Company (Nawah) has<br />

successfully started up Unit 1 of the<br />

Barakah <strong>Nuclear</strong> Energy Plant, located<br />

in the Al Dhafrah Region of Abu Dhabi,<br />

United Arab Emirates (UAE). This step<br />

is the most historic milestone to date in<br />

the delivery of the UAE Peaceful<br />

<strong>Nuclear</strong> Energy Program, as part of<br />

the process towards generating clean<br />

electricity <strong>for</strong> the Nation <strong>for</strong> at least<br />

the next 60 years.<br />

Since receipt of the Operating<br />

­License from the Federal Authority<br />

<strong>for</strong> <strong>Nuclear</strong> Regulations (FANR) in<br />

February <strong>2020</strong>, and the completion of<br />

fuel assembly loading in March <strong>2020</strong>,<br />

Nawah, the Joint Venture nuclear<br />

operations and maintenance subsidiary<br />

of ENEC and the Korea Electric<br />

<strong>Power</strong> Corporation (KEPCO), has<br />

been safely progressing through a<br />

comprehensive testing program, prior<br />

to successfully completing the start-up<br />

of the first nuclear energy reactor of<br />

the Barakah plant.<br />

The start-up of Unit 1 marks the<br />

first time that the reactor safely produces<br />

heat, which is used to create<br />

steam, turning a turbine to generate<br />

electricity. Nawah’s qualified and<br />

licensed team of nuclear operators<br />

focus on safely controlling the process<br />

and controlling the power output of<br />

the reactor. After several weeks and<br />

conducting numerous safety tests, Unit<br />

1 will be ready to connect to the UAE’s<br />

| United Arab Emirate: Barakah site during construction<br />

electricity grid, delivering the first<br />

megawatts of clean electricity to the<br />

homes and businesses of the Nation.<br />

Testing has been undertaken with the<br />

continued oversight of the UAE’s<br />

­independent nuclear regulator, FANR,<br />

and follows the World Asso­ciation of<br />

<strong>Nuclear</strong> Operator’s (WANO) completion<br />

of a Pre Start-up Review (PSUR) in<br />

January <strong>2020</strong>, prior to receipt of the<br />

Operating License, which ensures Unit<br />

1 is aligned with international best<br />

practice in the nuclear energy industry.<br />

H.E. Mohamed Ibrahim Al<br />

­Hammadi, Chief Executive Officer<br />

of ENEC, said: “Today is a truly historic<br />

moment <strong>for</strong> the UAE. It is the<br />

culmination of more than a decade of<br />

vision, strategic planning and robust<br />

program management. Despite the<br />

recent global challenges, our team has<br />

demonstrated outstanding resilience<br />

and commitment to the safe delivery<br />

of Unit 1. We are now another step<br />

closer to achieving our goal of supplying<br />

up to a quarter of our Nation’s<br />

electricity needs and powering its<br />

future growth with safe, reliable, and<br />

emissions-free electricity.<br />

“Through the realization of the<br />

vision of our Leadership, the Barakah<br />

<strong>Nuclear</strong> Energy Plant has become an<br />

engine of growth <strong>for</strong> the Nation. It will<br />

deliver 25 % of the UAE’s electricity<br />

with zero carbon emissions while also<br />

supporting economic diversification<br />

by creating thousands of high-value<br />

jobs through the establishment of<br />

a sustainable local nuclear energy<br />

industry and supply chain. We are<br />

grateful to the Leadership <strong>for</strong> their<br />

continuous support in making this<br />

remarkable achievement happen,<br />

along with the support of our UAE<br />

stakeholders and Korean partners,<br />

and congratulate everyone involved<br />

in the Program on this landmark<br />

occasion.”<br />

Once the unit is connected to the<br />

grid, the nuclear operators will<br />

continue with a process of gradually<br />

raising the power levels, known as<br />

<strong>Power</strong> Ascension Testing (PAT).<br />

Throughout, the systems of Unit 1 are<br />

continuously monitored and tested<br />

as the unit proceeds towards full<br />

electricity production in line with all<br />

regulatory requirements and the<br />

highest international standards of<br />

safety, quality and security. Once the<br />

process is completed over the course<br />

of a number of months, the plant will<br />

deliver abundant baseload electricity<br />

at full capacity to power the growth<br />

and prosperity of the UAE <strong>for</strong> decades<br />

to come.<br />

Commenting on this key milestone<br />

in UAE nuclear energy operations, Eng.<br />

Ali Al Hammadi, Chief Executive Officer<br />

of Nawah, said: “The start-up of<br />

Unit 1 is a significant milestone <strong>for</strong><br />

Nawah Energy Company as we fulfill<br />

our mandate to operate and maintain<br />

the plant in accordance with the<br />

highest international standards of safety<br />

and quality. The dedication of our<br />

people as well as our close collaboration<br />

with our Korean partners and cooperation<br />

with numerous international<br />

expert organizations has enabled this<br />

accomplishment. This ­reflects our<br />

commitment to upholding the highest<br />

safety, quality and operational transparency<br />

standards through out the entire<br />

commissioning and startup process<br />

by leveraging the expertise of the global<br />

nuclear industry.<br />

“I am especially proud of our<br />

­talented UAE National engineers and<br />

nuclear professionals who contributed<br />

to the construction of Unit 1, as<br />

well as the UAE National Senior<br />

Reactor Operators and Reactor Operators<br />

who have been certified to safely<br />

operate the plant, alongside our international<br />

experts, to ensure the safe<br />

and sustainable operations of the unit<br />

<strong>for</strong> decades to come.” concluded Eng.<br />

Ali Al Hammadi.<br />

The UAE is the first country in the<br />

Arab World, and the 33 rd nation<br />

globally, to develop a nuclear energy<br />

plant to generate safe, clean, and reliable<br />

baseload electricity. The Barakah<br />

plant is significantly contributing to<br />

the UAE’s ef<strong>for</strong>ts to move towards the<br />

electrification of its energy sector, and<br />

the decarbonization of electricity production.<br />

When fully operational, the<br />

plant will produce 5.6 gigawatts of<br />

electricity while preventing the release<br />

of more than 21 million tons of carbon<br />

emissions every year, equivalent to the<br />

removal of 3.2 million cars from the<br />

Nation’s roads annually.<br />

The UAE Peaceful <strong>Nuclear</strong> Energy<br />

Program commenced in 20<strong>09</strong>, ENEC<br />

has worked closely with international<br />

nuclear bodies, including the<br />

News


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Operating Results April <strong>2020</strong><br />

Plant name Country Nominal<br />

capacity<br />

Type<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Month Year Since<br />

commissioning<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Month Year Month Year<br />

457<br />

OL1 Olkiluoto BWR FI 910 880 699 590 640 2 562 054 272 027 524 97.04 98.54 89.17 95.35 89.17 95.93<br />

OL2 Olkiluoto BWR FI 910 880 720 652 413 2 674 197 262 038 283 100.00 100.00 99.23 99.76 98.49 100.13<br />

KCB Borssele PWR NL 512 484 720 366 399 1 474 330 169 455 764 99.47 99.01 99.05 98.89 99.58 99.45<br />

KKB 1 Beznau 1,2,6,7) PWR CH 380 365 395 151 006 992 465 131 301 285 54.86 88.80 54.63 88.75 54.86 89.94<br />

KKB 2 Beznau 6,7) PWR CH 380 365 720 273 585 1 107 912 138 404 695 100.00 100.00 100.00 99.83 100.05 100.50<br />

KKG Gösgen 7) PWR CH 1060 1010 720 761 249 3 <strong>08</strong>7 541 325 203 776 100.00 100.00 99.99 99.94 99.74 100.34<br />

CNT-I Trillo PWR ES 1066 1003 720 661 602 2 945 895 258 693 921 100.00 100.00 100.00 100.00 85.00 94.54<br />

Dukovany B1 PWR CZ 500 473 720 358 018 1 451 428 117 335 611 100.00 100.00 100.00 99.90 99.45 100.00<br />

Dukovany B2 PWR CZ 500 473 720 355 806 1 441 545 112 484 863 100.00 100.00 100.00 99.88 98.84 99.31<br />

Dukovany B3 2) PWR CZ 500 473 1 29 284 895 110 536 631 0.14 20.01 0.01 19.67 0.01 19.63<br />

Dukovany B4 2) PWR CZ 500 473 720 360 754 955 104 111 662 061 100.00 65.42 99.65 65.31 100.21 65.80<br />

Temelin B1 4) PWR CZ 1<strong>08</strong>0 1030 0 0 1 756 742 123 671 555 0 54.74 0 54.22 0 55.93<br />

Temelin B2 PWR CZ 1<strong>08</strong>0 1030 720 782 217 3 223 320 120 705 938 100.00 100.00 100.00 100.00 100.41 102.62<br />

Doel 1 2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0<br />

Doel 2 2) PWR BE 454 433 0 0 0 136 335 470 0 0 0 0 0 0<br />

Doel 3 PWR BE 1056 1006 720 775 8<strong>09</strong> 3 139 030 266 250 680 100.00 100.00 100.00 100.00 101.65 101.96<br />

Doel 4 PWR BE 1<strong>08</strong>4 1033 720 788 060 3 186 503 272 824 778 100.00 100.00 99.26 99.70 99.26 99.63<br />

Tihange 1 2) PWR BE 10<strong>09</strong> 962 0 0 0 307 547 424 0 0 0 0 0 0<br />

Tihange 2 PWR BE 1055 10<strong>08</strong> 720 754 589 3 042 824 261 <strong>09</strong>7 343 100.00 100.00 99.98 99.87 100.34 100.33<br />

Tihange 3 PWR BE 1<strong>08</strong>9 1038 720 778 353 3 142 871 283 705 448 100.00 100.00 100.00 99.97 99.91 100.03<br />

NEWS<br />

Plant name<br />

Type<br />

Nominal<br />

capacity<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Since Month Year Month Year Month Year<br />

commissioning<br />

KBR Brokdorf DWR 1480 1410 720 951 530 3 7<strong>08</strong> 586 364 429 6<strong>09</strong> 100.00 100.00 93.80 94.13 89.00 85.93<br />

KKE Emsland 4) DWR 1406 1335 720 1 003 407 3 995 435 361 595 636 100.00 100.00 100.00 100.00 99.17 97.89<br />

KWG Grohnde 2) DWR 1430 1360 266 359 894 3 265 485 391 540 331 36.96 84.36 37.84 84.58 34.73 78.17<br />

KRB C Gundremmingen 3) SWR 1344 1288 590 746 240 3 295 647 344 619 199 81.99 86.<strong>09</strong> 77.15 84.34 76.49 83.84<br />

KKI-2 Isar DWR 1485 1410 720 972 470 4 115 <strong>09</strong>0 369 877 559 100.00 100.00 99.98 99.99 90.37 95.04<br />

GKN-II Neckarwestheim DWR 1400 1310 720 962 560 3 930 560 344 168 804 100.00 100.00 100.00 99.97 95.54 96.81<br />

<strong>International</strong> Atomic Energy Agency<br />

(IAEA), and WANO, in line with the<br />

robust regulatory framework of FANR.<br />

To date, more than 255 inspections<br />

have been undertaken by FANR to<br />

ensure the Barakah plant and its<br />

people and processes meet the highest<br />

standards of nuclear quality and<br />

safety. These national reviews have<br />

been supported by more than 40 assessments<br />

and peer reviews by the<br />

IAEA and WANO.<br />

ENEC recently announced the<br />

construction completion of Unit 2,<br />

with operational readiness preparations<br />

now underway by Nawah.<br />

Construction of Units 3 and 4 of the<br />

Barakah <strong>Nuclear</strong> Energy Plant is in the<br />

final stages, with the overall construction<br />

completion of the four units now<br />

standing at 94 %.<br />

| www.enec.gov.ae (201121401)<br />

World<br />

IAEA Launches Initiative to<br />

Help Prevent Future Pandemics<br />

(iaea) The Director General of the<br />

­<strong>International</strong> Atomic Energy Agency<br />

(IAEA), Rafael Mariano Grossi,<br />

launched an initiative today to<br />

strengthen global preparedness <strong>for</strong> future<br />

pandemics like COVID-19. The<br />

project, called ZODIAC, builds on<br />

the IAEA’s experience in assisting<br />

countries in the use of nuclear and<br />

nuclear-derived techniques <strong>for</strong> the<br />

rapid detection of pathogens that<br />

cause transboundary animal diseases,<br />

including ones that spread to humans.<br />

These zoonotic diseases kill around<br />

2.7 million people every year.<br />

The IAEA Zoonotic Disease Integrated<br />

Action (ZODIAC) project will<br />

establish a global network to help<br />

national laboratories in monitoring,<br />

surveillance, early detection and control<br />

of animal and zoonotic diseases<br />

such as COVID-19, Ebola, avian influenza<br />

and Zika. ZODIAC is based on<br />

the technical, scientific and laboratory<br />

capacity of the IAEA and its partners<br />

and the Agency’s mechanisms to<br />

quickly deliver equipment and knowhow<br />

to countries.<br />

The aim is to make the world<br />

better prepared <strong>for</strong> future outbreaks.<br />

“Member States will have access<br />

to equipment, technology packages,<br />

expertise, guidance and training.<br />

Decision-makers will receive up-todate,<br />

user-friendly in<strong>for</strong>mation that<br />

will enable them to act quickly,”<br />

Mr Grossi told a meeting of the IAEA<br />

Board of Governors.<br />

Mr Grossi said COVID-19 had<br />

exposed problems related to virus<br />

*)<br />

Net-based values<br />

(Czech and Swiss<br />

nuclear power<br />

plants gross-based)<br />

1)<br />

Refueling<br />

2)<br />

Inspection<br />

3)<br />

Repair<br />

4)<br />

Stretch-outoperation<br />

5)<br />

Stretch-inoperation<br />

6)<br />

Hereof traction supply<br />

7)<br />

Incl. steam supply<br />

BWR: Boiling<br />

Water Reactor<br />

PWR: Pressurised<br />

Water Reactor<br />

Source: VGB<br />

News


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Operating Results May <strong>2020</strong><br />

458<br />

Plant name Country Nominal<br />

capacity<br />

Type<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Month Year Since<br />

commissioning<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Month Year Month Year<br />

NEWS<br />

OL1 Olkiluoto 1,4) BWR FI 910 880 570 485 956 3 048 010 272 513 479 76.61 94.07 71.00 90.38 71.00 90.84<br />

OL2 Olkiluoto 1) BWR FI 910 880 476 414 256 3 <strong>08</strong>8 454 262 452 539 63.97 92.65 60.56 91.76 60.52 92.05<br />

KCB Borssele 1,4) PWR NL 512 484 690 329 877 1 804 207 169 785 641 86.69 96.50 86.76 96.42 86.43 96.79<br />

KKB 1 Beznau 1,2,7) PWR CH 380 365 228 82 670 1 075 135 131 383 955 30.65 76.94 29.21 76.60 28.74 77.45<br />

KKB 2 Beznau 7) PWR CH 380 365 744 282 479 1 390 391 138 687 174 100.00 100.00 100.00 99.87 99.93 100.39<br />

KKG Gösgen 4,7) PWR CH 1060 1010 744 783 030 3 870 571 325 986 806 100.00 100.00 100.00 99.95 99.29 100.12<br />

CNT-I Trillo 1,2) PWR ES 1066 1003 411 416 266 3 362 161 259 110 187 55.19 90.86 52.06 90.22 51.99 85.86<br />

Dukovany B1 PWR CZ 500 473 744 368 <strong>08</strong>5 1 819 513 117 703 696 100.00 100.00 100.00 99.92 98.95 99.78<br />

Dukovany B2 PWR CZ 500 473 744 365 973 1 807 518 112 850 837 100.00 100.00 100.00 99.90 98.38 99.12<br />

Dukovany B3 PWR CZ 500 473 744 354 979 639 874 110 891 610 100.00 36.33 96.72 35.39 95.42 35.<strong>09</strong><br />

Dukovany B4 PWR CZ 500 473 744 370 278 1 325 381 112 032 339 100.00 72.47 100.00 72.39 99.54 72.68<br />

Temelin B1 PWR CZ 1<strong>08</strong>0 1030 4<strong>08</strong> 399 882 2 156 624 124 071 437 54.84 54.76 50.<strong>08</strong> 53.37 49.67 54.65<br />

Temelin B2 PWR CZ 1<strong>08</strong>0 1030 744 800 038 4 023 358 121 505 976 100.00 100.00 99.84 99.97 99.38 101.96<br />

Doel 1 2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0<br />

Doel 2 2) PWR BE 454 433 42 4 360 4 360 136 339 830 5.78 1.18 1.16 0.23 1.16 0.24<br />

Doel 3 PWR BE 1056 1006 744 798 061 3 937 <strong>09</strong>1 267 048 741 100.00 100.00 100.00 100.00 101.19 101.80<br />

Doel 4 PWR BE 1<strong>08</strong>4 1033 744 8<strong>08</strong> 168 3 994 671 273 632 946 100.00 100.00 99.40 99.64 98.57 99.41<br />

Tihange 1 2) PWR BE 10<strong>09</strong> 962 0 0 0 307 547 424 0 0 0 0 0 0<br />

Tihange 2 PWR BE 1055 10<strong>08</strong> 744 772 511 3 815 335 261 869 854 100.00 100.00 99.96 99.89 99.38 100.13<br />

Tihange 3 PWR BE 1<strong>08</strong>9 1038 744 799 791 3 942 662 284 505 238 100.00 100.00 100.00 99.97 99.27 99.87<br />

Plant name<br />

Type<br />

Nominal<br />

capacity<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Since Month Year Month Year Month Year<br />

commissioning<br />

KBR Brokdorf DWR 1480 1410 744 1 001 313 4 7<strong>09</strong> 899 365 430 922 100.00 100.00 94.33 94.17 90.70 86.90<br />

KKE Emsland 1,2,4) DWR 1406 1335 202 262 675 4 258 110 361 858 311 27.21 85.15 26.35 84.98 25.05 83.03<br />

KWG Grohnde 2) DWR 1430 1360 188 243 479 3 5<strong>08</strong> 965 391 783 810 98.32 87.21 96.87 87.<strong>08</strong> 22.69 66.85<br />

KRB C Gundremmingen SWR 1344 1288 744 996 571 4 292 218 345 615 771 100.00 88.93 100.00 87.54 98.90 86.91<br />

KKI-2 Isar DWR 1485 1410 744 1 003 559 5 118 649 370 881 118 100.00 100.00 100.00 99.99 90.23 94.06<br />

GKN-II Neckarwestheim DWR 1400 1310 744 1 025 940 4 956 500 345 194 744 100.00 100.00 100.00 99.98 98.70 97.19<br />

detection capabilities in many countries,<br />

as well as a need <strong>for</strong><br />

better communication between health<br />

institutions around the world. While<br />

the IAEA has been doing important<br />

work to help countries in these areas,<br />

such as through the provision of<br />

­COVID-19 tests, he said it was “essential<br />

to pull these diverse strands<br />

together into a coherent and comprehensive<br />

framework of assistance”.<br />

<strong>Nuclear</strong>-derived techniques, such<br />

as tests using real time reverse<br />

transcription–polymerase chain reaction<br />

(RT-PCR), are important tools in<br />

the detection and characterization<br />

of viruses. The IAEA is providing<br />

emergency assistance to some 120<br />

countries in the use of such tests to<br />

rapidly detect COVID-19.<br />

Zoonotic diseases are caused by<br />

bacteria, parasites, fungi or viruses<br />

that originate in animals and can be<br />

transmitted to humans. Many of these<br />

diseases are treatable if medication is<br />

available, such as E. coli- and brucella<br />

bacterial infections. But others<br />

have the potential to severely affect<br />

­humans, such as Ebola, SARS and<br />

COVID-19.<br />

ZODIAC builds on the experience<br />

of VETLAB, a network of veterinary<br />

laboratories in Africa and Asia that<br />

was originally set up by the Food<br />

and Agriculture Organization of the<br />

­United Nations (FAO) and the IAEA to<br />

combat the cattle disease rinderpest.<br />

VETLAB now supports countries in<br />

the early detection of several zoonotic<br />

and animal diseases, such as African<br />

swine fever and pest des petit ruminants<br />

(PPR).<br />

“About 70 per cent of all diseases in<br />

humans come from animals,” said<br />

Gerrit Viljoen, Head of the Animal<br />

Production and Health Section of<br />

the Joint FAO/IAEA Programme <strong>for</strong><br />

<strong>Nuclear</strong> Techniques in Food and<br />

­Agriculture.<br />

ZODIAC aims to help veterinary<br />

and public health officials identify<br />

these diseases be<strong>for</strong>e they spread.<br />

“We have seen an increase in the number<br />

of zoonotic epidemics in the last<br />

decades: first Ebola, then Zika, and<br />

now COVID-19. It’s important to monitor<br />

what is in the animal kingdom –<br />

both wildlife and livestock – and to<br />

act quickly on those findings be<strong>for</strong>e<br />

the pathogens jump to humans,”<br />

Mr ­Viljoen said.<br />

Following the One Health concept<br />

<strong>for</strong> a multidisciplinary collaborative<br />

approach between human and animal<br />

health authorities and specialists,<br />

­ZODIAC will benefit from the unique<br />

News


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

joint FAO/IAEA laboratories and from<br />

partners such as the World Health Organization<br />

(WHO) and the World Organisation<br />

<strong>for</strong> Animal Health (OIE).<br />

“We have a unique capacity to provide<br />

laboratory support and guidance<br />

to countries,” said Mr Viljoen, adding<br />

that ZODIAC will, <strong>for</strong> example, provide<br />

technical know-how and advice<br />

to laboratories on test per<strong>for</strong>mance<br />

and assist authorities in the interpretation<br />

of results and in devising containment<br />

measures.<br />

ZODIAC will also support R&D<br />

activities <strong>for</strong> novel technologies and<br />

methodologies <strong>for</strong> early detection and<br />

surveillance. Under the project, the<br />

IAEA will enhance its capacities to<br />

host scientists and fellows from<br />

Member States at its Seibersdorf<br />

­laboratories outside Vienna and to<br />

carry out research on immunological,<br />

molecular, nuclear and isotopic tests,<br />

as well as in the use of irradiation to<br />

develop vaccines against diseases<br />

such as avian influenza.<br />

| www.iaea.org (201711216)<br />

Research<br />

TUM FRMII: Improved welding<br />

process <strong>for</strong> turbine parts<br />

(tum) Scientists at the Heinz Maier-<br />

Leibnitz Zentrum (MLZ) use imaging<br />

techniques with neutrons to study<br />

specially bonded steel components.<br />

With their results they can improve<br />

the welding process <strong>for</strong> oil and gas<br />

pipelines and turbines.<br />

The “transient liquid-phase<br />

bonding” (TLPB) process is used <strong>for</strong><br />

joining metallic systems where<br />

standard welding methods cannot be<br />

used. For example in the repair of gas<br />

pipelines. By using special materials,<br />

the TLPB process achieves very good<br />

mechanical properties.<br />

Researchers at MLZ have now<br />

taken a closer look at this type of<br />

welded joint. With their method they<br />

can non-destructively test the quality<br />

of TLPB joints. In doing so, they<br />

discover harmful inclusions at the<br />

fusion joint. This is particularly important<br />

<strong>for</strong> the oil and gas industry, where<br />

the TLPB method is used particularly<br />

often. It is also used in the manufacture<br />

of turbine blades in jet engines.<br />

Neutrons make the joining<br />

seam visible<br />

Scientists Dr. Nicolás Di Luozzo and<br />

Dr. Marcelo Fontana from the Universidad<br />

de Buenos Aires (UBA) together<br />

with MLZ scientist Dr. Michael Schulz<br />

(Technical University Munich) have<br />

investigated TLP-bonded steel components<br />

with boron-alloyed foils.<br />

For this purpose they used neutron radiography<br />

and tomography at the instrument<br />

ANTARES.<br />

The neutrons make the “seam”<br />

between the metal parts, the microstructure<br />

of the compound and the<br />

subsequent base metal, visible. “With<br />

existing microscopic methods it is not<br />

possible to quantify the boron concentrations<br />

in a steel matrix,” Di Luozzo<br />

explains, “quite contrary to neutron<br />

radiography. The spatial resolution<br />

achieved by ANTARES is remarkable.”<br />

Researchers identify flaws in the<br />

welding process<br />

Within the joint, Di Luozzo and his<br />

colleagues identify two areas: they<br />

distinguish where all solidification<br />

phases are completed and where they<br />

are not. In case of incomplete solidification,<br />

the scientists find borides in<br />

addition to ferrite. This is an indicator<br />

of weak points in the joint: the TLPB<br />

process did not succeed perfectly at<br />

these points and the joint is defective.<br />

Using neutron tomography, the<br />

scientists were able to reconstruct the<br />

size, quantity and exact location of<br />

these weak points in three-dimensional<br />

<strong>for</strong>m. Here they still have to<br />

improve the welding process in order<br />

to obtain a perfect seam.<br />

Non-destructive testing thanks<br />

to neutrons<br />

In their experiment, the scientists<br />

showed that neutrons are suitable <strong>for</strong><br />

making borides visible on welding<br />

seams. This helps to assess the quality<br />

of TLPB welded joints. “The borides at<br />

the joint have an extremely negative<br />

influence on its mechanical properties,”<br />

explains Di Luozzo. In the worst<br />

case, this means that the weld is brittle<br />

and breaks, and the joined components<br />

no longer hold together.<br />

In theory, the TLPB process enables<br />

very good mechanical properties. “It is<br />

possible <strong>for</strong> joined materials to achieve<br />

properties comparable to those of the<br />

base material,” emphasizes Di Luozzo.<br />

The product is there<strong>for</strong>e as strong as if<br />

it were made in one piece. That's why<br />

the process is so important: the<br />

neutrons can detect defective spots<br />

in the welded end product – and in<br />

the case of an aircraft engine, this is<br />

vital <strong>for</strong> passengers' survival.<br />

Original publication:<br />

Di Luozzo, N., Schulz, M. & Fontana,<br />

M. Imaging of boron distribution<br />

in steel with neutron radiography<br />

and tomography. J Mater Sci 55,<br />

7927–7937 (<strong>2020</strong>). DOI: 10.1007/<br />

s1<strong>08</strong>53-020-04556-z<br />

| www.frm2.tum.de (202321930)<br />

Company News<br />

Framatome and Lockheed<br />

Martin join <strong>for</strong>ces to provide<br />

additional solution <strong>for</strong> US nuclear<br />

plant instrumentation<br />

and control<br />

(framatome) Framatome and Lockheed<br />

Martin recently signed a teaming<br />

agreement that will integrate Lockheed<br />

Martin’s Discrete Logic Solving<br />

System (DLSS) into proven Framatome<br />

instrumentation and control (I&C)<br />

nuclear plant modernization solutions.<br />

This additional analog solution combines<br />

the companies’ technologies and<br />

supports the safety and reliability of<br />

nuclear power plants.<br />

The I&C system serves as part of<br />

the plant’s “central nervous system.”<br />

It provides operators with critical<br />

in<strong>for</strong>mation on plant operation,<br />

allows them to control various plant<br />

safety systems during routine operations<br />

and automatically protects the<br />

reactor if needed.<br />

“With Lockheed Martin’s Discrete<br />

Logic Solving System now part of our<br />

I&C portfolio, U.S. customers will have<br />

access to additional analog-based<br />

safety solution options <strong>for</strong> upgrading<br />

their existing equipment,” said Clayton<br />

Scott, senior vice president global<br />

sales and deputy <strong>for</strong> the I&C Business<br />

Unit at Framatome. “While Fram atome<br />

focuses on helping our U.S. customers<br />

transition to digital I&C systems, it’s<br />

important that we con tinue to serve<br />

our non-digital customers with safe<br />

and reliable solutions.”<br />

DLSS is one of Lockheed Martin’s<br />

non-digital I&C solutions offering<br />

effective applications <strong>for</strong> nuclear<br />

systems that currently employ analog<br />

technology. It supports operational<br />

measures of nuclear plant systems<br />

by monitoring, calculating and activating<br />

plant interfaces.<br />

“Lockheed Martin is proud to<br />

partner with Framatome in support of<br />

incorporating DLSS on future nuclear<br />

energy I&C modernization ef<strong>for</strong>ts,”<br />

said John Pericci, Lockheed Martin<br />

<strong>Nuclear</strong> Systems & Solutions program<br />

director. “For more than 60 years, we<br />

have provided safety critical nuclear<br />

I&C systems to commercial and U.S.<br />

government customers, enhancing<br />

operation and reliability in the<br />

industry.”<br />

459<br />

NEWS<br />

News


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

460<br />

NEWS<br />

In 2019, Framatome and Lockheed<br />

Martin partnered to complete the first<br />

installation and site acceptance testing<br />

of a new Engineered Safety<br />

­Features Actuation System (ESFAS)<br />

replacement that included DLSS. A<br />

second installation at the same plant<br />

took place earlier this year. Both<br />

systems are fully operational and<br />

meeting all requirements.<br />

| www.framatome.com (201711204)<br />

Westinghouse program<br />

awarded £10m from<br />

UK Government advanced<br />

modular reactor project<br />

(westinghouse) Westinghouse Electric<br />

Company today announced their<br />

Lead-cooled Fast Reactor (LFR) program<br />

has successfully progressed to<br />

Phase 2 of the UK Government’s<br />

Department <strong>for</strong> Business, Energy<br />

and Industrial Strategy’s (BEIS)<br />

­Advanced Modular Reactor (AMR)<br />

Feasibility and Development project,<br />

receiving £10m ($12.5m) in funding<br />

from the BEIS Energy Innovation<br />

Portfolio. The funding will help<br />

to advance nuclear technology<br />

through innovation in order to deliver<br />

reliable, clean energy <strong>for</strong> future<br />

generations.<br />

As part of Phase 2, Westinghouse, in<br />

collaboration with industry, research<br />

centres and academic partners, will<br />

utilise the funding to undertake applied<br />

research and development activities.<br />

The award will be used to demonstrate<br />

LFR components and accelerate the<br />

development of high-temperature<br />

materials, advanced manufacturing<br />

technologies and modular construction<br />

strategies <strong>for</strong> the LFR.<br />

“Our progression to Phase 2 builds<br />

on our eighty-year history in the<br />

UK as a Strategic National Asset,” said<br />

Patrick Fragman, Westinghouse president<br />

and chief executive officer. “This<br />

is the perfect combination of reducing<br />

the cost of electricity and maintaining<br />

a leading edge of science, research<br />

and innovation <strong>for</strong> the UK.”<br />

The Westinghouse LFR, a 450 MWe-­<br />

class Generation IV reactor design, has<br />

the potential to have a trans<strong>for</strong>mative<br />

effect on the cost and market flexibility<br />

of new nuclear. The key features of<br />

the Westinghouse LFR include a<br />

­simplified design, flexible operations<br />

and fuel cycle capabilities, zero CO 2<br />

emissions, walk-away safety features<br />

and modular assembly. The Westinghouse<br />

LFR will also achieve a competitive<br />

Levelised Cost of Electricity<br />

(LCoE) to ensure economic competitiveness<br />

in the most challenging<br />

global electricity markets.<br />

As part of the AMR project, some of<br />

the development facilities will be<br />

established at the Clean Energy<br />

­Technology Park at Springfields. The<br />

Clean Energy Technology Park is<br />

contributing towards the UK’s Net<br />

Zero ambitions by leveraging the<br />

existing strategic national asset,<br />

Springfields, to support innovation<br />

and collaborative partnerships, whilst<br />

providing opportunities <strong>for</strong> bringing<br />

highly-skilled jobs to the North West<br />

of England and significant economic<br />

benefits to the UK.<br />

Westinghouse will deliver the<br />

Phase 2 program in collaboration with<br />

Ansaldo <strong>Nuclear</strong>e and ENEA, in<br />

addition to Bangor University, Frazer-<br />

Nash Consultancy, Jacobs, National<br />

<strong>Nuclear</strong> Laboratory (NNL), <strong>Nuclear</strong><br />

Advanced Manufacturing Research<br />

Centre (NAMRC), the University<br />

of Cambridge, the University of<br />

­Manchester and Vacuum Process<br />

­Engineering, Inc. (VPE).<br />

| www.westinghousenuclear.com<br />

(201711200)<br />

Orano acquires KSE and<br />

strengthens its position<br />

in industrial maintenance<br />

(orano) Orano announces the acquisition<br />

of the company KSB Service<br />

Energie (KSE) and of its subsidiaries<br />

KSB Service Cotumer (KSC) and the<br />

Société de Travaux et d’Ingénierie<br />

Industrielle (STII) from the German<br />

group KSB, a global player in the<br />

manufacturing of industrial pumps<br />

and valves.<br />

This acquisition, effective as of July<br />

1st, is part of Orano's strategy to<br />

develop its service activities, in<br />

particular in the area of industrial<br />

maintenance. KSE and its subsidiaries<br />

KSC and STII are recognized <strong>for</strong> the<br />

role they play in providing services<br />

to the French nuclear fleet and to<br />

the industry, whether it be carrying<br />

out interventions on valve systems,<br />

mechanical maintenance on rotating<br />

machines or boilermaking services<br />

(anchor points, supports, piping, etc.).<br />

With this acquisition, Orano<br />

enhances its service offer with new<br />

specialized resources which complement<br />

the nuclear maintenance<br />

activities in which the group is already<br />

present. More than 250 employees<br />

and the industrial capacities of<br />

the three entities will be joining<br />

<strong>for</strong>ces with Orano's “Dismantling<br />

and Services” unit, whose 1,600 employees<br />

already work on the French<br />

nuclear fleet on a daily basis in<br />

industrial logistics, site support and<br />

maintenance.<br />

Philippe Knoche, Chief Executive<br />

Officer of Orano, declared: “I ­welcome<br />

our new colleagues from KSE, STII<br />

and KSC to the Orano group. Their<br />

arrival is concrete evidence of our<br />

ambition to develop in the area of<br />

service activities and of our confidence<br />

in the future of nuclear, <strong>for</strong><br />

which needs in maintenance will<br />

remain strong over the years to come.”<br />

Alain Vandercruyssen, Senior<br />

­Executive Vice President in charge of<br />

Orano’s “Dismantling and Services”<br />

unit, added: “with this transaction,<br />

Orano rein<strong>for</strong>ces its expertise and its<br />

qualifications to achieve the critical<br />

mass necessary to become a major<br />

player in maintenance <strong>for</strong> EDF. The<br />

contribution that KSE and its subsidiaries<br />

bring to the table also consolidates<br />

our presence with nuclear<br />

operators on international markets.”<br />

| www.orano.group (202330732)<br />

People<br />

Daniel Oehr takes over from<br />

Dr Hannes Wimmer<br />

as Chairman of the<br />

GNS Management Board<br />

(gns) After nine very successful years<br />

as Chairman of the Management<br />

Board of GNS Gesellschaft für<br />

Nuklear- Service mbH, Dr Hannes<br />

Wimmer (56) will leave GNS at the<br />

end of this year by best mutual agreement<br />

with the shareholders.<br />

“On behalf of all shareholders, I<br />

would like to thank Hannes Wimmer<br />

<strong>for</strong> his great commitment, especially in<br />

the internationalisation and repositioning<br />

of GNS over the past years.<br />

With the spin-off of the interim storage<br />

and final disposal activities and the<br />

successful acquisition of Höfer &<br />

Bechtel as well as Eisenwerk Bassum,<br />

the management team with Mr.<br />

Wimmer has succeeded in developing<br />

GNS into a renowned cask manufacturer<br />

and service provider in the field<br />

of nuclear waste management,” says<br />

Dr. Guido Knott, Chairman of the<br />

Supervisory Board of GNS.<br />

Wimmer's successor will be<br />

Daniel Oehr (43), currently Head<br />

of Controlling and Per<strong>for</strong>mance<br />

Management at GNS shareholder<br />

Preussen Elektra GmbH, who has been<br />

familiar with GNS <strong>for</strong> many years.<br />

Daniel Oehr will join the management<br />

board on November 1, <strong>2020</strong> and will<br />

take over the tasks of Dr. Wimmer as<br />

of December 1, <strong>2020</strong>, in particular as<br />

chairman of the management board<br />

and CEO.<br />

News


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

Uranium<br />

Prize range: Spot market [USD*/lb(US) U 3 O 8 ]<br />

140.00<br />

120.00<br />

) 1<br />

Uranium prize range: Spot market [USD*/lb(US) U 3 O 8 ]<br />

140.00<br />

120.00<br />

) 1<br />

461<br />

100.00<br />

100.00<br />

80.00<br />

60.00<br />

40.00<br />

20.00<br />

0.00<br />

1980<br />

Jan. 20<strong>09</strong><br />

Yearly average prices in real USD, base: US prices (1982 to1984) *<br />

Jan. 2010<br />

1985<br />

Jan. 2011<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

1990<br />

Jan. 2012<br />

Jan. 2013<br />

1995<br />

Jan. 2014<br />

2000<br />

Jan. 2015<br />

Jan. 2016<br />

2005<br />

Jan. 2017<br />

2010<br />

Jan. 2018<br />

Jan. 2019<br />

2015<br />

Jan. <strong>2020</strong><br />

<strong>2020</strong><br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> <strong>2020</strong><br />

Year<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

) 1 ) 1<br />

prices, * Actual nominal USD not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> <strong>2020</strong><br />

| Uranium spot market prices from 1980 to <strong>2020</strong> and from 20<strong>09</strong> to <strong>2020</strong>. The price range is shown.<br />

In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.<br />

Separative work: Spot market price range [USD*/kg UTA]<br />

Conversion: Spot conversion price range [USD*/kgU]<br />

180.00<br />

26.00<br />

24.00<br />

160.00<br />

22.00<br />

140.00<br />

120.00<br />

100.00<br />

80.00<br />

60.00<br />

40.00<br />

20.00<br />

0.00<br />

Jan. 2021<br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> <strong>2020</strong><br />

80.00<br />

60.00<br />

40.00<br />

20.00<br />

0.00<br />

20.00<br />

18.00<br />

16.00<br />

14.00<br />

12.00<br />

10.00<br />

Jan. 20<strong>09</strong><br />

8.00<br />

6.00<br />

4.00<br />

2.00<br />

0.00<br />

Jan. 20<strong>09</strong><br />

Jan. 2010<br />

Jan. 2010<br />

Jan. 2011<br />

Jan. 2011<br />

Jan. 2012<br />

Jan. 2012<br />

Jan. 2013<br />

Jan. 2013<br />

Jan. 2014<br />

Jan. 2014<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

Jan. 2015<br />

Jan. 2015<br />

Jan. 2016<br />

Jan. 2016<br />

Jan. 2017<br />

Jan. 2017<br />

Jan. 2018<br />

Jan. 2018<br />

Jan. 2019<br />

Jan. 2019<br />

Jan. <strong>2020</strong><br />

Jan. <strong>2020</strong><br />

Jan. 2021<br />

Jan. 2021<br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> <strong>2020</strong><br />

NEWS<br />

| Separative work and conversion market price ranges from 20<strong>09</strong> to <strong>2020</strong>. The price range is shown.<br />

)1<br />

In December 20<strong>09</strong> Energy Intelligence changed the method of calculation <strong>for</strong> spot market prices. The change results in virtual price leaps.<br />

* Actual nominal USD prices, not real prices referring to a base year<br />

Sources: Energy Intelligence, Nukem; Bilder/Figures: <strong>atw</strong> <strong>2020</strong><br />

“The shareholders are pleased to<br />

have won Daniel Oehr <strong>for</strong> this<br />

challenging task. With him in the<br />

lead, GNS will be further developed<br />

into a quality and customer-oriented<br />

service provider with increasing<br />

business in third markets.”<br />

In addition, Sascha Bechtel (49)<br />

will, at the request of the shareholders,<br />

support the GNS management<br />

in the repositioning of GNS and<br />

will be responsible <strong>for</strong> LLW-ILW<br />

residues and waste and disposal projects.<br />

He will continue to per<strong>for</strong>m his<br />

duties as Managing Director of Höfer<br />

& Bechtel GmbH.<br />

| www.gns.de<br />

Market data<br />

(All in<strong>for</strong>mation is supplied without<br />

guarantee.)<br />

<strong>Nuclear</strong> Fuel Supply<br />

Market Data<br />

In<strong>for</strong>mation in current (nominal)<br />

­U.S.-$. No inflation adjustment of<br />

prices on a base year. Separative work<br />

data <strong>for</strong> the <strong>for</strong>merly “secondary<br />

market”. Uranium prices [US-$/lb<br />

U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 =<br />

0.385 kg U]. Conversion prices [US-$/<br />

kg U], Separative work [US-$/SWU<br />

(Separative work unit)].<br />

2017<br />

p Uranium: 19.25–26.50<br />

p Conversion: 4.50–6.75<br />

p Separative work: 39.00–50.00<br />

2018<br />

p Uranium: 21.75–29.20<br />

p Conversion: 6.00–14.50<br />

p Separative work: 34.00–42.00<br />

2019<br />

January to June 2019<br />

p Uranium: 23.90–29.10<br />

p Conversion: 13.50–18.00<br />

p Separative work: 41.00–49.00<br />

July to December 2019<br />

p Uranium: 24.50–26.25<br />

p Conversion: 18.00–23.00<br />

p Separative work: 47.00–52.00<br />

<strong>2020</strong><br />

January <strong>2020</strong><br />

p Uranium: 24.10–24.90<br />

p Conversion: 22.00–23.00<br />

p Separative work: 48.00–51.00<br />

February <strong>2020</strong><br />

p Uranium: 24.25–25.00<br />

p Conversion: 22.00–23.00<br />

p Separative work: 45.00–53.00<br />

March <strong>2020</strong><br />

p Uranium: 23.05–27.40<br />

p Conversion: 21.50–23.50<br />

p Separative work: 45.00–52.00<br />

April <strong>2020</strong><br />

p Uranium: 27.50–34.00<br />

p Conversion: 21.50–23.50<br />

p Separative work: 45.00–52.00<br />

May <strong>2020</strong><br />

p Uranium: 33.50–34.50<br />

p Conversion: 21.50–23.50<br />

p Separative work: 48.00–52.00<br />

June <strong>2020</strong><br />

p Uranium: 33.00–33.50<br />

p Conversion: 21.50–23.50<br />

p Separative work: 49.00–52.00<br />

| Source: Energy Intelligence<br />

www.energyintel.com<br />

News


<strong>atw</strong> Vol. 65 (<strong>2020</strong>) | Issue 8/9 ı August/September<br />

462<br />

Europe Can’t Discard <strong>Nuclear</strong> Investment<br />

in Quest <strong>for</strong> a Clean Energy Future<br />

NUCLEAR TODAY<br />

John Shepherd is<br />

editor-in-chief of the<br />

online publication<br />

New Energy 360 &<br />

World Battery News.<br />

Sources:<br />

IEA Energy<br />

Policy Review –<br />

https://bit.ly/39ZY4kJ<br />

Foratom report –<br />

https://bit.ly/2DITQBK<br />

DFC announcement –<br />

https://bit.ly/2XuxICk<br />

Of all the quirky expressions that creep into our conversations, a personal favourite is ‘throwing the baby out with the<br />

bathwater’. For anyone unfamiliar with the idiom, it means to discard something considered valuable or important<br />

while disposing of something that is thought to be worthless – such as an outdated idea or <strong>for</strong>m of behaviour.<br />

I’m told that the expression has its origins in German but,<br />

in the interests of accuracy, I should point out I have no<br />

hard facts to back that up.<br />

Nevertheless, the idiom does serve to underline the<br />

striking actions of those who seek to abandon all that is<br />

good <strong>for</strong> what is, arguably, a short-term gain.<br />

I’m thinking here of a recent study by the Paris-based<br />

<strong>International</strong> Energy Agency (IEA), which takes a<br />

clear-eyed look at the state of Europe’s energy policies and<br />

ambitions <strong>for</strong> the future energy mix. The ‘bottom line’ of<br />

the analysis should be a wake-up call <strong>for</strong> those in charge<br />

of mapping out Europe’s energy future.<br />

According to the IEA’s ‘Energy Policy Review <strong>for</strong><br />

Europe’, nuclear power capacity in the European<br />

Union could fall to 5 % by 2040 “without new policy ­<br />

action” at the level of the bloc’s national states. The<br />

nuclear power community has known <strong>for</strong> a long time<br />

that it is regarded as a Cinderella industry in the eyes of<br />

some states.<br />

The IEA’s study underlines just how blurred the<br />

European Commission’s long-term vision is. For example,<br />

nuclear in the EU is still expected to contribute to 15 % of<br />

electricity production in 2050. But that cannot be a tenable<br />

proposition if nuclear is allowed to wither on the energy<br />

infrastructure vine.<br />

As the IEA points out, Europe’s nuclear reactor fleet is<br />

ageing, with many nuclear plants being taken out of<br />

service and only a few reactors under construction and<br />

several planned. There<strong>for</strong>e, without a thorough review of<br />

policy planning, there could be serious implications “not<br />

only <strong>for</strong> the cost of electricity but also the security of supply<br />

at a regional level”.<br />

And if the message is still not getting through, the IEA<br />

spells the situation out in plain English: “To keep the<br />

nuclear energy option open <strong>for</strong> 2030 and beyond, the EU<br />

needs to maintain a level playing field <strong>for</strong> the financing of<br />

nuclear, to support lifetime extensions and new plants in<br />

countries where nuclear is accepted, and foster safety and<br />

waste disposal <strong>for</strong> the decommissioning of existing plants.”<br />

The corridors of power in Brussels do not have an<br />

­enviable reputation <strong>for</strong> being fans of ‘level playing fields’ –<br />

in other words, fairness – when it comes to incentivising<br />

nuclear energy investment.<br />

However, the IEA warned that the EU is “not yet on<br />

track” towards its targeted increase of a renewables share<br />

of up to 32 % (which was at 18 % in 2018) by 2030. Nor is<br />

the EU on course to achieve its wish of energy efficiency<br />

savings of 32.5 % by 2030. In fact, the IEA points out<br />

that today’s 2030 targets “will require a significant ­<br />

system trans<strong>for</strong>mation, even more so with the announced<br />

enhanced targets under the European Green Deal”.<br />

What is needed is to put dogma aside and to ensure the<br />

functioning of the internal energy market and the level<br />

playing field <strong>for</strong> energy technology development,<br />

­investment and sustainable financing in the EU to keep<br />

all technology options open <strong>for</strong> achieving net-zero<br />

emissions.<br />

There could be the faint breeze of a change in thinking<br />

emerging in at least one as yet staunchly non-nuclear EU<br />

state – the Republic of Ireland.<br />

Senator Ned O’Sullivan, a member of one of the<br />

governing parties in the country’s newly-<strong>for</strong>med coalition,<br />

is calling <strong>for</strong> a national debate on the introduction of<br />

nuclear power. The senator said in a newspaper interview<br />

that Ireland’s “over-reliance on wind power is turning our<br />

beautiful rural landscape into a <strong>for</strong>est of ugly windmills”.<br />

According to O’Sullivan, Irish “resistance to nuclear<br />

energy is not, in my opinion, anything like as strong as it<br />

was” in the past. And he pointed out that Ireland is<br />

“ increasingly interconnected to the British grid, so we are<br />

in fact importing energy from mixed sources, including<br />

nuclear”.<br />

“We are now in the era of smaller, modular nuclear<br />

reactors. They are safer, they create less waste, some<br />

models can regenerate power from their own waste, and<br />

they are less costly and quicker to build,” he said.<br />

The senator’s Fianna Fáil party shares government with<br />

the Greens, so any attempts to tilt towards nuclear are<br />

unlikely to result in change any time soon. But O’Sullivan’s<br />

comments indicate that nuclear might yet find fresh<br />

support in the unlikeliest of quarters in Europe.<br />

Meanwhile, contrast the current EU approach with that<br />

of the US, where the <strong>International</strong> Development Finance<br />

Corporation (DFC) confirmed it was lifting its prohibition<br />

on funding nuclear energy projects.<br />

The DFC – America’s development bank – said the<br />

decision “recognises the vast energy needs of developing<br />

countries as well as new and advanced technologies such<br />

as small modular reactors and microreactors that could be<br />

particularly impactful in these markets”.<br />

DFC said its new policy stance on nuclear energy “will<br />

help deliver a zero-emission, reliable, and secure power<br />

source to developing countries” and support competition<br />

in the US nuclear sector.<br />

The president and chief executive officer of the US<br />

<strong>Nuclear</strong> Energy Institute, Maria Korsnick, said the decision<br />

would “enable US nuclear exports to compete on a<br />

more level playing field against state-owned rivals from<br />

countries such as Russia and China”.<br />

The EU could perhaps learn something from the US<br />

move. Here, we have the European Investment Bank (EIB),<br />

which could be a channel of support <strong>for</strong> nuclear industry<br />

suppliers seeking to export their goods and talents beyond<br />

Europe – and <strong>for</strong> the benefit of EU jobs and economies.<br />

The EIB has provided finance to increase the safety of<br />

operating nuclear units in the EU, so why not extend such<br />

assistance to support exports of equipment and knowhow<br />

<strong>for</strong> beyond European shores? If the EU cannot bring itself<br />

to level the playing field <strong>for</strong> nuclear energy here at home,<br />

at least give the industry a helping hand to compete (on a<br />

level playing field) internationally.<br />

Author<br />

John Shepherd<br />

<strong>Nuclear</strong> Today<br />

Europe Can’t Discard <strong>Nuclear</strong> Investment in Quest <strong>for</strong> a Clean Energy Future ı John Shepherd


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