atw International Journal for Nuclear Power 2020-08/09

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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

nucmag.com

2020

8/9

ISSN · 1431-5254

24.– €

Current Status and

Prospects of Nuclear Power

Plant Decommissioning in

the Republic of Korea

Actual Research and Development

Activities in the Field of Dismantling

Neutronic Study of CAREM-25

Advanced Small Modular Reactor

Using Monte Carlo Simulation


Competence for

Nuclear Services

Operational Waste and D&D

Spent Fuel Management

Nuclear Casks

Calculation Services and Consulting

Waste Processing Systems and Engineering

GNS Gesellschaft für Nuklear-Service mbH

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


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Implications of the Corona/Covid-19 Crises

on Energy Supply and Nuclear Power

Dear Reader, What began in the early days of this year in Wuhan, China as a local outbreak of a newer form of

Corona virus has now and for months on end taken hold of the world.

Our lives are affected by the Corona virus like hardly any

other crisis before, in private as well as in economy, society,

social life and politics. Restrictions off our daily freedom of

movement are directly noticeable, and economic life has

been badly hit with slumps in gross national product of up

to 13 % in individual OECD countries and around 5 % of

the entire global economy.

The energy supply is also affected. This is clearly

­reflected in the global oil prices. At the beginning of the

year, around US$ 52 per barrel was still being quoted on

the spot market for the US grade WTI (West Texas Intermediate).

In the course of the Corona crisis, the price

slumped to US$ 13 and even reached negative values.

Currently, the price of crude oil has stabilized at around

US$ 40. The International Energy Agency (IEA) notes a

year-on-year decline in global crude oil demand from

8.1 million barrels a day to currently around 91.9 million

barrels a day, or around minus 9 percent. Electricity

consumption has fallen dramatically in some countries,

particularly in times of complete lockdown and the massive

cutback in all industrial production. In the UK and India,

the year-on-year decline in consumption was thus almost

30 percent, and overall in most OECD countries average

demand was at the lower level of the comparable Sundays

in the previous year, i.e. around 10 percent lower. With the

gradual easing of the complete lockdown from mid-2020

onwards, electricity demand rose again and in August

reached a level around 5 % below the previous year’s

­figures.

One aspect that is hardly noticed in the overall situation,

as already noted here in issue 5 (2020) on the same subject,

is the security of electricity supply. This is because hardly

anyone is aware of how important a secure power supply

was and is for communication, which is based almost

entirely on electronic channels, in times of a lack of social

“presence”. Neither telephone, whether line or mobile, nor

the many types of internet communication can do without

electricity. In countries with well-developed infrastructure

and largely existing precautionary measures for dealing

with large-scale restrictions caused by infectious diseases,

the continued stable supply of electricity was an expression

of the responsible and forward-looking action of companies

and their employees.

But even if Corona still has us in its grip now, the

expected time after that and the prospects for energy

supply and the contribution of nuclear energy should

perhaps be considered all the more intensively than before.

Let’s take a look at the current situation of nuclear

­power, worldwide: As of August 2020, 440 nuclear power

plant plants are in operation worldwide. With a gross

capacity of around 400,000 MW, they contribute to around

11 % of global electricity production and thus also to

around one third of low-emission technologies. In terms of

secure supply, i.e. uninfluenced by external, natural

conditions and technically plannable, nuclear power is the

technological leader thanks to its high flexibility, both in

terms of the load gradient, i.e. the speed of ramp-up and

ramp-down, and in terms of the scope of available

capacities, i.e. the available megawatts. This is also

demonstrated today by many plants that support the

integration of renewables in grids with high shares of this

volatile generation. But other results from the 2019

operating year are also impressive. For example, the

97 ­nuclear power plants currently in operation in the

United States reached a further peak value with a

­availability of 93.4 percent (based on gross rated output).

The Taishan 1 EPR reactor with 1,750 MW gross capacity

in its first full year of operation not only led the way in

2019 with around 12.99 billion kilowatt hours of ­electricity

generated, but also slightly exceeded the previous

annual top result of Chooz B1, 1,560 MW, France, with

12.97 ­billion kWh from 2012. All in all: nuclear energy still

has something to offer worldwide.

However, action is needed when it comes to medium

and long-term prospects. The nuclear power plant fleet is

also aging. Although the plants do not reach their ­primarily

technical and economic design limits as quickly as other

technologies, and runtimes of 80 or 100 years are not only

under discussion but a reality internationally, the share of

nuclear energy will decline without action as the world’s

hunger for energy continues to rise. The 54 nuclear power

plant plants currently under construction will bring about

75,000 MW of nuclear power plant capacity to the grid in

the coming years due to their capacity size. However,

governments and companies should also quickly tackle

the 104 other projects that are currently in an advanced

planning stage.

The socio-economic benefits of nuclear power, i.e.

secure, competitive and subsidy-free jobs, affordable,

reliable electricity for consumers and low-emission power

generation, should be addressed to policy makers. Nuclear

energy is an investment in the future. It can be safely

implemented in operation with responsibility. The radioactive

residues or waste are not a problem. They are a technical

and scientific challenge for the current generation.

And nuclear energy has two further perspectives: One

is the small and medium power reactors, SMR for short,

with a wide range of possible technologies. For many years

in the past, SMRs were part vision, part speculation. Today

they are part of the energy future in a whole range of

concrete projects for pilot plants - from China and Russia to

the UK, the USA and Canada. And nuclei can not only be

split, they can also be reassembled, in nuclear fusion. This

is where the international fusion project ITER in Cadarche

in the south of France recently reached another milestone

in its construction: with notable political participation, the

interior construction of the ring-shaped nuclear fusion

reactor began at the end of July 2020. The facility is

expected to be completed by the middle of this decade,

paving the way for the energy-producing fusion reactor.

Christopher Weßelmann

– Editor in Chief –

379

EDITORIAL

Editorial

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


atw Vol. 65 (2020) | Issue 8/9 ı August/September

380

EDITORIAL

Auswirkungen der Corona/Covid-19-Krise

auf Energieversorgung und Kernenergie

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

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

Unser Leben wird wie kaum eine andere Krise zuvor vom

­Coronavirus beeinflusst, im privaten Bereich ebenso wie in

Wirtschaft, Gesellschaft, sozialem Leben und Politik. Einschränkungen

unserer täglichen Bewegungsfreiheit sind

direkt spürbar, das Wirtschaftsleben ist schwer getroffen

mit Einbrüchen beim Bruttosozial produkt von bis zu 13 %

in einzelnen OECD-Ländern um rund 5 % der gesamten

Weltwirtschaft.

Auch die Energieversorgung ist betroffen. Offensichtlich

zeigt sich dies bei den weltweiten Ölpreisen. Anfang

des Jahres wurden auf dem Spotmarkt für die U.S.-

amerikanische Sorte WTI (West Texas Intermediate) noch

rund 52 US$ pro Barrel notiert. Im Laufe der Coronakrise

brach der Preis auf 13 US$ ein und erreichte sogar negative

­Werte. Aktuell hat sich der Rohölpreis bei rund 40 US$

­stabilisiert. Beim weltweiten Rohölbedarf notiert die

­Internationale Energieagentur IEA beim Vorjahresvergleich

einen Rückgang von 8,1 auf aktuell rund 91,9 Millionen

­Barrel pro Tag, also etwa 9 Prozent. Der Stromverbrauch verzeichnete

teils dramatische Einbrüche, vor allem in Zeiten

des vollständigen Lockdowns und dem damit ver bundenen

massiven Zurückfahren jeglicher Produktion in der Industrie.

In Großbritannien und Indien lag der Ver­brauchsrückgang

im Vorjahresvergleich so bei fast 30 % und insgesamt

lag in den meisten Ländern der OECD der durchschnittliche

Bedarf auf dem niedrigeren Niveau der vergleichbaren

Vorjahres-Sonntage, also ca. 10 % niedriger. Mit allmäh­licher

Lockerung des vollständigen Lockdowns ab Jahresmitte

2020 stieg der Strombedarf wieder an und erreichte im

­August ein Niveau rund 5 % unter den ­Vorjahreswerten.

Ein in der Gesamtsituation kaum wahrgenommener

­Aspekt, wie an dieser Stelle schon in der Ausgabe 5 (2020)

zum gleichen Themenkomplex angemerkt, ist die Versorgungssicherheit

bei der Stromversorgung. Denn kaum ist

breit bewusst geworden, wie wichtig in Zeiten fehlender

sozialer „Präsenzkontakte“ die sichere Stromversorgung für

die quasi vollständig auf elektronischen Wegen basierende

Kommunikation war und ist. Weder Telefon, ob Festnetz

oder Mobil, noch die vielfältigen Formen der Internetkommunikation

kommen ohne Strom aus. In den Ländern

mit gut ausgebauter Infrastruktur und weitgehend vorhandenen

Vorsorgemaßnahmen im Umgang mit großflächigen

Einschränkungen durch Infektionskrankheiten

war die weiterhin stabile Stromversorgung Ausdruck für das

verantwortungsvolle und vorausschauende Handeln der

Unternehmen und ihrer Mitarbeitenden.

Aber auch wenn uns Corona jetzt noch weiterhin im

Griff hat, sollte über die erwartete Zeit danach und die

Perspektiven der Energieversorgung und dem Beitrag

der Kernenergie vielleicht umso intensiver als vorher nachgedacht

werden.

Dazu ein Blick auf die aktuelle Situation der Kernkraft,

weltweit: Aktuell sind mit Stand August 2020 weltweit

440 Kernkraftwerksblöcke in Betrieb. Sie tragen mit rund

400.000 MW Bruttoleistung zu rund 11 % der weltweiten

Stromproduktion bei und damit auch etwa zu einem Drittel

der emissionsarmen Technologien. Bei der gesicherten

­Versorgung, d. h. von äußeren, natürlichen Gegebenheiten

unbeeinflusst und technisch planbar, ist die Kernenergie

durch ihren hohen Grad an Flexibilität, sowohl was den

Lastgradienten, also die Geschwindigkeit des Hoch- und

­Abfahrens, betrifft, als auch beim Umfang der verfügbaren

Editorial

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

Kapazitäten, also den verfügbaren Megawatt, technologischer

Spitzenreiter. Dies zeigen auch heutzutage viele

Anlagen, die die Integration von Erneuerbaren in Netzen mit

hohen Anteilen dieser volatilen Erzeugung unterstützen.

Doch auch weitere Ergebnisse aus dem Betriebsjahr 2019

beeindrucken. So haben die in den USA laufenden 97 Kernkraftwerke

mit einer Arbeitsverfügbarkeit von 93,4 Prozent

(bezogen auf die Brutto-Nennleistung) einen weiteren

Spitzenwert erreicht. Der EPR-Reaktor Taishan 1 mit

1.750 MW Bruttoleistung war im ersten vollen Betriebsjahr

mit rund 12,99 Milliarden erzeugten Kilowattstunden

Strom nicht nur führend in 2019, sondern hat auch das

­bisherige Jahres-Spitzenergebnis von Chooz B1, 1.560 MW,

Frankreich, mit 12,97 Mrd. kWh aus dem Jahr 2012, leicht

übertroffen. In Summe: Kernenergie hat weltweit noch

etwas zu bieten.

Was allerdings die mittel und langfristigen Perspektiven

betrifft, ist Handeln erforderlich. Auch der Kernkraftwerkspark

altert. Zwar kommen die Anlagen nicht so schnell an

ihre vor allem technisch-wirtschaftliche Auslegungsgrenze

wie andere Technologien, und international sind Laufzeiten

von 80 bzw. 100 Jahren nicht nur in der Diskussion sondern

Realität, aber dennoch wird der Kernenergieanteil ohne

Handeln mit dem weiter steigenden Energiehunger in der

Welt sinken. Die 54 aktuell in Bau befindlichen Kernkraftwerksblöcke

werden aufgrund ihrer Leistungsgröße in den

kommenden Jahren rund 75.000 MW Kernkraftwerkskapazität

ans Netz bringen. Die Regierungen und Unternehmen

sollten aber darüber hinaus die 104 weiteren

Projekte, die sich heute in einem weit fortgeschrittenen

­Planungsstadium befinden, zügig angehen.

Mit den sozioökonomischen Vorteilen der Kernenergie,

also sicheren, wettbewerbsfähigen und subventionsfreien

Arbeitsplätzen, preisgünstigem, versorgungssicherem Strom

für den Verbraucher und einer emissionsarmen Energieerzeugung

sollte an die politischen Entscheidungsträger

appelliert werden. Kernenergie ist eine Zukunftsinvestition.

Sie ist mit Verantwortung sicher im Betrieb umsetzbar.

Die radioaktiven Reststoffe oder Abfälle sind kein Problem.

Sie sind eine technisch-wissenschaftlich lösbare Herausforderung

für die aktuelle Generation.

Und Kernenergie hat zwei weitere Perspektiven: Diese

eine sind die Reaktoren kleiner und mittlerer Leistung,

kurz SMR, mit einem breiten Spektrum von möglichen

­Technologien. Über viele Jahre der Vergangenheit waren

SMR teils Vision, teils Spekulation. Heute sind sie in einer

ganzen Reihe von konkreten Projekten für Pilotanlagen –

von China über Russland bis hin zu Großbritannien, die USA

und Kanada – Teil der Energiezukunft. Und Kerne lassen

sich nicht nur spalten, sie lassen sich auch neu zusammenfügen,

bei der Kernfusion. Hier hat das internationale

­Fusionsprojekt ITER im Südfranzösischen Cadarche kürzlich

einen weiteren Meilenstein bei der Errichtung erreicht:

Unter namhafter politischer Teilnahme begann Ende Juli

2020 der Innenausbau des ringförmigen Kernfusionsreaktors.

Mitte dieses Jahrzehnts soll die Anlage fertiggestellt

sein und den Weg zum Energie liefernden Fusionsreaktor

bereiten.

Christopher Weßelmann

– Chefredakteur –


Kommunikation und

Training für Kerntechnik

Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik?

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

3 Atom-, Vertrags- und Exportrecht

Atomrecht – Das Recht der radioaktiven Reststoffe und Abfälle RA Dr. Christian Raetzke 20.10.2020 Berlin

Export kerntechnischer Produkte und Dienstleistungen –

Chanchen und Regularien

Atomrecht – Was Sie wissen müssen

Atomrecht – Ihr Weg durch Genehmigungs- und

Aufsichtsverfahren

RA Kay Höft M.A. (BWL) 04.11.2020 Berlin

RA Dr. Christian Raetzke

Akos Frank LL. M.

11.11.2020 Berlin

RA Dr. Christian Raetzke 20.01.2021 Berlin

3 Kommunikation und Politik

Public Hearing Workshop –

Öffentliche Anhörungen erfolgreich meistern

Dr. Nikolai A. Behr 10.11. - 11.11.2020 Berlin

3 Rückbau und Strahlenschutz

In Kooperation mit dem TÜV SÜD Energietechnik GmbH Baden-Württemberg:

3 Nuclear English

Stilllegung und Rückbau in Recht und Praxis

Das Strahlenschutzrecht und

seine praktische Umsetzung

Dr. Stefan Kirsch

RA Dr. Christian Raetzke

Dr. Maria Poetsch

RA Dr. Christian Raetzke

23.09. - 24.09.2020 Berlin

29.10. - 30.10.2020 Berlin

English for the Nuclear Industry Angela Lloyd 16.03. - 17.03.2021 Berlin

3 Wissenstransfer und Veränderungsmanagement

Erfolgreicher Wissenstransfer in der Kerntechnik –

Methoden und praktische Anwendung

Veränderungsprozesse gestalten –

Herausforderungen meistern, Beteiligte gewinnen

Dr. Tanja-Vera Herking

Dr. Christien Zedler

Dr. Tanja-Vera Herking

Dr. Christien Zedler

05.10. - 06.10.2020 Berlin

24.11. - 25.11.2020 Berlin

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

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

382

CONTENTS

Issue 8/9 | 2020

August/September

Contents

Editorial

Implications of the Corona/Covid-19 Crises

on Energy Supply and Nuclear Power E/G 379

Inside Nuclear with NucNet

South Africa / Policy Announcement Opens Door

to all Types of Nuclear Technology 384

Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . .385

Cover:

Dismantling the concrete shell of the

containment of NPP Philippsburg 1.

Courtesy of EnBW Kernkraft GmbH

Contents:

Manipulator Operated Laser Ablation

(MANOLA).

Courtesy of KIT Institute of Technology

Calendar . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .386

Feature | Decommissioning And Waste Management

Current Status and Prospects of Nuclear Power Plant

Decommissioning in the Republic of Korea 387

Spotlight on Nuclear Law

Cost Correction for Site Selection Procedures –

It Remains Confusing G 392

Decommissioning and Waste Management

Actual Research and Development Activities

in the Field of Dismantling 394

A Geopolymer Waste Form for Technetium, Iodine

and Hazardous Metals 397

Decommissioning of Nuclear Power Plants: Waste Streams

and Release Measurements 400

Ventilation Concepts for Nuclear Decommissioning 403

Steam Generator Rip and Ship – a Valuable Contribution

to Decommissioning and Dismantling of Nuclear Power Plants 406

Casks and Cask Stacks in Interim Storage Facilities

under Earthquake Loads 409

Radioactivity Calculation of the Concrete Shielding

of the Petten LFR and the Dodewaard BWR 414

Quality Assurance and Data Analysis in Automated Radiological

Characterization of Large Soil Volumes 418

Construction of a Dismantling Hall for Large Components

at Entsorgungswerk für Nuklearanlagen in Lubmin G 421

Environment and Safety

Current Procedure for Determining Release Parameters for a Plane

Crash on a Nuclear Facility in the Context of Accident Analyses 426

Research and Innovation

Evaluation of MACST Strategies for Extended Loss

of AC Electric Power Event in OPR1000 Nuclear Power Plants 430

Neutronic Study of CAREM-25 Advanced Small Modular Reactor

Using Monte Carlo Simulation 435

KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .440

Report

Operating results 2019 441

News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .456

Nuclear Today

Europe Can’t Discard Nuclear Investment

in Quest for a Clean Energy Future 462

G

E/G

= German

= English/German

Imprint 393

Insert: ICOND 2020

Contents


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Feature

Feature | Decommissioning

And Waste Management

383

CONTENTS

387 Current Status and Prospects

of Nuclear Power Plant Decommissioning

in the Republic of Korea

Joo Hyun Moon

Decommissioning and Waste Management

394 Actual Research and Development Activities

in the Field of Dismantling

Sascha Gentes and Nadine Gabor

397 A Geopolymer Waste Form for Technetium, Iodine

and Hazardous Metals

Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg

400 Decommissioning of Nuclear Power Plants:

Waste Streams and Release Measurements

Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann

Research and Innovation

435 Neutronic Study of CAREM-25 Advanced Small Modular Reactor

Using Monte Carlo Simulation

Saeed Zare Ganjaroodi and Ali Pazirandeh

Report

441 Operating results 2019

Contents


atw Vol. 65 (2020) | Issue 8/9 ı August/September

384

South Africa / Policy Announcement Opens Door to

all Types of Nuclear Technology

Country wants 2,500 MW of new reactors, but financing remains an obstacle

INSIDE NUCLEAR WITH NUCNET

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

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

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

The plan is to allow vendors to self-finance 100 % of the cost

which means the national government will not provide any

funding.

Mineral resources and energy minister Gwede Mantashe

said his agency will issue a request for information to assess the

market with a focus on SMRs. However, he said all options are

being explored and if the market indicates one design is more

affordable and can be built more efficiently, he will go with it.

He did not say when his agency expects a vendor to break

ground, nor did he specify a preference for any particular

reactor designs.

He said: “We may give a company a right to develop a

nuclear station on a build, operate, and transfer basis. It

means there is no immediate funding from the state.”

The nuclear plans are a significant development on what

has been a long and sometimes controversial effort by South

Africa to add to its existing commercial nuclear capacity.

In 2018 plans to expand nuclear capacity by building up to

9,600 MW of new plants were put on hold with nuclear

excluded from an integrated resource plan (IRP) because the

government saw electricity generation from other sources as

cheaper and because there was a lower demand for electricity

than forecast in an earlier plan in 2010.

The IRP called for nuclear capacity to remain at 1,860 MW

(net) by 2030, which means there will be no change.South

­Africa’s only nuclear station at Koeberg has two pressurised

water reactor units that have been in commercial operation

since 1984 and 1985. Their output accounts for 2.5 % of the

country’s energy generation.

The plan was drawn up a year after the Western Cape High

Court said a series of preliminary procurement deals for new

nuclear between the government of South Africa and Russia,

China, the US, South Korea and France were illegal.

The court ruled in April 2017 that the procurement ­process

was unconstitutional and illegal as it was not ­sufficiently ­public

and did not involve adequate environmental and financial

­assessments. After theverdict, then energy minister Mmamoloko

Kubayi said the government would go ahead with the signing of

five new intergovernmental agreements with potential international

partners, but this never happened and ambitious plans

for up to 9,600 MW of new nuclear ­capacity were dropped.

That plan, put forward by then president Jacob Zuma,

would have seen South Africa sign a deal with Russia’s state

­nuclear corporation Rosatom for eight 1,200 MW VVER ­nuclear

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

terms were that it would provide 50 % of the financing.

The plan died for three reasons. The first was that South

Africa couldn’t afford it, even with generous financial terms

from any vendor. The second was that Mr Zuma’s administration

was rife with allegations of corruption and nepotism. The

third was the lack of transparency related to how the procurement

process for the deal was done. It was said to have come

about as a result of a “secret” meeting between Mr Zuma and

Russian president Vladimir Putin in a side meeting at a development

conference in Brazil. No tender had been released for

the project before that meeting.

Separately, the nation’s economy has been hobbled by a

series of electricity blackouts due to a lack of electrical power

and an aging grid infrastructure.

Eskom, the state-owned utility that also operates Koeberg’s

two reactors, has been thwarted in its requests to raise rates as

the government uses cheap electricity as a way to address

poverty in the country. The government has also declined to

subsidise Eskom directly.

A proposed turnaround plan for Eskom has been put on hold

due to the coronavirus pandemic. Eskom’s turnaround plan

includes proposed debt transfer to the government, cost containment,

operational reforms and the company’s unbundling

into three separate entities: generation, transmission and

distribution.

The latest announcement from Mr Mantashe immediately

ran into significant challenges. Shadow minister of mineral

resources and energy Kevin Mileham questioned whether the

100 % vendor financed approach would work and said the timeline

for issuing and evaluating tenders might not be feasible.

Additionally, he pointed to the IRP, which he said makes no

mention of nuclear energy at least for the next decade. In fact,

the IRP makes brief mention of “preparations for nuclear

­energy”, but does not mention a specific level of generating

capacity or a timeline for a procurement.

Meanwhile, the need for new baseload generation remains.

Africa’s inability to generate enough electricity continues to

hamper economic growth, cutting 2 % to 4 % off GDP every

year, according to the Africa Progress Panel.

Nuclear could be part of the solution, but financing ­remains

an issue. Nuclear plants are relativelt cheap to operate, but are

expensive to build and require significant upfront capital with a

long wait for any return on investment.

The Nuclear Industry Association of South Africa (­Niasa)

has proposed six possible funding options for new nuclear, but

government officials have suggested the most likely is a “build,

own, operate and transfer” (Boot) model similar to that used

by Russia for project including Akkuyu in Turkey.

Niasa had earlier welcomed Mr Mantashe’s announcement

of plans to produce a roadmap for new nuclear power plants,

saying it gives the requisite policy certainty which “enables

industry to respond accordingly”.

The government remains cautious. Mr Mantashe said it

will “test the market” to hear what potential investors and

consortia have to say about building a new nuclear facility.

Options being considered include giving a “right to develop

a modular nuclear station on a build, own, operate and transfer

basis,” which means there may be no imme diate call for

state funding” he said.

“We are going to explore all options, when there is appetite

for nuclear in the market we will go ahead with it,”

Mr ­Mantashe added.

An International Energy Agency report said electricity

generation from nuclear energy is likely to increase only

slightly in South Africa by 2040, but could see a bigger jump if

policies are enacted to develop the continent’s energy sector.

Those policies would include faster economic expansion

accompanied by the full achievement of key sustainable development

goals by 2030. Those goals, including full access to

electricity, would allow economies to grow “strongly, sustainably

and inclusively”, the IEA said.

Author

NucNet – The Independent Global Nuclear News Agency

Editors responsible for this story:

David Dalton and Dan Yurman

Avenue des Arts 56 2/C

1000 Bruxelles

www.nucnet.org

www.neutronbytes.com

Inside Nuclear with NucNet

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


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Did you know...?

Enabling deep decarbonisation:

The main drivers of nuclear new build cost and ways to reduce them

In July 2020 the Nuclear Energy Agency (NEA) of the OECD released the

report “Unlocking Reductions in the Construction Costs of Nuclear: A Practical

Guide for Stakeholders” assessing opportunities to reduce construction costs

at the plant-level. The report highlights the role of the different stakeholders

– particularly policymakers. It aims to unlock a positive learning trend in

nuclear construction to enable the full potential of nuclear power as integral

part of future low-carbon energy generation and decarbonisation pathways.

The report con textualises the nuclear challenge with the necessity to reduce

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

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

Development Scenario of the International Energy Agency that requires the

commissioning of 15 GW new nuclear capacity globally per year. The main

conclusions and advice derived from the NEA case and study analysis of

which some key parameters are presented in the graphs below are the

following:

p Capitalise on lessons learnt from recent Gen III construction projects. A

window of opportunity exists for cost reductions with recently realized Gen III

designs where the nuclear industry and its supply chain have in large part

redeveloped their capabilities in several OECD countries. This requires timely

new-build decisions in the early 2020s.

p Prioritise design maturity and regulatory stability. Policy support mechanisms

should include requirements for design maturity and construction readiness.

The regulatory framework needs to remain stable and predictable throughout

construction.

p Consider committing to a standardised nuclear programme. A standardised

nuclear programme is the most promising way to achieve cost reductions by

the series effect, multi-unit con struction and design and process optimisation.

p Enable and sustain supply chain development and industrial performance.

New-build ambitions have to consider supply chain constraints and ensure

continuous activity to enable and sustain development in the framework of a

long-term energy policy commitment.

p Foster innovation, talent development and collaboration at all levels. Cost

reduction can arise from innovative nuclear technologies (i.e. SMRs and

Gen-IV reactors). Governments can support this by the timely development of

demonstration projects and an adequate licensing framework.

p Support robust and predictable market and financing frameworks. Nuclear

projects require long-term government commitments and market regulations.

In addition, financial support is currently essential in western OECD

countries to address market failure and deliver cost-competitive new nuclear

construction.

p Encourage concerted stakeholder efforts. Governments should create an

environment that fosters a social contract with industry and society to reduce

nuclear construction costs as shown by the Nuclear Sector Deal in the United

Kingdom.

p Tailor government involvement to programme needs. Government financial

support for fleet programmes will likely decrease gradually with industry

maturity and perceived risk levels falling, but restarting a nuclear programme

likely will require further government support.

DID YOU EDITORIAL KNOW...?

385

Cost breakdown for nuclear power

levelised cost of electricity (LCOE)

13 %

11 %

9 %

20 %

47 %

IDC and return of investment are

financing cost, i. e. 67 percent of total cost

p Interest during construction (IDC)

p Return of capital

p Overnight construction costs (OCC)

p Operations and maintenance (O&M)

p Fuel

Key drivers of Flamanville 3 EPR cost overruns

8,000

6,000

4,000

2,000

0

1,577

1,923

1,771

Flamanville ex-post

607

p Budgeted cost

p Design maturity

p Project management

(FOAK effect)

p Regulatory changes

p Delays

1,697 2,900

Taishan ex-post, pro-forma adjusted

Source: “Unlocking Reductions in the Construction Costs of Nuclear”, NEA, 2020

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

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

Source: NEA (2020), based on Folz (2019),

“Rapport au Président Directeur Général d’EDF:

La construction de l’EPR de Flamanville.”

LCOE of a new nuclear power plant project according to the cost of capital

160

p Fuel cycle costs p Operation & maintenance costs p Overnight construction costs p Cost of capital

120

80

40

0

0 % 1 % 2 % 3 % 4 % 5 % 6 % 7 % 8 % 9 % 10 % 11 % 12 %

For further details

please contact:

Nicolas Wendler

KernD

Robert-Koch-Platz 4

10115 Berlin

Germany

E-mail: presse@

KernD.de

www.KernD.de

Did you know...?


atw Vol. 65 (2020) | Issue 8/9 ı August/September

386

Calendar

This is not a full list. Dates are subject to change.

Please check the listed websites for updates.

2020

CALENDAR

07.10. – 08.10.2020

9 th ENPPS - EU Nuclear Power Plant Simulation

Forum. Brussels, Belgium, NRG Events,

www.nrg-events.com

11.10. – 15.10.2020

RRFM – European Research Reactor Conference.

Helsinki, Finland, European Nuclear Society,

www.euronuclear.org

Virtual Meetings 15.10. and 21.10.2020

Implementing Digital Innovation in a Nuclear

Environment 2020. Nuclear Institute,

www.nuclearinst.com

19.10. – 23.10.2020

International Conference on the Management

of Naturally Occurring Radioactive Materials

(NORM) in Industry. Vienna, Austria, IAEA,

www.iaea.org

Cancelled

ATH'2020International Topical Meeting

on Advances in Thermal Hydraulics.

Paris, France, SFEN, www.sfen-ath2020.org

26.10. – 30.10.2020

NuMat 2020 – 6 th Nuclear Materials Conference.

Gent, Belgium, IAEA, www.iaea.org

Virtual Meeting 29.10. – 30.10.2020

Nuclear Decommissioning & Used Fuel Strategy.

Reuters Events, www.nuclearenergyinsider.com

Virtual Meeting 04.11. – 06.11.2020

The Power & Electricity World Africa 2020.

Terrapinn, www.terrapinn.com

08.11. – 12.11.2020

Advancing Geological Repositories

from Concept to Operation. Helsinki, Finland,

OECD, Nuclear Energy Agency, www.oecd-nea.org

09.11. – 13.11.2020

International Conference on Radiation Safety:

Improving Radiation Protection in Practice.

Vienna, Austria, IAEA, www.iaea.org

Virtual Meeting 15.11. – 19.11.2020

ANS Winter Meeting and Technology of Fusion

Energy (TOFE 2020). American Nuclear Society,

www.ans.org

30.11. – 02.12.2020

European Power Strategy & Systems Summit.

Prague, Czech Republic, European Power

Generation, www.europeanpowergeneration.eu

07.12. – 10.12.2020

SAMMI 2020 – Specialist Workshop on Advanced

Measurement Method and Instrumentation

for enhancing Severe Accident Management in

an NPP addressing Emergency, Stabilization and

Long-term Recovery Phases. Fukushima, Japan,

NEA, www.sammi-2020.org

08.12. – 10.12.2020

World Nuclear Exhibition 2020. Paris Nord

Villepinte, France, Gifen,

www.world-nuclear-exhibition.com

Virtual Meeting 17.12. – 18.12.2020

ICNESPP 2020 – 14. International Conference on

Nuclear Engineering Systems and Power Plants.

WASET, www.waset.org

2021

07.03. – 11.03.2021

WM2021 – Waste Management Symposia.

Phoenix, Arizona, USA, X-CD Technologies,

www.wmsym.org

17.03. – 19.03.2021

KONTEC 2021 – 15 th International Symposium

“Conditioning of Radioactive Operational &

Decommissioning Wastes”. Dresden, Germany,

atm, www.kontec-symposium.de

23.03. – 26.03.2021

7 th international conference on Education and

Training in Radiation Protection. Groningen,

Netherlands, FuseNet, www.etrap.net

26.04. – 27.04.2021

AtomExpo 2021. Sochi, Russia, Rosatom,

www.2021.atomexpo.ru

26.04. – 30.04.2021

European Nuclear Young Generation Forum

(ENYGF). Tarragona, Spain, ENYGF, www.enygf.org

Postponed to 02.06. – 04.06.2021

HTR2020 – 10 th International Conference

on High Temperature Reactor Technology.

Yogyakarta, Indonesia, Indonesian Nuclear Society,

www.htr2020.org

Postponed to 30.08. – 03.09.2021

International Conference on Operational Safety

of Nuclear Power Plants. Beijing, China, IAEA,

www.iaea.org

Postponed to 08.09. – 10.09.2021

3 rd International Conference on Concrete

Sustainability. Prague, Czech Republic, fib,

www.fibiccs.org

27.09. – 01.10.2021

NPC 2021 International Conference on Nuclear

Plant Chemistry. Antibes, France, SFEN Société

Française d’Energie Nucléaire,

www.sfen-npc2021.org

Postponed to 07.09. – 10.09.2021

International Forum on Enhancing a Sustainable

Nuclear Supply Chain. Helsinki, Finland, Foratom,

https://events.foratom.org/mstf2020/

Postponed to 30.11. – 02.12.2021

Enlit (former European Utility Week and

POWERGEN Europe). Milano, Italy,

www.powergeneurope.com

Postponed to 2021

The Frédéric Joliot/Otto Hahn Summer School

on Nuclear Reactors “Physics, Fuels and Systems”.

Aix-en-Provence, France, CEA & KIT, www.fjohss.eu

Postponed to 2021

INDEX 2020: International Nuclear Digital

Experience. Paris, France, SFEN,

www.sfen-index2020.org

Postponed to 2021

4 th CORDEL Regional Workshop – Harmonization

to support the operation and new build of NPPs

including SMR. Lyon, France, World Nuclear

Association, www.events.foratom.org

2022

18.11. – 19.11.2020

INSC — International Nuclear Supply Chain

Symposium. Munich, Germany, TÜV SÜD,

www.tuvsud.com

23.11. – 25.11.2020

KELI 2020 – Conference for Electrical Engineering,

I&C and IT in generation plants. Bremen, Germany,

VGB PowerTech e.V., www.vgb.org

Postponed to 10.05. – 15.05.2021

FEC 2020 – 28 th IAEA Fusion Energy Conference.

Nice, France, IAEA, www.iaea.org

Postponed to 29.05. – 05.06.2021

BEPU2020 – Best Estimate Plus Uncertainty International

Conference, Giardini Naxos. Sicily, Italy,

NINE, www.nineeng.com

24.11. – 26.11.2020

ICOND 2020 – 9 th International Conference on

Nuclear Decommissioning. Aachen, Germany,

AiNT, www.icond.de

Postponed to 31.05. – 04.06.2021

20 th WCNDT – World Conference on

Non-Destructive Testing. Incheon, Korea,

The Korean Society of Nondestructive Testing,

www.wcndt2020.com

KERNTECHNIK 2022.

Germany, KernD and KTG,

www.kerntechnik.com

Calendar


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Current Status and Prospects of

Nuclear Power Plant Decommissioning

in the Republic of Korea

Joo Hyun Moon

Introduction As of August 2020, there are 30 nuclear power reactors in the Republic of Korea as shown in

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

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

permanently shut down.

South Korea is now facing the problem of safe decommissioning

of Kori unit 1 because it is the first commercial

power reactor to be decommissioned. Wolsong unit 1 was

declared to be permanently shut down in June 2018, and

an application for a permit of change in its operating

license, which is an initial step to decommission a nuclear

facility required by nuclear safety regulations in Korea,

was approved by the nuclear safety and security commission

(NSSC) that is the national nuclear regulatory

­authority in December 2019.

No. Name Reactor

Type

Net Capacity

(MWe)

Although it has experience on completing the decommissioning

of two research reactors (TRIGA MARK II and

III), Korea does not have any experience on the whole

decommissioning process of a commercial nuclear power

reactor. The Korean government has been preparing for

the safe decommissioning of Kori unit 1 since several years

prior to the expiration date of its renewed operating

license. The government has considered decommissioning

Kori unit 1 to be an opportunity to develop new technologies

to decommission nuclear facilities and extend its

Issue Date of Operating License

(First Critical Date)

1 Kori 1 PWR 587 1972.05.31

(1977.06.19)

Expiration Date

of Operating License

Design Life

(Year)

Status

2017.06.18 30 Permanent

shutdown

2 Kori 2 PWR 650 1983.08.10 2023.08.09 40 Operational

3 Kori 3 PWR 950 1984.09.29 2024.09.28 40 Operational

4 Kori 4 PWR 950 1985.08.07 2025.08.06 40 Operational

5 Wolsong 1 PHWR 679 1978.02.15

(1982.11.21)

2022.11.20 30 Permanent

shutdown

6 Wolsong 2 PHWR 700 1996.11.02 2026.11.01 30 Operational

7 Wolsong 3 PHWR 700 1997.12.30 2027.12.29 30 Operational

8 Wolsong 4 PHWR 700 1999.02.08 2029.02.07 30 Operational

9 Hanbit 1 PWR 950 1985.12.23 2025.12.22 40 Operational

10 Hanbit 2 PWR 950 1986.09.12 2026.09.11 40 Operational

11 Hanbit 3 PWR 1,000 1994.09.09 2034.09.08 40 Operational

12 Hanbit 4 PWR 1,000 1995.06.02 2035.06.01 40 Operational

13 Hanbit 5 PWR 1,000 2001.10.24 2041.10.23 40 Operational

14 Hanbit 6 PWR 1,000 2002.07.31 2042.07.30 40 Operational

15 Hanul 1 PWR 1,000 1987.12.23 2027.12.22 40 Operational

16 Hanul 2 PWR 1,000 1988.12.29 2028.12.28 40 Operational

17 Hanul 3 PWR 1,000 1997.11.08 2037.11.07 40 Operational

18 Hanul 4 PWR 1,000 1998.10.29 2038.10.28 40 Operational

19 Hanul 5 PWR 1,000 2003.10.20 2043.10.19 40 Operational

20 Hanul 6 PWR 1,000 2004.11.12 2044.11.11 40 Operational

21 Shin Kori 1 PWR 1,000 2010.05.19 2050.05.18 40 Operational

22 Shin Kori 2 PWR 1,000 2011.12.02 2051.12.01 40 Operational

23 Shin Kori 3 PWR 1,400 2015.10.30 2075.10.29 60 Operational

24 Shin Kori 4 PWR 1,400 2019.02.01 2079.01.31 60 Operational

25 Shin Wolsong 1 PWR 1,000 2011.12.02 2051.12.01 40 Operational

26 Shin Wolsong 2 PWR 1,000 2014.11.14 2054.11.13 40 Operational

27 Shin Hanul 1 PWR 1,400 - - 60 Under

construction

28 Shin Hanul 2 PWR 1,400 - - 60 Under

construction

29 Shin Kori 5 PWR 1,400 - - 60 Under

construction

30 Shin Kori 6 PWR 1,400 - - 60 Under

construction

387

FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT

| Tab. 1.

Nuclear Reactors in the Republic of Korea. (As of August 2020)

Feature

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 388

| Fig. 1.

Nuclear power plant sites in the Republic of Korea.

own nuclear industrial capabilities to cover all areas of the

nuclear fuel cycle.

In April 2019, the Korean government established ‘a

business strategy to promote domestic industry in the field

of nuclear facility decommissioning’ [1] to technically

support the decommissioning of Kori unit 1 and to develop

the domestic nuclear industry’s technical competence in

regards to the decommissioning of nuclear facilities. With

this strategy, the Korean government set an ambitious

target to foster the domestic nuclear industry’s technical

competence; the goal is to be a contender in the global

market for nuclear facility decommissioning by 2035.

The Korea Hydro and Nuclear Power Corporation

( KHNP) operates all nuclear power plants in the Republic

of Korea and is responsible for the safe and successful

decommissioning of Kori unit 1. KHNP came up with and is

now collecting opinions from the local residents on the

decommissioning plan (DP) for Kori unit 1 as required by

our national nuclear safety regulations. Main decommissioning

activities will start immediately after the DP is

approved by the Nuclear Safety and Security Commission

(NSSC). For successful decommissioning of the first

commercial nuclear power reactor in Korea in cooperation

with domestic and foreign corporations, the KHNP has

prepared extensively from the development of technologies

needed for decommissioning and securing funds for

decommissioning to organizing a department responsible

for leading the decommissioning project. These developments

have been in motion since several years before the

expiration date of Kori unit 1’s renewed operating license.

Few papers have been written on the preparation status

of the decommissioning of Kori unit 1 and prospects

regarding the decommissioning of nuclear power plants in

the Republic of Korea. Hence, this paper will examine the

preparation status of the decommissioning of Kori unit 1

by the KHNP and the Korean government, and the

prospects of decommissioning of other nuclear power

reactors in the Republic of Korea.

Status of Kori unit 1

Kori unit 1 is the first commercial nuclear power reactor in

Korea and operated for about 40 years starting June 1977

when it reached first criticality. Kori unit 1 had a capacity

of 587 MW and produced a total of 155,260 GWh of electricity

in that period. The original operating lifetime for

Kori unit 1 was 30 years, but was extended by 10 years

through license renewal in 2007. In June 2015, the board

of directors of KHNP decided to permanently shut

down Kori unit 1 with a renewed operating license

­termination date of 18 June 2017.

For its permanent shutdown, pursuant with nuclear

safety regulations, the KHNP had to submit a written

application for a permit of change in operating license on

24 June 2016 to the NSSC, and won the approval from

­NSSC on 9 June 2017. The KHNP then shut down Kori unit

1 permanently on 18 June 2017. Then, spent nuclear fuel

was discharged from the reactor and transported to the

temporary storage water pool. Since then, Kori unit 1 had

been kept in the cold shutdown state.

Pursuant to an immediate dismantling strategy, KHNP

decided to complete the decommissioning of Kori unit 1

with a budget of 812.9 billion Korean Won (KWN) as the

2019-year present value (about 677 million US dollars at

exchange rate of $1 = 1,200 KWN) within about 15 years.

The strategy consists of four main stages: 1) management

of permanent shutdown conditions and preparation for

decommissioning; 2) beginning of decommissioning and

construction of facilities for radioactive waste treatment;

3) decontamination/dismantling and waste treatment;

and 4) site restoration and report of completion of

decommissioning, as shown in Figure 2 [2]. The main

tasks in each stage are as follows:

Stage 1

p Management of spent nuclear fuel and inspections

­conducted by a regulatory body;

| Fig. 2.

Four main stages of decommissioning of Kori unit 1 [2].

Feature

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


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 3.

Development status of technologies and devices for decommissioning of Kori unit 1 [2].

p Development of DP and the design and procurement of

waste treatment facilities;

p Preparation of decommissioning planning documents

and application for DP approval from the NSSC.

Stage 2

p Dismantling of non-radioactive areas, installation and

operation of utilities for decommissioning;

p Construction of radioactive waste treatment facilities;

p Transport of spent nuclear fuel to off-site storage

facility.

Stage 3

p Decontamination and dismantling of radioactive

­systems and structures;

p Operation of radioactive waste treatment facilities

( decontamination, cutting, volume reduction, packaging

etc.);

p Evaluation and verification of radioactivity measurements.

Stage 4

p Site restoration;

p Final status survey and inspection for closure of decommissioning

processes;

p Termination of operating license for Kori unit 1.

To secure technical competence in the decommissioning of

Kori unit 1, the KHNP formulated and implemented the

roadmap for the development of 17 decommissioning

technologies in 2017, as shown in Figure 3. The KHNP

developed the technology tree to decommission nuclear

power plants and identified 58 technologies to complete

decommissioning. The gap analysis identified 17 technologies

that were lacking or insufficient for use onsite and

must be developed with urgency. As of December 2017,

13 of 17 technologies were under development and the

development of the other four technologies will be

completed by 2021. In parallel with technologies development,

the key decommissioning devices with high value

added are being developed.

Although the DP for Kori unit 1 is subject to change

because it has not been approved by the NSSC, the KHNP is

going to carry forward with the decommissioning project

using the project management system shown in Figure 4.

In this system, KHNP is in overall charge of this project and

will be a licensee for the decommissioning of Kori unit 1.

The maintenance of Kori unit 1 till initiation of the main

decommissioning activities will be undertaken by the

KHNP in cooperation with the existing contractors.

For efficient project management, the KHNP would

categorize the decommissioning projects into the six

elementary businesses: comprehensive design of decommissioning;

system decontamination, construction and

operation of waste treatment facilities and decontamination/dismantling;

cutting and dismantling of the nuclear

reactor system; radiation measurement, evaluation, and

verification; and site restoration. For each elementary

business, the KHNP will choose and enter into a business

contract with a well-equipped company or consortium

through competitive bids.

There are three matters that might have significant

­influence on the schedule of the Kori unit 1 decommissioning

project. The first is transport of spent nuclear fuel out

of the temporal storage pool. Pursuant to nuclear safety

regulations, the main decommissioning activities shall not

FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 389

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FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 390

| Fig. 4.

Project management system for decommissioning of Kori unit 1 [2].

| Fig. 5.

Number of nuclear reactors to be decommissioned by time period.

be started until all spent nuclear fuel stored at the temporal

storage pool of Kori unit 1 has been transported to other

storage pools or sites. At present, it seems to be hard to

secure enough space at the other sites to accommodate all

nuclear fuel accumulated at the temporal storage pool of

Kori unit 1. It is also uncertain when the interim storage

site for spent nuclear fuel could be secured due to an

aversion to spent nuclear fuel in Korea. Due to this fact, the

decommissioning of Kori unit 1 may be delayed by several

years.

The second is disposal of radioactive waste. KHNP has

set 14,500 200L drums as the target for final disposal

amount of all low- and intermediate-level radioactive

waste generated over the whole decommissioning period.

Before applying any volume reduction techniques, about

several hundred thousand drums of decommissioning

waste including intermediate-, low-, and very low-level

waste would be generated during the whole process

of decommissioning, which is huge amount even for a

relatively short period of about 15 years. It is big challenge

to classify, condition, and treat huge amounts of waste

­adequately and finally reduce the waste volume to about a

tenth of raw waste volume. Even if successful in reducing

the volume, 14,500 drums would be also big burden to the

operator of the disposal facility because all drums need to

be inspected to comply with waste acceptance criteria.

Because unit disposal costs per 200L drum of low- and

intermediate-level waste is about 15.2 million KWN (about

12,667 US $ at exchange rate of $1 = 1,200 KWN) in

Korea, the more drums to be disposed of, the higher the

disposal cost and the higher the total cost of decommissioning.

To reduce disposal burden and cut down total

costs of decommissioning Kori unit 1 as much as possible,

it is essential to minimize the quantity of waste drums to be

finally disposed.

The third matter is the related impacts on the operation

of neighboring nuclear reactors. At the Kori site, there are

a total of four nuclear reactors including Kori unit 1.

Because Kori unit 2 is immediately adjacent to Kori unit 1,

the decommissioning works of Kori unit 1 might influence

the operation of Kori unit 2. Hence, the decommissioning

works should carry forward only within the limits so as to

not cause inconvenience in the operation of Kori unit 2; the

whole schedule of Kori unit 1 could be affected by the

operation schedule of Kori unit 2.

Prospect of nuclear decommissioning in Korea

In December 2017, the Ministry of Trade, Industry and

­Energy (MOTIE) released ‘8 th Basic Plan for Long-term

Electricity Supply and Demand (2017–2031) (8 th plan)’

[3]. According to the 8 th plan, license renewal beyond the

designated lifetimes of any existing nuclear reactor is not

allowed. Without considering lifetime extension, the

number of nuclear power reactors to be permanently shut

down with each coming decade is shown in Figure 5.

The annual expenses expected to be spent on

the decommissioning of the 26 reactors in Korea are

roughly estimated without considering inflation. Among

30 reactors in Table 1, the decommissioning start dates of

four reactors which are being under construction were

not able to be fixed and are therefore not included in this

estimation. For this estimation, the main assumptions are

as follows:

Considering the 8 th plan, the license renewal for the

designated lifetime extension for all nuclear power plants

is not considered. The decommissioning of a nuclear

Feature

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


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 6.

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

power reactor will be completed within 15 years. Stage 1

for decommissioning would be initiated immediately after

the permanent shutdown date.

The total expense for decommissioning of a nuclear

­reactor will be 812.9 billion KRW which is equivalent to

about $677M US at the 2019-year present value, which is

subject to review and revision every two years by MOTIE.

The decommissioning cost consists of labor costs,

dismantling costs, decontamination costs, waste disposal

costs (including waste transportation costs), and miscellaneous

costs such as insurance fees, taxes, and utility costs.

The annual expense varies according to activities that

would be performed by year.

Based on the above assumptions, the annual decommissioning

cost of the 26 nuclear power reactors were shown

in Figure 6. Accordingly, the annual decommissioning cost

and the decommissioning market size will grow gradually

and reach the maximum size in 2037 when the decommissioning

of 13 nuclear reactors will be carried out

simultaneously. From that point, the decommissioning

market size will gradually decrease until 2069 when the

decommissioning of Shin Wolsong unit 2 is completed.

There would be no decommissioning works over the next

five years from 2070 to 2074, and in 2075, the decommissioning

of Shin Kori unit 3 would be initiated. Those

estimated timings might be altered due to changes in the

future circumstances.

Conclusion

This paper reviews the current status of the decommissioning

process of Kori unit 1. Because Kori unit 1 is the

first commercial nuclear reactor to be decommissioned in

the Republic of Korea, the Korean government and the

KHNP have prepared to ensure the safety of the project.

In particular, the Korean government considered this to

be an opportunity to strengthen its nuclear industry’s

competence and foster domestic specialized companies in

the field of decommissioning of nuclear facilities. Thus,

the Korean government formulated ‘a business strategy to

promote the domestic industry in field of nuclear facility

decommissioning.’ The KHNP has made preparations to

decommission Kori unit 1 through measure such as the

formulation and implementation of roadmaps for technological

developments, formulation of a decommissioning

plan and making license-related documents, and etc. This

paper estimates the number of nuclear power reactors to

be permanently shut down with each coming decade and

the expected annual expenses of decommissioning 26

­nuclear power reactors in Korea from 2017 to 2094.

­Although the estimations presented are rough, they show

that the size of the decommissioning market will grow

gradually and reach the maximum in 2037 when the

decommissioning of 13 nuclear reactors would be carried

out simultaneously. This paper is expected to be helpful to

get insight on the status and prospects of decommissioning

of nuclear facilities in the Republic of Korea.

Acknowledgements

This present research was conducted by the research fund

of Dankook University in 2019.

References

[1] Korean Government, “a business strategy to promote the domestic industry in field of nuclear facility

decommissioning (in Korean),” April 2019.

[2] Young Gi Choi, “Plan for Decommissioning of Kori Unit 1 (in Korean),” Nuclear Industry, 2017(12),

pp.40-49, December 2017.

[3] Ministry of Trade, Industry and Energy, “8th Basic Plan for Long-term Electricity Supply and Demand,”

December 2017.

Author

Joo Hyun Moon

jhmoon86@dankook.ac.kr

Department of Nuclear Engineering

Dankook University

Cheonan-Si,

Chungnam 31116, Rep. of Korea

FEATURE | DECOMMISSIONING AND WASTE MANAGEMENT 391

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

392

Kostenkorrekturen für Standortauswahlverfahren –

Es bleibt unübersichtlich

Tobias Leidinger

SPOTLIGHT ON NUCLEAR LAW

Gesetzliche Kostenregelungen sind häufig kompliziert, unübersichtlich und zuweilen auch nicht (mehr)

sachgerecht. Mit dem Ende Juli 2020 vom Bundeskabinett beschlossenen „Gesetz zur Anpassung der Kostenvor­schriften

im Bereich der Entsorgung radioaktiver Abfälle sowie zur Änderung weiterer Vorschriften“ zielt der Gesetzgeber u. a.

auf eine nachträgliche Korrekturmöglichkeit bei der Kostenverteilung für das Standortauswahlverfahren nach

­Standortauswahlgesetz (StandAG) ab. Dieser grundsätzlich begrüßenswerte Schritt fällt bei näherer Betrachtung

allerdings nicht konsequent aus.

Kostentragung für die Endlagersuche

Die Kosten für das komplexe Standortauswahlverfahren

in Deutschland für das Endlager für hochradioaktive

­Abfälle werden nach § 29 Abs. 1 StandAG auf die Umlagepflichtigen

anteilig umgelegt. Das bedeutet konkret,

dass das Bundesumweltministerium (BMU) jedes Jahr

Bescheide zur Festsetzung der im vorangegangenen Jahr

umlagefähigen Kosten einerseits und zur Vorauszahlung

dieser Kosten für das laufende Jahr andererseits gegenüber

den Beitragsverpflichteten erhebt.

Das Problem: Überholter Kostenschlüssel

Problematisch ist diese Kostenregelung deshalb, weil der

Maßstab, nach dem sich der jeweils zu entrichtende Anteil

eines Umlagepflichtigen an den umlagefähigen Kosten bemisst,

bereits seit längerer Zeit nicht mehr „passt“. § 29

Abs. 2 StandAG legt insoweit fest, dass dafür der Aufwandschlüssel

aus § 6 der Endlagervorausleistungsverordnung

(EndlagerVlV) entsprechend heranzuziehen ist. Dieser

„Schlüssel“ ist für die danach Zahlungspflichtigen der

Gruppe der „Nicht-KKW-Betreiber“ (§ 6 Abs. 1 Nr. 2 lit. c)

EndlagerVlV) seit geraumer Zeit nicht mehr aufwandsgerecht.

Denn der diesem Schlüssel zugrundeliegende

Sachverhalt hat sich geändert: Das Endlager Konrad

für schwach- und mittelradioaktive Abfälle wurde 2007

rechtskräftig planfestgestellt, der „Schlüssel“ in § 6

­EndlagerVIV – und dementsprechend auch die Umlagebescheide

des BMU nach StandAG – gehen indes noch

­immer von einer Addition der Abfälle für das Endlager

Konrad und das Endlager für hochradioaktive Abfälle

(HAW) aus („Ein-Endlager-Prinzip“). Seit der Bestandskraft

des Planfeststellungsbeschlusses für das Endlager

Konrad am 26. März 2007, dem Tag der Zurückweisung

der Nicht-Zulassungsbeschwerde gegen das die Klagen

­zurückweisende Urteil des OVG Lüneburg durch das

BVerwG, steht indes fest, dass die Konradmengen nicht im

HAW-Endlager endgelagert werden.

Das StandAG unterstellt indes ausweislich seiner

Gesetzes­begründung – vgl. BT-Drs. 17/13471, S. 19 – dass

nur die nicht in das Endlager Konrad einzulagernden

­Abfälle in das nach StandAG zu suchende Endlager eingebracht

werden sollen. Es verweist in § 29 Abs. 2 StandAG

aber gleichwohl auf den bisherigen Kostenschlüssel in

§ 6 der EndlagerVIV, der diese Trennung gerade nicht

­berücksichtigt. Eine verursachungsgerechte Verteilung

der Kosten hätte mithin seit März 2007 (also vor mehr als

10 Jahren!) erfolgen müssen. Denn eine Addition

der ­Mengen für Konrad und das HAW-Endlager ist

seitdem nicht mehr zulässig. Die Regelung nach § 6 Abs. 3

Satz 3 EndlagerVIV ist mithin nicht aufwandsgerecht.

Diese sachlich überholte Rechtslage führt dazu, dass

die ­Umlagebescheide nach StandAG, die auf den

Kostenschlüssel in § 6 EndlagerVlV aufsetzen, zumeist mit

Rechtsmitteln angegriffen werden.

Die Lösung: Nachträgliche Korrektur

ermöglichen – Richtiger Maßstab fehlt aber

nach wie vor

Das geplante „Kostenanpassungsgesetz“ sieht nunmehr

mit der Neu-Einführung von § 35a StandAG eine Regelung

vor, die die abschließende Berechnung der Umlagebeträge

am Ende der Standortsuche für das Endlager (also nach

2031) vorsieht. In diese Neu-Berechnung können die nach

dem 1. Januar 2021 festgesetzten oder in der Vergangenheit

erhobene Beiträge einbezogen werden, bei denen

die Festsetzung entweder noch nicht bestandskräftig ist

( wegen Einlegung von Rechtsbehelfen) oder bei denen die

Festsetzung rechtswidrig erfolgt ist, also die Voraussetzungen

für eine Rücknahme der Bescheide nach § 48

des Verwaltungsverfahrensgesetzes (VwVfG) vorliegen.

Diese Neuregelung ist im Grundsatz zu begrüßen, weil

damit eine nachträgliche „Umverteilung“ des Kostenaufwands,

also eine Korrektur unrichtiger Beitragserhebungen,

ermöglicht wird.

Die Neuregelung in § 35a StandAG ist aber nicht

­ausreichend: Es fehlt jeglicher Anhaltspunkt sowohl im

Gesetzestext als auch in der Gesetzesbegründung, nach

welchem inhaltlichen Maßstab die spätere Neu berechnung

der Umlagebeträge, d. h. nach der finalen Feststellung des

Endlagerstandorts erfolgen soll. Da eine Veränderung bei

einem Zahlungspflichtigen dazu führt, dass sich die

­Umlagebeträge auch für alle anderen Zahlungspflichtigen

ändern, bedarf es für eine Neuberechnung eines klar

­definierten inhaltlichen Maßstabs. Daran fehlt es indes.

Vielmehr bleibt auch weiterhin der längst überholte

­Kostenschlüssel in § 6 EndlagerVlV maßgebend, wie sich

aus dem Verweis der Neuregelung in § 35a StandAG auf

die unverändert in Bezug genommenen Regelungen in

§§ 29, 31, 32 und 35 StandAG ergibt. Das verwundert

schon. Zumindest in der Gesetzesbegründung wäre ein

Hinweis zu erwarten gewesen, dass der bisherige Schlüssel

aus § 6 EndlagerVlV absehbar novelliert wird, so dass er

auf den Sachverhalt nach StandAG „passt“. Trotz der Neuregelung

in § 35a StandAG ist also festzustellen, dass es

auch danach an einem realitätsgerechten Maßstab für die

Bestimmung der Umlagen nach StandAG fehlt und zwar

nicht nur für die Zeit bis zur finalen Feststellung des

Endlagerstandorts, sondern auch für den Zeitpunkt der

abschließenden Berechnung nach § 35a StandAG selbst.

Keine Zinsregelung vorgesehen

Unter beitragsrechtlichen Gesichtspunkten nicht unproblematisch

ist die Tatsache, dass – trotz des sehr langen

Zeitraums, der noch bis zur endgültigen Berechnung der

Spotlight on Nuclear Law

Cost Correction for Site Selection Procedures – It Remains Confusing ı Tobias Leidinger


atw Vol. 65 (2020) | Issue 8/9 ı August/September

finalen Umlagebeiträge nach § 35a StandAG verstreichen

wird – eine Zinsregelung für überzahlte, d. h. am Ende

zurück zu erstattende Beiträge nicht vorgesehen ist. Die

Gesetzesbegründung zu § 35a StandAG verweist dazu

­lediglich auf § 34 StandAG, also eine unverzinste Rückerstattung.

Damit weicht § 35a StandAG für das HAW-­

Endlager von der Regelung für das Endlager Konrad ab:

§ 8 EndlagerVlV sieht hier eine Verzinsung von 3 % über

dem Basiszinssatz gesetzlich vor, wenn überzahlte Beträge

an die Beitragspflichtigen zurück zu erstatten sind. Warum

dies nicht auch im Rahmen der Erstattung von Zahlungen

nach § 35a StandAG gelten soll, erschließt sich nicht. Eine

Begründung dafür fehlt.

Fazit

Die nach § 35a StandAG eröffnete Korrektur der Kostenverteilung

nach Abschluss des Standortauswahlverfahrens

ist grundsätzlich zu begrüßen. Damit ist es möglich,

­sachliche Veränderungen in die abschließende Umlageberechnung

einfließen zu lassen. Das neue Gesetz lässt

aber den für die finale Berechnung erforderlichen inhaltlichen

Maßstab vermissen. Darüber hinaus bleibt der für

die weiteren Festsetzungs- und Vorauszahlungsbescheide

nach StandAG sachlich längst überholte „Schlüssel“ nach

§ 6 EndlagerVlV auch weiterhin unverändert. Damit sind

entsprechende Umlagebescheide auch zukünftig angreifbar.

Die Anzahl der Rechtsbehelfe wird also kaum

­abnehmen, die Situation für die Beitragspflichtigen bleibt

damit bis auf Weiteres unübersichtlich. Von einer „echten

Lösung“ kann also aktuell keine Rede sein, sie wurde

vielmehr in die weite Zukunft verschoben.

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SPOTLIGHT ON NUCLEAR LAW 393

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Prof. Dr. Tobias Leidinger

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ISSN 1431-5254

Spotlight on Nuclear Law

Cost Correction for Site Selection Procedures – It Remains Confusing ı Tobias Leidinger


atw Vol. 65 (2020) | Issue 8/9 ı August/September

394

DECOMMISSIONING AND WASTE MANAGEMENT

Actual Research and Development

Activities in the Field of Dismantling

Sascha Gentes and Nadine Gabor

The shutdown of nuclear facilities is attracting more and more public attention, not only because of their limited

life cycle, but primarily due to the political decision to phase out nuclear power. For the engineers involved, the

complete deconstruction and decommissioning of such facilities represents an extremely complex problem with

countless constraints and variables which have constantly to be taken into account and incorporated into the process.

Deconstruction work often relies on standard construction equipment, but this has to be modified and refined for each

application and for each part of the structure.

| Fig. 1.

Patented novel milling cutter to remove highly reinforced concrete.

The deconstruction costs run into

several hundred million euros, depending

on the actual design of the

power station, and the demolition

work itself takes about ten years. The

phasing out of nuclear energy and the

attendant switching off of Germany’s

nuclear power stations by 2022 have

focused public attention even more on

this particular issue. More than 440

nuclear power stations are in operation

around the globe, and they will

all need to be decommissioned at

some point. These facts serve to illustrate

the great potential and the vast

amount of research which this field

entails.

The professorship of Deconstruction

and Decommissioning of Conventional

and Nuclear Buildings at

Karlsruhe Institute of Technology

(KIT) was established in 2008 and is

dedicated exclusively to the last life

cycle of a building, its deconstruction

and decommissioning.

There is no doubt that a nuclear

power station can be safely decommissioned

nowadays. Neither is there any

doubt that optimization potential

exists for many technologies and processes.

There is also scope to further

enhance the effectiveness of the automation

and the robotics.

This is precisely where this department

comes in, concentrating its

efforts on R&D projects, and always

working in close collaboration with

industry. These projects can then

be comprehensively tested in the

Institute’s own testing facility.

The individual R&D projects address

the following issues:

p Reduction of secondary waste

p Automation and remote operation

of the processes

p Performance optimization of existing

processes

p Development of new technologies

p Management methods for decommissioning

and deconstruction

By way of example, we present a

brief overview of the DefAhS (Defined

ablation of highly reinforced concrete

structures) project, a collaborative

project between Kraftanlagen Heidelberg

GmbH, Herrenknecht AG, and

KIT. The project objective was the

in-depth removal of highly reinforced

concretes using one single tool and in

one single operation. In 2019, the project

achieved 2 nd place in the Innovation

Awards at the Bauma, the world’s

| Fig. 2.

Detail of the combination of impact cutters and steel cutting inserts.

largest construction machinery trade

fair, and the innovation has been

granted patents in numerous countries.

The equipment innovatively

combines cutting tools for concrete

and steel in one milling drum. For the

first time ever, it is now possible to

automate the deep milling of cracks,

for example.

Another of our current projects is

MASK (Magnetic Separation Method

for the Reduction of Secondary Waste

from the Water Abrasive Suspension

Cutting Technique (MASK)), because

the cutting up and disposal of the

­reactor pressure vessel (RPV) and its

related installations poses a considerable

challenge during the decommissioning

of a nuclear facility. One of

the cutting techniques is the Water

­Abrasive Suspension Cutting Technique

(WASS), which is characterized

by the high degree of flexibility of its

modular application and the fact that

it is impervious to the mechanical and

thermal stresses in the material being

cut. The abrasive that has to be

Decommissioning and Waste Management

Actual Research and Development Activities in the Field of Dismantling ı Sascha Gentes and Nadine Gabor


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 3.

Prototype rig for the separation.

admixed to facilitate the cutting process

combines with the metal cuttings

from the RPV, and this mixture itself

has then also to be disposed of. The

amount of secondary waste produced

thereby is considerable, roughly doubling

the total volume of radioactive

waste. The disposal of secondary

waste is very expensive and so this

method of cutting up the reactor

pressure vessel has fallen out of use

despite its technical advantages.

MASK was preceded by the research

project known as NENAWAS (New

Disposal Methods for the Secondary

Waste of the Water Abrasive Suspension

Cutting Technique), a close

­collaboration between KIT and ­AREVA

GmbH which succeeded in developing

a separation method which can

process the secondary waste from the

water abrasive suspension cutting

technique. It utilizes a prototype magnetic

separation rig to separate the

metal cuttings from the mixture of

steel and abrasive material produced

in the cutting process. The microscopic

analysis of the separated abrasive

shows that it is still contaminated

with metal cuttings, however. Before

the bulk of the secondary waste can be

cleared for release, further investigation

is required, and this is being done

in the current MASK project. The

objective here is to undertake basic

research which allows the quality of

the separation to be optimized to such

a degree that the secondary waste

can be disposed of in a conventional

manner and thus allow the cutting

technique to be used in the future for

| Fig. 4.

Secondary waste: generation.

the large number of decommissioning

projects still to be undertaken. To

this end, further experiments (with

different types of steel to be cut and

different processed mixtures) are

being carried out on the existing test

rig. Moreover, a numerical flow simulation

of the magnetic filter is being

produced. In the controlled area of

the laboratory belonging to our project

partner (Institute for Nuclear

Waste Disposal), the separation is

being tested and evaluated under

realistic conditions with a small,

laboratory-scale test setup with active

and activated materials to assess its

suitability for the treatment of the

radioactive waste.

Robotics is also making greater

and greater inroads into the field of

deconstruction and decommissioning.

This has led to the setting up of

the ROBDEKON (Robotic Systems

for Decontamination in Hazardous

Environments) project, a competence

center dedicated to research into fully

autonomous and semi-autonomous

robotic systems. The aim for the future

is for such systems to carry out

decontamination work autonomously,

obviating the need for humans to

­enter hazardous zones. As of mid-

June 2018, the Federal Ministry of

Education and Research has provided

ROBDEKON with twelve million euros

of funding as part of the »Research

for Civil Security« program. It will

ini tially run for four years, but the aim

is for the competence center to continue

in the long term. ROBDEKON

is coordinated by the Fraunhofer

Institute of Optronics, System Technologies

and Image Exploitation

IOSB. In addition to the Fraunhofer

IOSB, other research institutions

involved in the project are Karlsruhe

Institute of Technology (KIT), the

­German Research Center for Artificial

Intelligence (DFKI), and the FZI

Research Center for Information

Technology. The industrial partners

in the consortium are Götting KG,

Kraftanlagen Heidelberg GmbH, ICP

Ingenieurgesellschaft Prof. Czurda

und Partner mbH, and KHG Kerntechnische

Hilfsdienst GmbH.

The intention is for ROBDEKON

to become the national point of

contact for issues appertaining to

robotic systems for decontamination

in hazardous environments. The

competence center aims to set up a

network of experts and users, and

create an innovative environment for

new technologies for robot-assisted

decontamination for its partners from

science and industry.

ROBDEKON has been established

to explore and develop novel types of

robotic systems for decontamination

tasks. Its research topics are mobile

robots for difficult terrain, autonomous

construction equipment,

robotic manipulators, and also decontamination

concepts, planning

algorithms, multi-sensorial 3D environment

mapping, and teleoperation

by means of virtual reality. Methods

from the field of artificial intelligence

enable the robot to perform the

tasks assigned to it either autonomously

or semi-autonomously. While

DECOMMISSIONING AND WASTE MANAGEMENT 395

Decommissioning and Waste Management

Actual Research and Development Activities in the Field of Dismantling ı Sascha Gentes and Nadine Gabor


atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 396

| Fig. 5.

Prototype of automated milling system.

the competence center is being set up,

the work will initially concentrate on

three relevant areas of application:

the remediation of landfills and contaminated

sites, the decommissioning

and deconstruction of nuclear facilities,

and the decontamination of

facility components. By involving

users at an early stage, we ensure that

practical systems which reduce the

risk for human operators and protect

them from hazards are developed

expeditiously.

The Department of Deconstruction

and Decommissioning of Conventional

and Nuclear Buildings at Karlsruhe

Institute of Technology is to explore

and develop an approach that will

help find a solution which uses robotic

systems for the automated decontamination

and clearance measurement

of building structures in nuclear facilities.

Automated decontamination

focuses on treating a contaminated

concrete wall and contaminated indoor

spaces with the aid of a mobile

work platform.

To this end, the department is

developing and constructing an

automated surveying system which

will utilize the latest measurement

systems to explore its surroundings.

It will display the contaminated locations

on a 3D chart of its environment

as a function of the radiation level,

thus allowing the subsequent decontamination

work to be carried out

with greater efficiency.

This overview is intended to show

that R&D work is still required in the

field of nuclear facility decommissioning,

even in 2020. The emphasis

here is always on the involvement of,

and close collaboration with industry,

since this is the only way to work out a

practical solution.

In addition to issues relating

­spe­cifically to nuclear facilities, great

importance is also attached to all

aspects concerning the demolition of

conventional buildings. Some brief

details are provided below. Basically,

the waste products from the demolition

process (asphalt, concrete, masonry,

asbestos, man-made mineral

fiber, …) are categorized as either

hazardous or non-hazardous waste.

The laws, ordinances and guidelines

are extremely extensive, specific to a

particular federal state in some cases,

and subject to continual revisions

and amendments. The thresholds

above which asphalt is classified as

hazardous waste differ by several

­hundred mg PAH/kg across the individual

federal states, for example.

The figures stated below show

just how important the demolition

of conventional buildings is. In

­Germany, around 209 million tonnes

of waste are classified as „construction

and demolition waste“ every year. In

2015, this amounted to over 50 percent

of the total waste produced

(402.2 million tonnes). In conjunction

with the Circular Economy Act (Kreislaufwirtschaftsgesetz)

and the stipulations

on recycling rates, this fact

clearly emphasizes the considerable

research potential in this field. The

research topics addressed are the

automated separation of different

types of waste, optimization potentials

in relation to environmental

release and pollution during demolition,

and also automation and remote

handling. This relates in particular

to the handling of “hazardous waste”.

The task is to recognize these potentials,

develop optimization approaches,

and implement pilot projects

with a specific objective.

Selective demolition requires that

questions about how to handle

hazardous waste and pollutants, and

how to comply with the stipulations

regarding type-specific collection and

disposal of the demolition material,

be clarified in advance. All these

issues mean the demolition of conventional

buildings is an exciting field

with extensive research potential.

The redevelopment of existing

building structures and building on

“brown field sites”, too, will become

ever-expanding fields of research in

the future. To do justice to the

demands, the Department of Deconstruction

and Decommissioning of

Conventional and Nuclear Buildings is

addressing this future-oriented field

in its research, science and teaching.

Now more than ever, the graduates of

today must have a good general

grounding in topics such as the construction,

operation, and also the

decommissioning of buildings. To

illustrate this, we have included part

of a survey carried out at universities

and universities of applied sciences

on the topic of “demolition-related

course content in civil engineering”.

The survey reveals that “demolition”

and “decommissioning” and other

related subjects are not covered in

detail. Lectures on more detailed

issues of decommissioning and deconstruction

are not usually part of the

curriculum.

This is precisely where we come in,

by offering lectures which train young

graduates in these important and

sustainable subjects.

The deconstruction and decommissioning

of conventional and

nuclear buildings is therefore a

discipline whose graduates will be

in great demand and which will

guarantee secure employment in the

future.

Authors

Prof. Dr.-Ing. Sascha Gentes

sascha.gentes@kit.edu

Dr.-Ing. Nadine Gabor

Karlsruhe Institut of Technology

Institute of Technology and

Management in Construction

Am Fasanengarten

76131 Karlsruhe, Germany

Decommissioning and Waste Management

Actual Research and Development Activities in the Field of Dismantling ı Sascha Gentes and Nadine Gabor


atw Vol. 65 (2020) | Issue 8/9 ı August/September

A Geopolymer Waste Form for

Technetium, Iodine and Hazardous Metals

Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg

We have developed geopolymer waste forms for the solidification of waste streams (Table 1)

that will be generated at the Hanford site in the State of Washington, USA. At the Hanford site,

about 205,000 m 3 of liquid radioactive waste is stored in 177 underground tanks. These wastes will

be immobilized at the US Department of Energy’s (DOE) “ Hanford Tank Waste Treatment and

Immobilization Plant” (WTP), which is under construction. The wastes will be separated into

high-level waste (HLW) and low-activity waste (LAW), both of which will be vitrified. The vitrified

highlevel waste will be disposed in an offsite underground repository. The vitrified LAW will be

disposed onsite in the “Integrated Disposal Facility” (IDF).

Vitrification of LAW and HLW will generate secondary ­liquid

wastes, called Hanford secondary waste (HSW), which includes

process condensates and liquid effluents from ­off-gas

treatment systems. HSW will be sent to an ­effluent treatment

­facility (ETF) for further treatment and solidification

(not vitrification) and then disposed in the IDF as well [1].

To support the evaluation and selection of waste forms

suitable for the solidification and disposal of ­secondary

waste streams from the WTP, the Pacific Northwest ­National

Laboratory (PNNL) in Richland, WA, conducted a waste

form testing program for the DOE [2, 3]. Our geopolymers

(patented as DuraLith [4]) and two other materials (cast

stone [5], and ceramicrete [6]) par ticipated and were tested

under the conditions laid out by PNNL. Cast Stone consists

of Portland cement, ground granulated blast furnace slag

and Class F fly ash (about 10, 45, 45 wt. %) and Ceramicrete

of 12.5 wt. % MgO, 42.5 wt. % KH 2 PO 4 and, 45.0 wt. %

Class C fly ash. For the DuraLith geopolymer the most

important results will be reported here.

Materials and methods:

DuraLith geopolymer waste forms consist of two binders,

i.e., ground granulated blast furnace slag (FS) and

metakaolin (MK), an additional source of silica, river sand,

and an activator, here the waste solution with extra sodium

hydroxide added. To support fixation of 99 Tc (replaced by

ReO 4 - ) SnF 2 was added to reduce Re 7+ (Tc 7+ ) to Re 4+

(Tc 4+ ), which forms insoluble ReO 2 (TcO 2 ). IONEX Ag 900,

an Ag-zeolite (Ag-Z) was used to precipitate I - as AgI in the

nano pores of the zeolite.

HSW will be generated as soon as vitrification operations

begin. Table 1 shows four projected compositions of

HSW streams (S1 to S4) that were provided for testing by

PNNL. S1 to S3 are alkaline solutions (high OH - ) and of

similar composition. S4 is not alkaline and contains

ammonia salts and has the highest concentrations of

­iodine and technetium. All wastes contain the heavy

­metals Cr, Cd, Pb, Ag, As, Hg with concentrations ranging

from about 1 ppm (Cd) to 1000 ppm (Cr). Mercury was not

included in this study for safety reasons.

The solid constituents were mixed and the activator

solution was added. Mixing was continued until a pour able

paste formed, which took a few minutes. The paste was

transferred into cylindrical plastic molds where it hardened

within 1 to 3 hours, depending on composition. All samples

were cured at room temperature for at least 28 days before

tests were conducted. No bleeding, swelling, salt deposition,

or cracking was observed. Example recipes for MKand

FS-based waste forms used for waste stream S1 are

Planned entry for

shown in Table 2 for a typical lab-scale batch of about 4 kg.

Loading of waste solids was about 2.5 wt. %.

IDF waste acceptance requires that materials testing

follows the pro cedures referred to below, that the quality

assurance programs NQA-1 is in place, and that the laboratory

is audited. Our laboratory meets these requirements.

Compressive strength was measured after 28 days

­following ASTM C39. Pieces of the crushed cylinders were

used to prepare the required particle size fraction for the

“Toxicity characteristic leaching procedure” (TCLP),

SW846 Test Method 1311. This procedure was applied to

measure the release of hazardous metals. The method

yields data that can be compared to the U.S. Environmental

Protection Agency”s Universal Treatment Standards

(UTS) [7], which are part of the waste acceptance criteria

for HSW waste forms at the IDF. Direct current plasma

atomic emission ­spectroscopy (DCP-AES) was ­employed to

determine the release of Cr, Ag, As, Cd, Cu, Pb, Sn from the

DuraLith waste forms during leaching. A small amount of

the crushed cylinders was used to prepare samples for

SEM/EDS analysis.

Best Paper

Award

The papers

“A geopolymer waste

form for technetium,

iodine and hazardous

metals” by

Werner Lutze,

Weiliang Gong, Hui

Xu and Ian L. Pegg

and “Code and data

enhancements of the

MURE C++ environment

for Monte-Carlo

simulation and

depletion” by

Dr. Maarten Becker

(featured in previous

atw) have been

awarded as

Best Papers of

KERNTECHNIK 2020,

which unfortunately

had to be cancelled

due to Covid-19.

Constituents S1 S2 S3 S4

Al(OH) 3 9.39E-02 1.14E-01 9.22E-02 4.24E-02

Si 1.88E-03 2.04E-03 7.74E-04 1.39E-02

K 5.82E-04 6.51E-04 2.18E-03 2.87E-02

NH 4

+

- - - 4.41E-01

OH - 3.98E-01 4.35E-01 2.45E-01 1.02E-08

NO 3

-

3.28E-01 1.90E-01 3.97E-01 1.13E+00

CO 3

2-

2.28E-02 4.66E-02 3.94E-02 1.04E-02

Cl - 2.25E-02 2.17E-02 2.91E-02 1.04E-02

NO 2

-

1.20E-02 1.05E-02 3.83E-02 4.31E-02

PO 4

3-

6.87E-03 4.85E-03 6.03E-03 5.10E-03

SO 4

2-

4.41E-03 5.81E-03 5.14E-03 4.36E-02

F - 5.57E-04 3.75E-04 4.42E-04 1.02E-08

Cr 2.03E-04 2.03E-04 2.03E-04 1.09E-03

Ag 6.27E-06 6.27E-06 6.27E-06 2.35E-05

As 3.48E-05 3.48E-05 3.48E-05 1.61E-05

Cd 1.57E-06 1.57E-06 1.57E-06 2.16E-05

Hg 1.13E-05 1.13E-05 1.13E-05 5.30E-06

Pb 8.99E-06 8.99E-06 8.99E-06 8.28E-06

Tc 1.81E-05 1.81E-05 1.81E-05 5.59E-04

I - 4.62E-06 4.62E-06 4.62E-06 6.29E-05

Total organic carbon 9.39E-02 1.14E-01 9.22E-02 4.42E-02

| Tab. 1.

Compositions of Hanford secondary waste stream simulants (mol/l).

DECOMMISSIONING AND WASTE MANAGEMENT 397

Decommissioning and Waste Management

A Geopolymer Waste Form for Technetium, Iodine and Hazardous Metals ı Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg


atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 398

Order

of addition

Chemical/

material

Assay

S1-2xMKR

Mass (g)

S1-2xFSR

1 HSW simulant 1.00 658.4 654.7

2 SnF2 0.98 15.0 15.0

3 KOH 0.90 499.4 232.6

4 NaOH 0.98 59.5 118.6

5 Fumed silica 0.96 466.1 333.2

6 Metakaolin 0.96 833.5 477.9

6 Furnace slag 1.00 547.5 1183.8

6 Fine river sand 1.00 733.0 723.3

6 Ag-Z 900 1.00 38.6 38.1

7 Silica fume filler 1.00 77.2 38.1

| Tab. 2.

Example recipes for FS-based and MK-based DuraLith waste forms applied to S1.

Sample ID Ag As Cd Cr Cu Pb Sn

S1-2xFS


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Table 4 shows the results obtained with the ­ANSI/ANS-

16.1 test for the MK-based geopolymer samples. The results

for the FS-based samples are similar. Results are

shown for rhenium and sodium. Iodine will be discussed

separately. Most of the LI values for rhenium are >9, i.e.,

greater than the required lower limit of 9.0 for technetium.

The LI values for sodium are about 9, i.e., three orders of

magnitude higher than the required minimum value of 6.

The objective of this work was to develop a waste form

that is easy to make at ambient temperature and exhibits

high retention for technetium, iodine and the hazardous

metals contained in HSW. To find out how well the

simulation of technetium by rhenium would be, we shared

a MK-based DuraLith recipe with Pierce et al. [8] who

conducted testing with 99 Tc. These authors used the EPA

Method 1315, which is equivalent to ANSI/ANS-16.1.

Pierce et al. [8] reported an average leaching index of

about 11, which is one to two orders of magnitude higher

than our index for Re. In another study, Mattigod et al. [9]

conducted leaching experiments with 99 Tc and Re, using

one of our FS-based DuraLith recipes. The authors

concluded that rhenium does not simulate technetium

very well. The leachability indices for Re were one to three

orders of magnitude lower than those for 99 Tc.

Ag-Z was employed to fixate iodine by precipitating AgI

in the nanopores of the zeolite. We found that at least some

of our chemicals and/or raw materials used to prepare the

geo polymer waste forms contained iodide as an impurity.

The level of the contamination was significant enough to

compromise the ANSI/ANS-16.1 leach tests. Therefore,

the results for iodine could not be quantified correctly.

Pierce et al. [8] reported an iodine LI of around 7 for a

MK-based ­DuraLith geopolymer for S1 with Ag-Z, spiked

with stable 127 I. This value is com parable to data collected

by Pierce et al. for two other waste forms, Cast Stone and

Ceramicrete [8, 10, 11]. None of these waste forms met the

minimum iodine LI of 11 for disposal at the IDF. An evaluation

by PNNL concluded that Cast Stone should be the

­final choice, because of its greater maturity concerning

large-scale production of the waste form.

We have studied iodide fixation by Ag-Z directly in the

waste solution (S1). For example, with a molar ratio

Ag/I = 5, Ag-Z sequestered 99.9 % of the iodide in the

simulated HSW solution (2M Na + ), spiked with 100 ppm

iodide. Though silver-based scavengers are very efficient in

removing iodine from HSW, its long-term stability may be

impaired as the AgI encapsulated in the zeolite may be

destabilized by sulfide released ­during alkali-activation of

blast furnace slag. More work needs to be done to evaluate

the significance of this process. ­Solidification of the waste

stream ­after iodide separation may yield a sufficiently high

LI for the geo polymer.

Scale-up testing:

The feasibility of larger-scale batches of FS-based geopolymer

waste forms was tested [12]. For example, a batch

was produced to fill a 55-gallon (208 liters) drum.

Core-drilling and cutting showed that a visibly homogeneous

monolith was obtained. Further more, on the

same scale we studied effects of composition changes on

workability and initial setting with 6 M Na + HSW simulant

(S1). Blast furnace slag varied in concen tration and parts

of the slag were ­replaced by Class F fly ash (ASTM C618).

The less reactive class F fly ash extended the workable time

(up to the time of initial setting) up to 15 hours, depending

on the weight fraction of fly ash. This ­observation gave rise

to an in-depth study in the laboratory on effects of composition

on pro duction-related and some other properties

Sample ID Concentration (mg/l) Leachability index

of these geopolymer waste forms. This study has been

completed and will be reported elsewhere.

Conclusions

A geopolymer waste form was ­de­veloped to immobilize

Hanford secondary lowlevel radioactive waste streams.

Two other materials were ­under consideration for a final

selection. The final decision was between Cast Stone and

the geopolymer, which performed equally well. Cast Stone

was selected because of greater experience with technicalscale

production. None of the waste forms complied with

the required degree of fixation of iodine. We have shown

that iodine separation by Ag-Z prior to ­solidification is very

effective. The decontamination factor is about 10 3 . We

have not yet tested whether adding the precipitated iodine

during solidification of the depleted waste stream would

increase LI to 11.

References

[1] Pacific Northwest National Laboratory, Hanford Site Secondary Waste Roadmap, PNNL-18196

(2009) Pacific Northwest National Laboratory, Richland, WA

[2] R.L. Russell, M.J. Schweiger, J.H. Westsik, Jr., P.R. Hrma, D.E. Smith, A.B. Gallegos, M.R. Telander,

S.G. Pitman, Low temperature waste immobilization testing, PNNL-16052 Rev 1 (2006), Pacific

Northwest National Laboratory, Richland, WA

[3] E.M. Pierce, R.J. Serne, W. Um., S.V. Mattigod, J.P. Icenhower, N.P. Qafoku, J.H. Westsik, Jr., R.D.

Scheele, Review of potential candidate stabilization technologies for liquid and solid secondary

waste streams, PNNL-19122 (2010), Pacific Northwest National Laboratory, Richland, WA

[4] W. Gong, W. Lutze, I.L. Pegg, U. S. Patent No. 7,855,313 B2 (2010)

[5] S.K. Sundaram, K.E. Parker, M.E. Valenta, S.G. Pitman, J. Chun, C.-W. Chung, M.L. Kimura, C.A.

Burns, W. Um, J.H. Westsik, Jr., Secondary waste form development and optimization -Cast

Stone, PNNL-20159 (2011), Pacific Northwest National Laboratory, Richland, WA

[6] D. Singh, R. Ganga, J. Gaviria, Y. Yusufoglu, Secondary waste form testing: Ceramicrete

phosphate bonded ceramics, ANL-11/16 (2011), Argonne National Laboratory, Chicago, IL

[7] 40 CFR 268. 2002, Land disposal restrictions; Code of Federal Regulations, U.S. Environmental

Protection Agency, Washington, DC

[8] E.M. Pierce, R.J. Serne, W. Um., S.V. Mattigod, J.P. Icenhower, N.P. Qafoku, J.H. Westsik, Jr., R.D.

Scheele, Review of potential candidate stabilization technologies for liquid and solid secondary

waste streams, PNNL-19122 (2010), Pacific Northwest National Laboratory, Richland, WA

[9] S.V. Mattigod, J.H. Westsik, Jr., C.W. Chung, M.J. Lindberg, and K.E. Parker, Waste acceptance

testing of secondary waste forms: Cast Stone, Ceramicrete and DuraLith, PNNL-20632 (2011),

Pacific Northwest National Laboratory, Richland, WA

[10] S.K. Sundaram, K.E. Parker, M.E. Valenta, S.G. Pitman, J. Chun, C.-W. Chung, M.L. Kimura, C.A.

Burns, W. Um, J.H. Westsik, Jr., Secondary waste form development and optimization -Cast

Stone, PNNL-20159 (2011), Pacific Northwest National Laboratory, Richland, WA

[11] D. Singh, R. Ganga, J. Gaviria, Y. Yusufoglu, Secondary waste form testing: Ceramicrete

phosphate bonded ceramics, ANL-11/16 (2011), Argonne National Laboratory, Chicago, IL

[12] G.B. Josephson, J.H. Westsik, Jr., R.P. Pires, J.L. Bickford, M.W. Foote, Engineering-scale

demonstration of DuraLith and Ceramicrete waste forms, PNNL20751 (2011), Pacific Northwest

National Laboratory, Richland, WA

Authors

Prof. Werner Lutze

Weiliang Gong

Hui Xu

Ian L. Pegg

The Catholic University of America

620 Michigan Ave NE

20064 Washington, DC, USA

Na Re Na Re

S1-2xMKR-L01 4.56 0.004 9.6 10.1

S1-2xMKR-L02 5.30 0.002 9.3 10.6

S1-2xMKR-L03 9.91 0.003 9.3 10.8

S1-2xMKR-L04 10.24 0.016 9.2 9.3

S1-2xMKR-L05 7.86 0.041 9.2 8.2

S1-2xMKR-L06 14.60 0.041 8.5 8.1

S1-2xMKR-L07 90.82 0.041 8.4 8.0

S1-2xMKR-L08 90.86 0.078 8.7 9.3

S1-2xMKR-L09 114.2 0.077 8.7 9.5

S1-2xMKR-L10 127.11 0.073 8.6 9.6

| Tab. 4.

Results of ANSI/ANS-16.1 leach test for selected DuraLith waste forms.

DECOMMISSIONING AND WASTE MANAGEMENT 399

Decommissioning and Waste Management

A Geopolymer Waste Form for Technetium, Iodine and Hazardous Metals ı Werner Lutze, Weiliang Gong, Hui Xu and Ian L. Pegg


atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 400

Planned entry for

Decommissioning of Nuclear Power

Plants: Waste Streams and Release

Measurements

Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann

The material from nuclear power

plants must be checked for radioactive

contamination before it can be recycled.

Roughly spoken, three classes

exist in respect to radioactivity:

release material, material with short

living radioactive isotopes and material

with long living radioactive isotopes

like Uranium or Plutonium.

The release measurement procedure

is of highest importance because

the material can be immediately

recycled and does not need special

handling and storages.

This paper will give a survey over

measurement methods, performance

consideration, data management and

quality control. All these topics have

also to be considered under economical

aspects.

Measurement methods

The inventory of nuclear power plants

consists of isotopes emitting gammas,

beta particles, alpha particles and

neutrons (other modes like internal

conversion, spontaneous fission etc.

are not considered here).

Alpha particles are difficult to

measure with high sensitivity because

of the high stopping power by material

or air. Neutrons can easily penetrate

material, but the detection limit

is very high because of the weak

During the next 20 years a large number of nuclear installations have to be taken out of operation

and have to be decommissioned up to the level of “green field”. According to German laws, all

material should be recycled, whenever possible. Otherwise the material has to be transported to

long term storages or disposals, which are actually not planed finally. From the economical point of

view, the recycling is therefore the preferred solution.

interaction with the detectors. For this

reason, alphas and neutrons are rarely

measured. This issue can be solved,

because almost all isotopes emitting

alphas or neutrons also emit gamma

radiation. The situation concerning

beta particles is similar. The main

difference is that there are some isotopes

emitting betas, but without

measurable gammas like Sr-90. In

this case the key nuclide method

must be applied, which means, that

gamma emission of those isotopes is

measured, which mostly occur

together with the hidden beta emitter

(e.g. Cs-137 is a key nuclide for Sr-90).

In principle the betas can be

measured with beta counters, but this

is only possible for isotopes near the

surface of the material and is not good

enough to estimate the total beta

emitters in the waste.

In conclusion the measurement of

gamma emitters is the best way

because it is non-invasive method and

is partly isotope specific. There is also

the possibility to analyse the material

in a radiochemical laboratory, where

alpha or neutron emitters can be

measured with high sensitivity and

precision. But this is not a 100 %

­efficiency measurement. Therefore,

the radiochemical analysis is only

used for quality control, where the big

effort for sample taking, transport to

the laboratory, waiting for results has

not the crucial role.

To measure the gamma emission

detectors are required, which have

high sensitivity and high energy

resolution. The energy resolution is

important to identify the gamma

emitting isotopes and to measure

gamma lines with low detection

limits. Nowadays only one detector

type exists that completely fulfils

these requirements: the HPGe detector.

The price which must be paid for

its excellent characteristics is the

cooling of the detector crystal. This

can be made by using liquid Nitrogen

or by electric cooling systems. Twenty

years before, these detectors were

utilised only in laboratory environments

because of their sensitivity to

sound waves and electric fields from

servo-drives, etc. Actually, devices are

available which are resistant against

rough environmental conditions and

ready for field application.

The measurement, data readout

and spectra analysis can be performed

automatically, including the un folding

of spectra with overlapping peaks

from different isotopes. Different

programs to analyse the spectra,

which were tested by the IAEA [1], are

existing today. One of these programs

| Fig. 1.

Waste ideal to build a waste stream (after processing: i.e. crushing, sieving etc.).

| Fig. 2.

Waste to be pre-processed before building a waste stream.

Decommissioning and Waste Management

Decommissioning of Nuclear Power Plants: Waste Streams and Release Measurements ı Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 3.

Automated system for crashing, sieving, weighing and filling on a conveyor belt.

is the program GAMMA-W, which was

upgraded to calculate uncertainties

according the current norm DIN ISO

11929 [2].

Waste streams

The measured activities do not only

depend on the intensity of measured

signals of the detector, but also

from absorption effects of the waste.

Therefore, the measured activities

have to be corrected to compensate

such ­influences. Considering large

quantities of waste under measurement

during decommissioning, waste

streams have to be defined, which

consist of similar materials in respect

to absorption properties and mass

densities (Figure 1 and Figure 2).

Inside a stream similar correction can

be applied. Typical streams are rubble

from building demolition, soil, metal

waste, etc. The homogeneity of the

material can be enhanced by crushing

and sieving of the material (Figure 3).

The material preparation for the

measurement is an important aspect

because this influences the reliability

of the results and allows increases the

material throughput.

For large quantities two solutions

exist in respect to the measurement

arrangement: The first solution is the

| Fig. 4.

Drum Measurement System.

filling of the material from a stream

into drums with a typical volume

between 200 l and 400 l. The drums

are placed on a rotary table and the

HPGe detector is placed at the side of

the drum (Figure 4).

Different measurement modes are

possible like the measurement of the

whole drum or a slice measurement

by using a horizontal slit collimator

and changing the HPGe position vertically.

An alternative solution is the filling

of the material on a conveyor band as

a thin (typically 10 cm high) layer

(Figure 5). The material moves by the

conveyor and passes the detector

system mounted above the conveyor

(Figure 6). The absorption effects are

minimized because no steel is used

like in the case of drums and the waste

thickness is small. To enhance the

­detection efficiency, the detector

system can be constructed by an array

of HPGe detectors.

The advantage of this construction

is the high material throughput of up

to 100 tons per hour (practical value

of tests at Hanau, Germany). The

disadvantage is that the measurement

time cannot be changed freely during

the measurement. Therefore, it is

important to run the system with

material of the same material property

(i.e. from the same waste stream).

The throughput for a drum configuration

is smaller: using a 400 l

drum filled with material having a

density of 2.3 g/cm 3 , measuring

3 minutes and with additional

2 ­minutes for drum handling (transport

to the turn table, positioning at

the table, removing from the table and

transport to a store position) gives a

throughput of 400*2.3/1000/5*60

= 11 tons per hour. This seems to be

an upper limit where the drum filling

and emptying procedure are not taken

into account.

The values as given above are

taken from a realistic scenario. The

requirements in respect to detection

limits, matrix properties of the waste

and configuration details like the

number of detectors etc. can influence

the throughput.

The metal waste can have three

­different final destinations: unrestricted

use, melting and filling at

disposal sites. In all these cases the

waste is released after measurement

of radioactive materials.

For the different sub-classification

of released material (i.e. free release,

restricted release, disposal) the

measured data have to be analysed.

The analysis starts with the data taken

DECOMMISSIONING AND WASTE MANAGEMENT 401

| Fig. 5.

Rubber (left) and soil* (right) filled on a conveyor belt (*with courtesy of FBFC International).

Decommissioning and Waste Management

Decommissioning of Nuclear Power Plants: Waste Streams and Release Measurements ı Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann


atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 402

| Fig. 6.

HPGe detectors, liquid Nitrogen cooling (left), electrical cooling (right).

for individual batches: a batch can be

a drum filled with waste or a certain

section of the conveyor belt. The batch

is characterised by the material and

the mass of waste contained. After

that, for each batch the activity limit is

calculated as the upper limit under

consideration of all possible errors.

This limit value must be lower than

the limit values as given by law.

If batches with very low limits are

separated from batches with higher

limits and if the two groups are filled

into separate containers, then the

content of the low value container

may have significantly lower activity

limits compared with the individual

batches. The reason is that the average

mass is increased and equalises the

fluctuation of the values of the individual

batches. In case of the use of

conveyor belts, this analysis can be

made online and can separate the

batches into streams for the different

release classifications.

Up to this point, data (i.e. spectra)

are analysed from measurement of

the complete batch. It is also possible,

to read out the spectra in short time

intervals and analyse the measured

data in respect to Hot Spots. This

special analysis is parallel to the batch

activity analysis. If Hot Spots are

found, the batch is separated, the Hot

Spot can be eliminated by e.g. hand

measurements and the rest of the

batch is given back for a new run.

Quality control

The measurement results depend

from the programs used to analyse

measured data and from the error estimation.

Because the detection limits

for release waste are very low, small

differences in the used algorithms or

in the method of error calculation may

influence the decision for release.

Therefore, it is very important to

check the quality of measurements

independently. This can be done by

sampling and analysing in a radiochemical

laboratory. As long as the

activities measured in the radiochemical

laboratory are lower than

the values from the batch measurements

there is no reason to distrust

the values from the batch measurements.

Otherwise, safety factors have

to be introduced into the data analysis.

The details of data analysis and

result interpretation have to be agreed

with the state authorities.

Economic considerations

The release measurement is a balance

act between environmental requirements

(protecting the population

from radioactive radiation) and financial

expenses for the measurement.

The most important parameters are

detection limits and throughput.

It must be considered that detection

limits depend not only from detector

sensitivities and matrix absorption

effects, but also from the natural

occurring radioactive materials at the

location where the measurement system

should be erected. If necessary

special shielding has to be provided.

The throughput depends not only

from a high-speed measurement but

also from the speed of material preparation

like crashing, sieving or

weighing of the batches and from

the removal of the material after

measurement. The slowest process

­defines the throughput.

The investment for the equipment

is high and therefore it has also to be

considered what to do with the

equipment after the measurement

campaign. A modular structured

system which can be containerised

for easy transport is a good solution

to use the equipment for a long time

period. The concept for decontamination

of the equipment has also to be

included in the considerations.

Conclusion

For the release measurement of a

large amount of waste there are two

general solutions: drum systems and

conveyor belt systems. Both solutions

are complementing each other. The

conveyor belt system is favoured in

respect to the highest throughput.

The waste stream concept enhances

the quality of the measurement and

optimises the throughput at the same

time. From economical point of view,

the careful planning of material

throughput as well as the reuse of the

installed systems for next projects

helps to save financial resources. The

installations must be carefully designed

and agreed with the responsible

authorities. Data available for commissioned

installations are helpful to

optimize the layout.

References

[1] M. Blaauw et al., The 1995 IAEA intercomparison of gamma

ray spectrum analysis software, Nuclear Instruments and

Methods in Physics Research A 387 (1997) 416-432

[2] Dr. Westmeier GmbH, Private information from Dr. Westmeier,

Ebsdorfergrund-Mölln, Germany

Authors

Dr. Marina Sokcic-Kostic

marina.sokcic-kostic@

nukemtechnologies.de

Dr. Christoph Klein

Dr. Frank Scheuermann

NUKEM Technologies Engineering

Services GmbH

Industriestr. 13

63755 Alzenau, Germany

Decommissioning and Waste Management

Decommissioning of Nuclear Power Plants: Waste Streams and Release Measurements ı Marina Sokcic-Kostic, Christoph Klein and Frank Scheuermann


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Ventilation Concepts

for Nuclear Decommissioning

Tobias Finken and Peter Hausch

During decommissioning, both the functions and functional areas of the various building

­sections must to some ­extent be considered in significantly different ways and extensively adapted

to the requirements of the decommis sioning. Particularly in the case of nuclear facilities, the

planning of which often did not take into account the eventual needs of decommissioning, there are

great challenges in making such buildings as safe as technically possible during decommissioning.

Importantly, these considerations must also cover the air distribution concepts of the building

sections. The original ventilation must be shut down before the power plant is decommissioned

because it is necessary to put in place a new air distribution concept that suits all necessary requirements

for the period to follow.

In order to design an air distribution

concept for the decommissioning

phase, a detailed inventory of both the

previous ventilation system and all

components of the building is necessary.

Moreover, the decommissioning

concept itself must be analysed in

terms of what ventilation technology

it may require. As the work pro­gresses,

an increasingly large proportion of the

total area is incorporated from the

black zone into the white zone.

Inside the white zone, which has

already been made free of contaminants,

the focus is on the building

itself and on it lasting through the

decommissioning phase, which may

take several years. For the actual areas

of operation, the focus is on the safety

of the people working there and,

on the buildings’, immediate surroundings.

All areas of the system

must be designed using a suitable

control concept so that, on the one

hand, safety is guaranteed at all times

even if individual parts fail and, on the

other hand, interdependent system

components are automatically shut

down in the event of partial failures.

In addition to the applicable legal

regulations, aspects of public perception

and opinion must also be

taken into account in nuclear engineering

projects. Especially given the

awareness that the building exhaust

automatically means some interaction

with the environment, there is a

legitimate public interest that, in

­addition to the best possible filtering

of the air, there is also continuous

monitoring so that any abnormality,

even if it is within the legal limits, is

noted and, in case of any doubt,

triggers a safe shutdown of the system

with a cause analysis to follow.

Another aspect that distinguishes

nuclear projects from conventional

applications is the necessary period

of observation. There can be a

| Fig. 1.

Schematic representation of the entire building.

comparatively long period of time

between the start of planning and its

implementation.

If further technological developments

that can better increase safety

take place within this time frame, these

developments must be taken into

consideration in the overall concept.

As a last point to be considered, the

costs associated with implementing

air distribution systems must not be

overlooked. For nuclear engineering

projects, these generally take a back

seat to safety aspects, but the public

may have legitimate questions about

the use of public funds, be it through

direct expenditure or necessary

subsidies.

If one now takes a closer look at the

air distribution requirements during

decommissioning, there are two main

areas of application:

p The ventilation of the buildings

themselves with their different

areas (permanent negative pressure

separation of areas via

pressure cascades), and

p Local work area-related ventilation

with detection and filtering of

radioactive particles and potentially

contaminated dust particles

as close as possible to the point of

release.

Planned entry for

In the following, both fields of application

will be explained in more detail

using specific examples.

Figure 1 shows a schematic representation

of a cross-section of a reactor

building and auxiliary buildings with

an air distribution concept suitable

for decommissioning already implemented.

Since the focus here is no

longer on the comfort of the indoor

climate, additional processing of the

supply air can often be dispensed

with. Instead, additional appropriately

dimensioned inflow openings are

added to the buildings. Here the air

enters the building through simple

­filter walls. Decen­tralized, mobile

­HEPA filter de­vices come into use

directly wherever the decommissioning

or dismantling of old systems is taking

place. In addition, an exhaust air

system is connected from the outside in

a separate con tainer building to ensure

continuous negative pressure and

­filtered air ­discharge from the building

in order to prevent carryover of radioactive

particles. The machine exhaust

air shown in the ­figure is used to filter

the total air inside the building.

Figure 2 shows a top view of

a newly constructed exhaust air

container. This is a completely prefabricated

module with a filter wall

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Ventilation Concepts for Nuclear Decommissioning ı Tobias Finken and Peter Hausch


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DECOMMISSIONING AND WASTE MANAGEMENT 404

| Fig. 2.

Exhaust air building (container).

made of F9 + H13 filter elements and

a fan that blows directly into a newly

constructed exhaust air stack. In order

to ensure that the building exhaust

air released into the environment is

actually safe, a fail-safe measurement

system for identifying radioactive

paricles in the air has been integrated

into the exhaust air stack ( Figure 3).

| Fig. 3.

Measuring system for monitoring of the central building exhaust air.

By installing pressure relief dampers,

pressure cascades and differing

areas of use can be put into effect at

different phases of the decommissioning.

For the safety of the people working

in black zones, the most important

point during decommissioning and

dismantling is that the large amounts

of dust present there accumulate

safely in order to ensure exhaust air or

recirculated air that is completely free

of contamination.

While particulate matter filters

(HEPA filters) are able to filter even the

smallest particles


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 5.

Bag in/bag out technology on particulate removal and when changing the filter.

| Fig. 7.

Pressure curve of differential pressure across filters in the test with plasma cutting.

­necessary filter changes, through

its recleaning mechanism. Figure 5

shows how the safe changing of

the filter cells takes place using bag

in/bag out technology. Here too, it is

very important to consider unplanned

operating situations and their associated

risks. For systems that are

connected to a central exhaust air

system, there is the risk that negative

pressure in the housing could develop

due to a leak or a defective flap. The

grooves into which the fastening

bands are tensioned are therefore

shaped in such a way that they

can hold onto the protective bag

even when it is sucked into the filter

housing.

For the optimal design of the system

for online cleaning of the filter

cells, it is first important to consider

the buildup of dust deposits on the

surface of the filter material. Figure 6

shows schematically how such dust

deposits are distributed during the

cleaning process. It has been tested

that the high working pressure of >5

bar is necessary for the dust to be effectively

removed from the tightly

folded filter paper. Since there is

always a small amount of dust in the

| Fig. 6.

Build-up on the filter material.

filter cell, a ­special dust is applied as

precoating when it is first put into operation

and after a filter change. This

forms the base layer on the filter and

enables more effective cleaning of the

actual radioactive dust.

To determine the most sensible

parameters for a basic setting, investigations

were carried out using plasma

dust from a cutting process. The

diagram in Figure 7 shows the

­pressure curve of a recleanable filter

stage over several hours of continuous

operation.

As can be clearly seen, the chosen

triggering pressure difference of

1500 Pa leads to a uniform cleaning

frequency and to a constant negative

discharge pressure. With additional

offline cleaning at the end of each

assignment, the system can virtually

return to its initial pressure loss of

ca. 500 Pa. In this way, only the radioactive

dust is collected, and the filter

cell does not have to be replaced.

To ensure safety, numerous parameters

are continuously monitored.

These include the filter load ­capacity

and filter breakage of both stages. If a

filter cell perforates, the system immediately

shuts down automatically.

If the damaged cell is from the first

stage, the automatic cleaning that

usually occurs after the system stops is

also prevented.

For systems that are further

connected to a central exhaust air system,

it can also make sense to connect

them to the central control room in

order to test external release signals or

to react automatically to error messages.

Through the combination of

centralized systems for building air

distribution and local mobile systems

in individual work areas for air conditioning

or for filtering radioactive

dusts close to the place of origin,

overall safe operations of such air

distribution systems can be achieved

both for workers and a plant’s

immediate surroundings. The special

challenges that the decommissioning

phase of a nuclear facility presents can

thus likewise be met.

Authors

Tobias Finken

Dr.-Ing. Peter Hausch

Krantz GmbH

Uersfeld 24

52072 Aachen, Germany

DECOMMISSIONING AND WASTE MANAGEMENT 405

Decommissioning and Waste Management

Ventilation Concepts for Nuclear Decommissioning ı Tobias Finken and Peter Hausch


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DECOMMISSIONING AND WASTE MANAGEMENT 406

Planned entry for

Steam Generator Rip and Ship – a Valuable

Contribution to Decommissioning and

Dismantling of Nuclear Power Plants

Heiko Herbell, Arne Larsson, Gregor Krause and Véronique Bouilly

Practical examples are given from

past projects such as Ringhals NPP

(Sweden) or Stade NPP (Germany) but

also from ongoing developments such

as the Fessenheim decommissioning

plan (France). These examples are

used to understand major challenges

during removal of the SGs and transportation

e.g. shipment of the SGs.

The radionuclide history and

inventory is of major interest for the

further treatment. Necessary input

information for transportation and

offsite treatment of the SG is

described.

Concluding, the advantages are

described for off-site treatment of

large components.

2 Description of rip & ship

concept

The rip & ship concept can be divided

into the rip and the ship part. The

rip part summarizes all activities

necessary to remove the SG from its

original position and place it outside

the containment at ground level.

Here, the ship part begins.

a Removal of steam

generators

Fessenheim decommissioning plan is

described in order to understand the

challenges during removal of SG. This

1 Introduction The removal and management of large components like steam generators

(SGs) is one of the major tasks during decommissioning and dismantling of a PWR nuclear power

plant. In case the decommissioning schedule is a key parameter to meet the budget, which it usually

is, transportation and treatment of large components to an external facility (Rip & Ship) is

­beneficial. This paper gives a short introduction of existing available technology for Rip & Ship

combined with advanced treatment.

example does not necessarily fits for all

plant geometries and conditions but

can be used to outline a specific

sce nario for NPPs without the practical

possibility to remove SGs as one

piece.

For the SG decommissioning EDF

chose the following scenario:

p Cutting of each steam generator in

2 pieces (i.e. separation of steam

dome) inside the reactor building

as for the previously replaced

steam generators).

p It generates 6 pieces of less than

200 t of which three are only

slightly contaminated and three

contains more than 99 % of the

component radioactivity

p Handling out of reactor building by

a specific handling platform or

polar crane,

p Storage in a dedicated building on

site (same than the one used for

steam generator replacement),

p Treatment off site in a EDF Group

facility aiming for recycling of

material and minimization of the

waste volume

Prior to the removal of the SGs from

the reactor hall, the previously

replaced SGs will have to be eva cuated

from the storage building and sent off

site for treatment. This is scheduled to

take place near term.

The definition and implementation

of optimized waste routes, before

starting the dismantling, improves the

project schedule, cost effectiveness

and its success. This solution contributes

to the:

p reduction of interfaces and risks

in decommissioning projects by

inte gration across the value chain,

p reduction of waste management &

disposal costs,

p optimization of scarce radioactive

disposal capacity.

For the rest of the French fleet, it’s not

necessary to remove the steam dome

from the steam generators to bring

them out of the reactor building. Onepiece

removal is the preference.

These examples show that the rip

& ship concept can be used for intact

SGs or be combined with a partial

segmentation of the SG in order to

ease transportation within building

structures.

b Transportation of steam

generators

Any treatment outside the reactor

building will likely, independent of

distance, require a transport qualification

& licensing as well as the

involvement of a heavy lifting and

transport company. The additional

practical efforts for the licensee

| Steam generator replacement upper part handling in the reactor building.

| SG lower part handling in the reactor building.

Decommissioning and Waste Management

Steam Generator Rip and Ship – a Valuable Contribution to Decommissioning and Dismantling of Nuclear Power Plants ı Heiko Herbell, Arne Larsson, Gregor Krause and Véronique Bouilly


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 1.

Ringhals SG entering the treatment facility (Cyclife Sweden AB).

to remove the SGs from the site

compared to an on-site waste

treatment facility are fairly limited.

However, the movement of large

contaminated components over long,

trans-boundary distances will require

pre-studies and qualifications, special

licenses, stakeholder interactions and

practical transport arrangements.

Cyclife has, both in practice and

in studies, demonstrated that the by

road or water for large contami nated

components within Europe are open

or can be opened when necessary.

The same is expected also for other

parts of the world, based on studies

conducted.

In total, hundreds of large, contaminated

components have been

shipped to the Swedish facility for

treatment. More than 20 of them have

exceeded 300 tonnes in weight.

3 Description of treatment

technology

Already in the early phases of the

planning for a SG treatment project, it

is important to understand the radionuclide

inventory and distribution.

Whether the SG has been chemically

decontaminated as a part of a full

­system decontamination, a specific

object decontamination or only has

decayed down to reasonable levels

for handling and off-site treatment

has a major impact on the optimisation

of the treatment.

A chemical decontamination prior

to disconnection from the primary

­circuit significantly decreases the

collective dose exposure during dismantling,

handling and treatment as

well as for the return of the treatment

residues (as most of the inventory was

removed already at the NPP).

Also detailed drawings, engineering

details, metallurgic data,

operational history records that are

available are of great importance for

the handling, segmentation and decontamination

activities and recycling.

A key parameter is data on the level

for contamination, if any at all of any

significance, on the secondary side.

Other, in most cases even more

important parameters related to the

operational history, are the data

regarding the tubing: number and

location of inserted sleeves and plugs.

Also, how the sleeves and plugs are

designed and have been fixed to the

tubes are of prime interest.

Cyclife Sweden, formerly Studsvik

Waste Treatment, has performed

several SG treatment projects, in total

treated 13 full size SGs.

p Nine SGs of Westinghouse design

originating from the Ringhals NPP

in Sweden, approximately 300 Mg

each.

p Four SGs of Siemens design

originating from the Stade NPP in

Germany, approximately 165 Mg

each.

The conducted projects demonstrate

that SGs can be authorized for

shipment and shipped as one piece,

to an external treatment plant.

The treatment process for recycling

of full-size SG’s was started up in

2005 and has been enhanced stepwise

over the years. The driving force for

treatment of those components was

primarily to reduce the waste volume

for final disposal, although there are

| Fig. 2.

Cross section of 300 Mg SG of Westinghouse design.

other benefits such as reduction of

on-site/project related risks as well

the elimination of a time critical

project during on-going decommissioning.

SGs received at the treatment

facility are handled, further characterized,

decontaminated and size

reduced. The entire secondary

side material do not usually require

any special decontamination (but

careful separation of non-contaminated

parts) as the contamination on

the secondary side should be very

limited, if any. The parts exposed to

the pri mary circuit, i.e. the water

chamber, parts of the tube plate and

the tubes, need special decontamination

or material segregation to be

candidates for clearance after treatment,

if possible at all. The main

challenge is to make the tubes candidate

for clearance. It is currently

not within reach, in practice, for

Inconel-600 tubes due to the intergranular

cracks with embedded

activity the treatment.

Depending on the cost for waste

storage and disposal, as well as the

availability of repository volume, the

plant owners may have different

preferences. If the disposal cost is low,

it may be sufficient to treat the entire

secondary side for clearance and

dispose the tubes, the tube plate and

the water chamber as contaminated

material. On the other hand, if the

disposal cost is high, there is likely an

interest to minimise the amount of

waste also at higher treatment cost. In

this case the preference may be that

only the tubes and the residues from

the treatment should be returned for

disposal.

The secondary side material may, if

con sidered free from contamination

of any significance, be decontaminated

and undergo clearance procedure

for recycling back to industry

either by implementing melting or by

direct clearance. Both alternatives are

available, the Cyclife preferred concept

is to combine direct clearance

and clearance after melting. In most

DECOMMISSIONING AND WASTE MANAGEMENT 407

Decommissioning and Waste Management

Steam Generator Rip and Ship – a Valuable Contribution to Decommissioning and Dismantling of Nuclear Power Plants ı Heiko Herbell, Arne Larsson, Gregor Krause and Véronique Bouilly


atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 408

| Fig. 3.

Treatment of Ringhals SG in the Cyclife Swedish facility.

cases a vast majority of the secondary

side tonnage goes to direct clearance.

By applying melting on all the metal,

the results are more predictable but at

a somewhat higher cost.

The primary side parts will likely

have to be mechanically decontaminated

by a physical removal of the

surface layer to be candidates for

clearance after treatment. This has

to be optimized with the customers

needs. Based on the past experiences,

the decontamination methods have

been enhanced incorporating new

technologies and new equipment.

The tube plates need a careful

analysis prior to selection of the

treatment approach. Multiple technologies

are applied by the SG

­manufacturers to fix the tubes in the

tube plate and to secure that no

leakage will occur. Of these reasons,

the treatment of the tube plates will

have to be tailored. Depending on

the complexity the efforts will differ

to make the tube plate a candidate

for clearance.

By application of chemical decontamination

prior to SG removal,

the dose rates for handling and

­treatment can be reduced significantly

and the management of the residual

waste simplified. The residual waste

contains most of the radioactive

­inventory but in a significantly lower

volume.

The SG recycling rate is expected

to be in the order of

60-80 % depending on the degree

of treatment aiming for clearance,

i.e. 20-40 % of the original tonnage

will have to be disposed as radioactive

waste.

The residues from the treatment

will have to be returned as radioactive

waste for disposal. Figure 4 shows

how the secondary waste from treatment

will look like.

Likely all tubes will have to be

returned as waste. Depending on

customer preferences, the tube

material can be compacted, chopped

and flattened or melted as illustrated

in Figure 5.

The treatment time for one,

300 Mg steam generator, using the

latest technology is estimated to

3-5 months (depending on the

selected scope of services and the

­specific SG properties).

In a volume perspective, which is

the most important as most waste

repositories charge per volume

­disposed, only 7-20 % of the original

volume for a full size SG will have to

be disposed. The percentage depends

both on the degree of treatment for

clearance but also on the efforts to

­reduce the final volume.

4 Conclusion

Examples from past projects such

as Ringhals NPP or Stade NPP but

also from ongoing projects such as

Fessenheim underline flexibility of

the rip & ship concept in order

to overcome challenges during

on-site treatment as eq logictic for

>1000 tonnes of SG material during

treatment in the narrow reactor

building.

The Fessenheim projects also

highlights the issue with previously

replaced SGs stored on site. Removal

of them from the storage is an

important decommissioning preparation

activity to be done well

ahead of the dismantling start.

Transportation of large contaminated

components over long, transboundary

distances require prestudies

and qualifications, special

licenses, stakeholder interactions and

practical packaging arrangements.

Besides these efforts, offsite treatment

of large components enable

­significant advantages such as:

p Less waste handling and treatment

on site.

p Potential for a significant reduction

of the decommissioning schedule.

p Risk mitigation by transfer of work

to specialists with proven pro cesses

and experienced workforce in facilities

that have been designed

and built for the purpose.

p A significantly lower volume of

waste for disposal. In addition, this

option has the benefit of recycling.

p A predictable result and a fixed

cost.

Abbreviations

EDF – Électricité de France

NPP – Nuclear Power Plant

PWR – Pressurized Water Reactor

SG – Steam Generator

SGR – Steam Generator Replacement

| Fig. 4.

Secondary waste from treatment.

Authors

Dr. Heiko Herbell

heiko.herbell@framatome.com

Framatome GmbH

Paul Gossen Str. 100

91058 Erlangen, Germany

Arne Larsson

Gregor Krause

Cyclife Sweden AB

Box 610

SE-61110 Nyköping, Sweden

Véronique Bouilly

Cyclife Engineering

196, Avenue Thiers

69006 Lyon, France

| Fig. 5.

Alternatives for the tube material not subject to clearance.

Decommissioning and Waste Management

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

Casks and Cask Stacks in Interim Storage

Facilities under Earthquake Loads

Nina Wieczorek

Interim storage of casks In Germany, there are interim storage facilities built directly Planned entry for

next to nuclear ­power plants as well as three central interim storage facilities sited in Ahaus,

­Gorleben and near Lubmin. These interim storage facilities differ in the classification of casks

stored, differentiating between radioactive waste with negligible heat generation and dry cask

storage of spent fuel and heat-generating waste. Depending on the type of waste, different types of

casks are stored. Besides containers of various types, cylindrical casks like e.g. shipping and storage casks of type

­CASTOR® V/19, V/52, MTR3 and cast iron casks of type MOSAIK® are used. The containers and casks are stored

separately as well as stacked, whereat common ISO edges are used for containers. The casks are stacked directly or with

stacking aids. This paper deals with cylindrical casks stored separately as well as stacked.

Guidelines and standards

ESK guidelines

The ESK guidelines (ESK: Nuclear

Waste Management Commission)

contain protection goals with which

the storage of radioactive waste has to

comply. The requirements resulting

from these are the avoidance of

any unnecessary radiation exposure

or contamination of man and the

environment and the minimization of

any radiation exposure or contamination

of man and the environment.

Depending on the type of radioactive

waste, the ESK established

guidelines to ensure the compliance

with the protection goals, see [1, 2],

whereat external hazards like the

“safety shutdown earthquake” (SSE)

focussed in this paper is addressed as

well. Ensuring the compliance with

the protection goals for the load case

SSE is feasible by proof of stability of

the containers and casks.

KTA series 2201

The KTA series consists of six parts

and deals with the design of nuclear

power plants against seismic events,

whereat its scope of application is

extended by the ESK guidelines

[1, 2]. KTA 2201.1 [3] contains

general requirements concerning the

design basis earthquake (DBE) and

the ­verification. KTA 2201.2 [4] deals

with the subsoil. While KTA 2201.3

[5] contains detailed requirements

concerning the verification of civil

structures, KTA 2201.4 [6] contains

detailed requirements concerning the

verification of components.

All six parts had been amended

successively between 2011 to 2015.

Especially with the new version of

KTA 2201.1 [3], the requirements

regarding the DBE were augmented

compared to the previous version.

Thus, the DBE has to be specified with

an intensity of at least VI and a probability

of exceedance of at least 10 -5 /a.

External hazard – “safety

shutdown earthquake”

Concerning the existing interim

storage facilities, seismological expert's

reports define a site-specific

DBE and seismo-engineering parameters

(e.g. , intensity, probability of

exceedance and strong motion

­duration) on the basis of KTA 2201.1

[3] as well as soil parameters (e.g.

shear modulus, density, Poisson's

­ratio, damping ratio). According to

KTA 2201.1 [3], the DBE is defined

as a free field response spectrum,

meaning a ground acceleration

response spectrum for a reference

horizon in the subsoil, where the

­oscillation properties are not influenced

by building structures. No

soil-structure interaction is considered.

Generally, the reference

horizon is equal to the ground level or

the geological layer boundary of a

­sufficiently stiff ground layer. The

scatter band of the soil profile as well

as uncertainties are covered within

the computations according to KTA

2201.1 [3]. The DBE results from

smoothed, broadened and enveloped

spectra and is defined for the horizontal

resultant as well as the

horizontal and vertical component.

Figure 1 (a) contains exemplarily free

field ­response spectra for a damping

ratio of 5 % for an arbitrary site in

Germany. The peak ground acceleration

(pga) for the horizontally resultant

direction a hr is 0.26 g, for

the vertical component a v it is 50 % of

a hr (0.13 g).

Floor response spectra applied

as design spectra

KTA 2201.4 [6] defines the design

spectrum to be enveloping, widened

DECOMMISSIONING AND WASTE MANAGEMENT 409

a

| Fig. 1.

Free field response spectra for a site in Germany (a) and FEM model of an interim storage facility (b).

b

Decommissioning and Waste Management

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 410

and smoothed. One differentiates

between ground acceleration spectrum

as primary spectrum, building

response spectrum as secondary

spectrum, and component response

spectrum as tertiary spectrum.

a

Generally, based on the sitespecific

DBE, floor response spectra

(FRS) for the base plate and the

crane runway are computed on a FEM

model of the building of the interim

storage facility, exemplarily shown in

Figure 1 (b). In order to consider

soil-structure interaction, besides the

structural building the soil is contained

in the numerical model as well.

According to KTA 2201.1 [3] and

2201.3 [5], diverse variations of the

| Fig. 2.

Floor response spectra of the base plate of an interim storage facility for a damping ratio of 4 %; x-direction (a), y- direction (b) und z- direction (c); comparison of the required response

spectra and the response spectra transformed back from the compatible time histories (TH, two per direction).

b

c

a

| Fig. 3.

Comparison of time histories from structural analyses (SA; blue line) with time histories compatible with the required response spectra (orange line); exemplarily plotted for the model

„ maximum shear modulus (stiff soil) / maximum load“; x-direction (a), y- direction (b) and z- direction (c).

b

c

a

b

| Fig. 4.

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

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).

c

d

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Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


atw Vol. 65 (2020) | Issue 8/9 ı August/September

reference model are investigated considering

the scatter band of the soil

profile and the various loading conditions.

Concerning the soil, the

average shear modulus (G mid ) is

varied between a lower (G min ) and

upper (G max ) bound. The various

loading conditions are considered by

additional masses ranging from a

lower (M min ) and upper bound (M max )

as well.

Response spectra are computed

for the various models, whereat the

excitation is applied as time histories

compatible with the DBE according

to KTA 2201.1 [3] and 2201.3 [5].

­Afterwards, FRS are identified as

smoothed, broadened and enveloping

design response spectra (DRS).

Figure 2 contains DRS with a

damping ratio of 4 % for the base

plate of an interim storage facility.

Furthermore, Figure 2 contains

response spectra transformed back

from time histories, which are

compatible with the DRS according to

KTA 2201.1 [3] and 2201.3 [5],

and taken as a basis for component

verification.

Time histories compatible

with required/design response

spectrum

If quasi-static methods are not

­sufficient, time history analyses are

performed in order to verify the

­stability of the cask/cask stack for the

load case SSE with the help of a

FEM model of the cask/cask stack.

As ­mentioned above, time histories

compatible with the required response

spectrum (RRS) are generated

complying with the requirements of

KTA 220.1 [3] and used as excitation

for the FEM model, see Figure 2.

Time histories taken directly

from structural analysis

Besides the approach of time histories

compatible with the RRS, KTA 2201.1

[3] allows an alternative approach by

applying time histories taken directly

from the structural analyses of the

building. Here, it has to be considered

that the scatter band of the structural

analyses is covered. That is, analogue

to the structural analyses, the regard

of variation of soil stiffness (G min ,

G max ) and loading conditions (M min ,

M max ). Since the time histories taken

from the structural analyses do not

contain the conservatives from the

RRS, this approach allows the application

of lower excitations to the FEM

model, but results in an increase of

computational cost, see Figure 3.

FEM model

Modelling and computations are

performed with the commercial

FEM program ANSYS including the

­LS-DYNA solver. In order to represent

relative displacements of the cask/

cask stack on the base plate, including

tilting, trundling and sliding, besides

the cask/cask stack, the numerical

model also contains a section of the

base plate. Between the (bottom) cask

and the base plate and between the

casks in a stack contact definitions are

applied to the FEM model. For the

contact pairs the friction coefficient

are varied between a lower and an upper

bound depending on the material

(µ cask-base, min = 0.2 und µ cask-base, max =

0.6, µ cask-cask/stacking aid, min = 0.1 and

µ cask-cask/stacking aid, max = 0.3). The

lower bound allows a conservative

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Decommissioning and Waste Management

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 412

a b c

| Fig. 5.

Displacements of a cask of type CASTOR® V/52 relative to the base plate applying various friction coefficients for the contact pair cask/base plate; excitation using time histories

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

a friction coefficient of µ = 0,2.

a

| Fig. 6.

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

two time history combinations compatible with the RRS; friction coefficient of µ = 0.6 for the contact pair cask/base plate and µ = 0.2 for the contact pair cask/cask (a); friction coefficient of

µ = 0.2 for the contact pair cask/base plate and µ = 0.1 for the contact pair cask/cask (b).

b

approach concerning sliding, while

the upper bound allows a conser vative

approach concerning tilting and

trundling. The excitation within the

nonlinear computations is applied to

the base plate section. Figure 4 contains

FEM models of various casks and

cask stacks of the types CASTOR®

V/52, MTR3 and MOSAIK®, with and

without stacking aids.

Nonlinear time history analyses

For nonlinear time history analyses,

loading conditions of the cask and

within the cask stack are investigated.

Furthermore, for cask stacks without

stacking aids imperfections due to

handling accuracy need to be considered.

Depending on the computed

­displacements of the cask/cask stack

relative to the base plate section,

minimum distances can be necessary

in order to avoid collisions of neighbouring

casks for the load case SSE.

Results – time histories

compatible with design spectra

Figure 5 contains displacements of a

single cask of type CASTOR® V/52

relative to the base plate section with

an average (µ = 0.4) and a minimum

(µ = 0.2) friction coefficient for

the contact pair cask/base plate.

­Applying an average friction coefficient

leads to trundling movements

with a maximum horizontal displacement

of ­approx. 70 mm for the cover

plate (Figure 5 (a)) and a maximum

vertical displacement of approx.

25 mm for the bottom plate with a

phase shift visible by the displacements

of two ortho gonal result

nodes on the bottom ( Figure 5 (b)).

Applying the lower bound of the

­friction ­coefficient leads to a sliding

movement with a com paratively little

maximum horizontal displacement of

approx. 6 mm.

One can see clearly larger horizontal

displacements in Figure 6,

whereat these displacements come

from a cask stack of four casks of type

MOSAIK® without stacking aids.

Exemplarily, two different time

history combinations are contrasted,

respectively, for each limit value

investigation of the friction coefficients.

While the influence of the

time history combination on the

horizontal stack displacement is

quiet low concerning the lower

bounds of the friction coefficients (TH

comb. 1 approx. 33 mm vs TH comb. 2

approx. 30 mm), concerning the ­upper

bound it is significant (TH comb. 1

­approx. 250 mm vs TH comb. 2 ­approx.

430 mm). Like the single cask

shown in Figure 5, the whole stack

trundles. As one can see in ­Figure

7 (c), the relative displacements

for the horizontally resultant direction

of two ortho gonal nodes, respectively,

are in-phase. The cask no. 4 (top

cask) slides less than 1 mm on the

cask no. 3.

Decommissioning and Waste Management

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


atw Vol. 65 (2020) | Issue 8/9 ı August/September

a

b

| Fig. 7.

Cask stack of four casks of type MOSAIK®; excitation using time histories compatible with the RRS; friction coefficient of µ = 0.6 for the contact pair cask/base plate and friction coefficient

of µ = 0.3 for 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

relative to the cask no. 3 (c).

c

DECOMMISSIONING AND WASTE MANAGEMENT 413

a

b

c

| Fig. 8.

Cask stack of four casks of type MOSAIK®; excitation using time histories taken directly from the structural analyses; friction coefficient of µ = 0.6 for the contact pair cask/base plate and

friction coefficient of µ = 0.3 for 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 relative to the cask no. 3 (c).

Results – approach of times

histories taken directly

from structural analyses

Since the displacements of the

cask stack of four casks of type

­MOSAIK® computed applying time

histories compatible with the DRS

and the upper bounds of the friction

coefficients are very large, in further

computations time histories taken

directly from the structural analyses

of the building are applied to the

FEM model complying with the

­requirements of KTA 2201.1 [3] and

2201.4 [6] as well. Figure 8 gives a

similar image like Figure 7, the cask

stack trundles, but the displacements

are con siderably lower. For the top

of the cask stack, the maximum

horizontal displacement is approx.

220 mm. Even though the computational

cost increases, the maximum

displacements and thereby

the necessary minimum distance

of two neigh bouring casks/cask

stacks can be reduced in order to

avoid a collision.

Conclusions

In order to proof stability of a cask/

cask stack in interim storage facilities

for the load case “safety shutdown

earthquake” tilting, trundling and

sliding of the cask/cask stack need

to be investigated. With the help of

FEM, various computations on a

suitable numerical model containing

the cask, stacking aids if utilised

and a section of the base plate can

be carried out applying friction

­coefficients for both contact partners,

cask/base plate and cask/cask, within

a certain range and varying the mass

distribution of the cask stack. In

doing so, the numerical results show

that not the sliding of the cask/

cask stack is significant but the

­trundling leading to a predefinition

of a minimum distance between

neighbouring casks/cask stacks.

Literature

[1] ESK – Nuclear Waste Management Commission:

Recommendation of the Nuclear Waste Management

Commission: Guidelines for dry cask storage of spent fuel and

heat- generating waste; Revised version of 10.06.2013

[2] ESK – Nuclear Waste Management Commission:

Recommendation of the Nuclear Waste Management

Commission: Guidelines for the storage of radioactive waste

with negligible heat generation; Revised version of

10.06.2013

[3] KTA Rule 2201.1 (11-2011): Design of Nuclear Power Plants

against Seismic Events. Part 1: Principles; Version November

2011

[4] KTA Rule 2201.2 (11-2012): Design of Nuclear Power Plants

against Seismic Events. Part 2: Subsoil; Version November

2012

[5] KTA Rule 2201.3 (11-2013): Design of Nuclear Power Plants

against Seismic Events. Part 3: Design of structural

components (civil structures); Version November 2013

[6] KTA Rule 2201.4 (11-2012): Design of Nuclear Power Plants

against Seismic Events. Part 4: Components; Version

November 2012

Author

Dr.-Ing. Nina Wieczorek

n.wieczorek@gmx.de

Technical Advisor

“Structural Dynamics and

Earthquake Engineering”

TÜV NORD EnSys GmbH & Co. KG

Bahnstr. 31

22525 Hamburg, Germany

Decommissioning and Waste Management

Casks and Cask Stacks in Interim Storage Facilities under Earthquake Loads ı Nina Wieczorek


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DECOMMISSIONING AND WASTE MANAGEMENT 414

Radioactivity Calculation of the

Concrete Shielding of the Petten LFR

and the Dodewaard BWR

Lino Salamon, Perry Young and Lojze Gačnik

Planned entry for

1 Introduction After nearly 30 years of operation the Nuclear Power Plant Dodewaard

(KCD) was ­permanently shut-down in March 1997. The KCD was a General Electric design boiling

water reactor (BWR) with thermal capacity 183 MWt, and an electrical capacity of 54 MWe. After

shutdown data was collected on the radioactive inventory in preparation for Safe Enclosure and future

dismantlement. One of the larger reactor components undergoing significant neutron ­activation

is the concrete biological shield (BS). Gamma spectroscopy measurements were made with samples of the concrete BS to

know the mass specific activity of the nuclides at different positions in the BS and to estimate the amount of radioactive

waste that would need to be sent to the Dutch disposal company COVRA at different dismantlement dates [1].

The Low Flux Reactor (LFR) at the

NRG Petten site was permanently

shut-down in November 2010 and

decommissioning was completed

­February 2019. The LFR is a JASON

variant of the Argonaut class reactor

supplied by Hawker Siddeley, with

a thermal capacity of 30 kWt.

Similarly to KCD, gamma spectroscopy

measure ments were made to

assess the radioactivity of the LFR’s

concrete shielding [2].

The aim of this work was to predict

the specific BS activities of KCD and

LFR via activation calculations and to

determine the delimitation between

regions of free-release material and

radioactive material that must be sent

to COVRA. The amount of radioactive

waste estimated from the measurements

was compared to the calculated

values. The ability to predict the

extent of radioactivity in the concrete

via calculation should permit better

planning of a nuclear facility’s decommissioning

and could possibly reduce

costs.

2 Calculation Methodology

The calculation of the activation

inventory was done in two steps. First,

the neutron fluxes and their spectra

were calculated in the BS using the

Monte Carlo radiation-transport code

MCNP6.2 [3]. In the second step, the

flux tallies were used in a FISPACT-II

[4] activation calculation. Based on

the free release activity limits [5] the

concretes were then categorized as

radioactive or free-release and

hence the amount of radioactive

waste were determined for different

dismantlement dates. Below in section

2.1 is detailed the methodology

for Dodewaard (KCD). The methodology

for the LFR can be found in Ref.

[2]; they are broadly similar, although

the MCNP geometric model of the LFR

is of far greater detail, explicitly modelling

almost all of the features in the

reactor and shielding.

2.1 KCD Geometric Model

The MCNP geometry of Dodewaard

from [1] was used to calculate the

neutron fluxes in BS, which is a

circular wall around the containment

wall. Two separated geometrical

models were available: a model of

the lower part extending between the

bottom of the concrete floor up to the

top of the core and a model of the

upper part extending from slightly

below the reactor core up to the top of

the biological shield above the reactor

vessel. The model of the Dodewaard’s

lower part is only a quarter of

the whole geometry, with reflective

surfaces at 0 and 90. The lower

and upper model are visualized in

Figure 1. Both models were updated

to ENDF/BVIII.0 nuclear data libraries

for the neutron transport using

temperatures broadly equivalent to

those during operation. These models

are very simple and lack much definition.

To wit, the nuclear fuel region is

a smeared homogenous zone, mixing:

fuel, cladding and water. For the

indicative purposes of this work, they

should be adequate.

| Fig. 1.

The MCNP models of Dodewaard’s lower (left) and upper part (right).

2.1 Material Composition

There are two types of concrete in the

BS of Dodewaard: heavy concrete with

density ρ ≈ 3.5 g/cm 3 (most of the inner

shielding and some of the BS) and

light concrete with density ρ ≈ 2.3 g/

cm 3 (most of the outer part of BS).

The material composition of the

heavy and light concrete was updated

Decommissioning and Waste Management

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

with the composition obtained from

the X-Ray Fluorescence (XRF),

­ICP-AES, ICP-MS, Loss of Ignition

(LOI), Total Carbon (TC) and Total Organic

Carbon (TOC) measurements

carried out by TCKI and TNO-Utrecht.

Measurements were performed using

10 free-release heavy concrete and

light concrete samples obtained from

Dodewaard. These had been part of

the previous core extractions [1], and

were being held on site within a

cabinet.

In the calculations, the XRF

measurements were used for the mass

fractions of the major constituent

elements and ICP for the trace elements

(Eu, Mn and Co). The LOI, TC and

TOC were used together to deduce the

moderating media (H, C and O). Based

on the experience comparing the LFR

gamma measurements to calculations

in [2], two cases were defined: a Best

Estimate case (BE) and a Conservative

case (CON). In these cases, per

concrete type, the mass fractions used

for the elements were as follows:

BE:

p MCNP: maximum values of the

measurements for hydrogen and

carbon, mean values of the

measure ments for the rest

p FISPACT: median values of the

measurements

CON:

p MCNP: mean values of the

measurements

p FISPACT: maximum values of the

measurements

2.2 Variance Reduction

Due to the deep penetration nature of

the problem, some kind of variance

reduction technique had to be applied

to efficiently propagate neutrons in

the concretes. For this reason, the

­ADVANTG code [6] was used to create

space and energy dependent mesh

based weight windows using threedimensional

(3D) discrete ordinates

(SN) solutions of the direct and

adjoint deterministic transport equations

that are calculated by the

­Denovo package. An MCNP geometry

of the upper part overlaid with the

weight-window mesh is presented in

Figure 2.

2.3 MCNP Flux Calculations

MCNP calculations were done in

multiple steps. First, a k-code criticality

calculation was performed to

generate a volumetric fission source

(SSW) in the nuclear fuel. This source,

together with the weight-windows

produced by ADVANTG, was subsequently

used in a fixed source

| Fig. 2.

Upper part of the Dodewaard geometry model

overlaid with a weight-windows spatial mesh.

calculation. The per-source-particle

neutron flux (neutrons/cm 2 ) in 172

energy groups was tallied in small

voxels over the concrete regions using

Cartesian and cylindrical mesh geometry.

2.4 FISPACT Activation

Calculations

Using the obtained flux tallies, the specific

activities (in Bq/g) of concrete

were estimated with FISPACT. In ­order

to get the actual neutron fluxes, the

calculated tallies first need to be scaled

with a scaling factor PNF as ­defined in

[7]. For this scaling, Q= 193 MeV, ¯v f =

2.565 and k eff = 0.98902 (the last two

calculated in MCNP), were used.

­Finally, the scaled fluxes were multiplied

in each time period by the power

history of KCD [8] in that time

period. Specific ­activities in FISPACT

were ­calculated with the 709-energy

group activation library ENDF/BVIII.0

( except for Eu-151 and Eu-153 that

used the ­EAF-2010 library, following

Eu C/M results for the different

datasets found in [2]).

Release

date

Ref.

[1]

2.5 Radioactive and

Free-Release Zones

Radioactivity of the material is

­defined according to standards

­defined in [5]. Each nuclide i has its

own specific activity limit, I i , that it is

not allowed to exceed in order to

qualify as free release. If a sample

has multiple active nuclides, then an

I 0 ratio is defined as:

where a i is the specific activity of

nuclide i. For I 0 < 1 the sample is

considered to be free release and for

I 0 ≥ 1 the sample is considered as

radioactive. The I 0 criteria was used to

map the radioactive and free-release

regions of Dodewaard KCD. Finally,

the total mass of concrete voxels

with I 0 ≥ 1 was used to estimate the

radioactive waste at different dismantlement

dates.

3 Results

3.1 Dodewaard

Figure 3 presents the radioactive

(in red) and free-release (in green)

regions at different axial positions of

the Dodewaard model (indicated in

Figure 1), approximately 43 years

after the KCD shutdown. Colored

areas represent the heavy and light

concrete of the biological shield, but

also concretes inside the containment

wall. The largest radial spread of

radioactivity in BS is at the heights

surrounding the active core (at

z ∼ 2300 cm) and it (at z ∼ 2800 cm),

where the BS is mainly made of light

concrete.

Figure 4 compares the radioactivity

of the BS at z = 2750 cm for

two time steps: a reactor shut-down

(26/03/1997) and the planned end of

Safe Enclosure (01/01/2040). The

radioactivity in the radial direction is

significantly decreased after 43 years

of decay under Safe-Enclosure. However,

according to calculations the

concretes inside the containment wall

and the inner parts of BS are still above

the specific activity limit (I 0 ≥ 1).

Radioactive waste [Mg]

CON

A2 Limits [5]

BE

A2 Limits [5]

BE

KCD Limits [1]

26/03/1997 270 1208 929 939

01/01/2010 230 510 322 467

01/01/2020 200 386 223 341

01/01/2030 180 272 157 244

01/01/2040 140 189 123 193

| Tab. 1.

Estimated radioactive waste for KCD at different dismantlement dates.

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 416

| Fig. 3.

Radioactive (in red) and free-release (in green) regions of BS at different axial positions of the Dodewaard model.

| Fig. 4.

Radioactivity of BS at z = 2750 cm for two time steps. Left: reactor shut-down (26/03/1997).

Right: the end of Safe Enclosure (01/01/2040).

Table 1 gives an overview about

the estimated radioactive waste

masses in accordance with the release

date. The second column in Table 1

reports the values from [1], which are

based on the gamma spectrometric

measurements of samples taken from

the BS at different heights and depths.

Radially symmetric distribution was

assumed, meaning that one sample

was representative for specific activity

at a certain height and depth in the BS.

The calculated quantities are 4x

higher at the end of life of the reactor,

but at future dates (2010, 2020,..)

they are between 2x-1x the values in

[1]. The best agreement between the

calculated values and estimations in

[1] are at year 2040, with differences

of 10 % – 40 %, depending on the

case.

One of the reasons for the difference

between [1] and the calculated

values could be due to a number

of short-lived nuclides, e.g. Mn-54,

Zn-65 and Cs-134, which were not

considered in Ref [1]. In [1] it is also

not clear what clearance levels are

used to determine the rad-waste

quantities. However, there are several

limits mentioned that could be inferred

to be the clearance levels

but several of these vary slightly

from [5]. The principal results given

here used the ADR A2 limits [5], but

additional BE results used the inferred

KCD limits (Depository Clearance

Levels) [1].

Furthermore, several assumptions

and simplifications on the layout of

facility were made in the original

MCNP model. For some of the regions

in BS there is not enough information

in [1] to assign the right concrete type

(heavy or light) in the MCNP model.

Ref. [1] also assumes radial distribution

of activities in BS, but there are

regions where at the same radius and

height the concrete type varies. Even

with the same type of concrete at

certain height and radius the activity

can vary. This is demonstrated in

Figure 5, where the area just above

the chimney (at z ∼ 2600 cm, look at

Figure 1) is more radioactive than

other parts at the same radius. In

addition to that, the results on LFR

reported in Ref. [2], which used the

same calculation procedure as this

work, show that calculations can

overestimate the measured activities

up to a factor of 4.

3.2 LFR

A similar calculation procedure for the

BS specific activity estimation was

­applied on LFR. Specific activities in

FISPACT were calculated with the

172-energy group activation library

EAF-2010. Details of the procedure

can be found in [2]. The 709-energy

group ENDF/B VIII.0 was not used

here due to calculation time constraints

(i.e. large amount of voxels

was required to capture the asymmetrical

layout of the facility).

However, the test calculations on

Dodewaard case indicate that the

waste estimates with ENDF/B VIII.0

would be somewhat lower.

Figure 6 compares the radioactivity

map at z = 81 cm (about

50 cm above the active core) for the

CON and BE cases. Purple and blue

areas designate the regions with

poor statistics, i.e. the regions with

flux ­uncertainty more than 10 %

and zero flux regions, respectively.

In CON case the radioactivity is

spread over a larger region around

the reactor core.

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Radioactivity Calculation of the Concrete Shielding of the Petten LFR and the Dodewaard BWR ı Lino Salamon, Perry Young and Lojze Gačnik


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 5.

Radioactivity map (Cartesian mesh)

just above the chimney.

disposed

at COVRA

In Table 2 the mass of radioactive

concretes of different type disposed at

COVRA between years 2017-2019 is

given along with those BE and CON

calculated as radioactive as of January

2014. The radioactive foundation

­concrete disposed at COVRA seems to

lie between the two calculated values.

The disposed barite concrete is 2x-4x

higher than the calculated values. The

barite blocks are large cubes with

sides of about 1 m. One possible

reason for the discrepancy is due to

the voxel nature of the calculation

which would cut up the barite cubes.

It is possible that during decommissioning

if one portion of barite was

found to be radioactive the whole

cube was considered radioactive,

which would lead to greater amounts.

The amount of radioactive scrap

concrete is 5x higher comparing to

calculations. One possible reason for

difference here is that the COVRA

quantity includes the ‘irradiation

wagon’, i.e. moveable shielding

blocks, whereas the calculation does

not consider this structure. Similarly

to the barite, the scrap concrete is

located in 4 large top-shield

structures. These would likely not

be cut up, but rather disposed of

as a whole.

Conclusions

In this work the radioactivity of

BS in two nuclear facilities (i.e. the

KCD and LFR) was mapped out.

The spread of calculated radioactivity

can give us insight of more important

Radioactive waste [Mg]

CON

calc Jan. 2014

regions when planning the Safe

Enclosure and dismantling, and can

save cost and time for the activation

measurements. Based on the

mapped radioactive regions, the

quantities of radioactive waste were

estimated.

Overall, the results from BE calculations

match better the radioactive

waste estimated from the activity

measurements. However, even the

CON case at the planned dismantlement

date differs by less than a factor

of 2 and 3 for Dodewaard and LFR,

respectively. These reasonable results

demonstrate that conservative quantities

of radioactive waste at different

dates could be estimated beforehand,

using the calculation procedure from

this work.

References

[1] GKN report 99-001/PID/R: “Activation measurements of the

Biological Shield”, Ministerie van Infrastructuur en Milieu.

(2017). Europese overeenkomst voor het internationale

vervoer van gevaarlijke goederen over de weg (ADR)

[2] P. Young, et al. “Characterizing the Radioactivity of the

Concrete Shielding during Decommissioning of the LFR”,

Proc. Int. Conf. Nuclear Energy for New Europe 2019,

Portorož, Slovenia

[3] C.J. Werner, et al., “MCNP6.2 Release Notes”,

LA-UR-18-20808, 2018

BE

calc Jan. 2014

Barite 22.7 9.3 4.78

Scrap 9.9 1.94 1.73

Foundation 2.6 6.77 0

| Tab. 2.

Estimated radioactive waste for LFR.

| Fig. 6.

Comparison of the radioactivity spread in BS (in XY direction) 50 cm above the active core for CON (left) and

BE (right) case of LFR.

[4] M. Fleming, et al., The FISPACT-II User Manual,

UKAEA-R(18)001, January 2018

[5] Europese overeenkomst voor het internationale vervoer van

gevaarlijke goederen over de weg (ADR) Bijlagen, 2017

[6] S. W. Mosher, et al., “ADVANTG – An Automated Variance

Reduction Parameter Generator”, ORNL/TM-2013/416 Rev. 1,

August 2015

[7] L. Snoj, M. Ravnik, “Calculation of Power Density with MCNP in

TRIGA reactor”, Proc. Int. Conf. Nuclear Energy for New Europe

2006, Portorož, Slovenia

[8] https://pris.iaea.org/PRIS/CountryStatistics/

ReactorDetails.aspx?current=422, accessed 18/12/2019

Authors

Lino Salamon

salamon@nrg.eu

Perry Young

Lojze Gačnik

Nuclear Research and Consultancy

Group NRG

Westerduinweg 3

1755 Le Petten, Netherlands

DECOMMISSIONING AND WASTE MANAGEMENT 417

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Radioactivity Calculation of the Concrete Shielding of the Petten LFR and the Dodewaard BWR ı Lino Salamon, Perry Young and Lojze Gačnik


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DECOMMISSIONING AND WASTE MANAGEMENT 418

Quality Assurance and Data Analysis

in Automated Radiological

Characterization of Large Soil Volumes

Christoph Klein, Marina Sokcic-Kostic and Felix Langer

Planned entry for

At the dismantling of FBFC International’s fuel fabrication plant in Dessel/Belgium, more than

34.000 tons of soil have already been successfully characterized and sorted by NUKEM’s system

FREMES until end of July 2020.

The measurement and sorting process is executed fully automated and can process up to 13 tons

per hour, having involved recording and evaluation of up to 16.500 gamma spectra per day.

To ensure the constant quality of the results, a clear-cut algorithm as well as regular consistency checks and

supervision are essential.

This article describes the principal data evaluation of the system from recording to sorting, and gives an overview about

the taken measures of data analysis and quality assurance. Practical representations and typical examples

from operational experience illustrate, how the system reliably performs highly numerous and frequent gamma

measurements, without necessary expertise of the site operators.

| Fig. 1.

Overview of soil characterization and sorting with FREMES: Material is filled in by truck (right),

then buffered and transported by conveyors to the measurement/sorting belts (blue containers),

and afterwards sorted to piles and containers (left side).

1 Measurement operation

The system FREMES for characterization

and sorting of bulk material can

process large quantities of soil, building

rubble and other, in order to release

or dispose of it properly. It provides

measurement with a high throughput

(up to 13 tons per hour for the soil

measured in Dessel) by uninterrupted

gamma-spectrometric measurements

of a continuous material stream on the

belt, which is then directly evaluated,

characterized and sorted to the resulting

destination. (See Figure 1) By this

direct examination and sorting of

100 % of the material, conservative

assumptions about the activity (and

therefore the amount of radioactive

waste) are minimized.

The central measurement process is

performed according to the following

basic steps:

p The material is buffered and a

­continuous stream with a defined

geometry suitable for measurement

is created. (See Figure 2)

p The material is virtually divided

into cells, for which gamma spectra

are subsequently taken by HPGe

detectors and a scale takes the

weight.

p Immediately after recording,

the data is evaluated and for

each separable portion the contained

specific U-235 activity is determined

automatically, in cluding

uncertainty values ac cording to

ISO 11929.

p The result is compared to legal

limits (including a safety margin)

and classified into the categories:

p Free Release (FR)

p Conditional Release (CR)

p (Low Level) Radioactive Waste

(RW)

p Material suspicious of a hot spot

(HS) or other inconclusive

| Fig. 2.

Material is moved in continuously below the

detectors in a defined geometry suited for

measurement.

activity pattern, which is

collected and re-measured later

p According to the obtained characterization,

the system directs the

material to its storage destination

via sorting belts.

p All results and raw data are stored

on the PC, from which documentation

is automatically provided at

material export.

To guarantee high throughput, all

these steps – including the gamma

spectrum evaluation – run fully automated

in a clear-cut and well-proven

process and therefore can be supervised

by an operator, who does not

need to be an expert for radiation

measurement.

Considering the large amount

of material, this is an important

advantage, especially for costefficiency.

In the application here,

already around 34,000 tons have been

examined over more than 2 years,

­corresponding to around 1.5 million

single gamma measurements and

their evaluation.

However, many aspects of the

system (especially of the gamma

measurement) still need to be supervised

regularly by an expert, to make

sure that the automated evaluation

works reliably and without unintended

effects. To ensure this, NUKEM

employs numerous measures for

prevention of issues, regular maintenance

and quality assurance, both

remotely and together with the personnel

operating the system on-site.

2 Quality assurance

of the results

Regular activities and provisions of

various types make sure that the

results of the system are reliable and

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| Fig. 3.

The operators ensure reliable results and high throughput by super vising

the process constantly over cameras and software displays. The system

also can be controlled outside of the working station by a tablet PC.

that possible issues are prevented or

quickly fixed:

p Operational / Mechanical:

p Constant observation of the

­operators over material flow

and automated operation

p Direct communication of

operators to NUKEM experts

for remediation of issues (telephone,

remote PC connection,

cameras)

p Regular maintenance visits and

checks of the system hardware

p Radiological:

p Daily measurements of radiation

background and check

for changes (performed by the

operators)

p Regular reference measurements

(1-2 days) to ensure

correct function of gamma

detectors and weighing system

p Taking of representative

samples from the material

stream and comparison with

an independent measurement

p Regular consistency checks and

deeper analyses of the recorded

data by a radiological measurement

expert

p Automated synchronization

and transfer of the data to

ensure safe storage

Since the system mechanically processes

a large volume of material at all

seasons (temperatures, weather conditions)

issues of the mechanical systems

and irregularities in the material

flows are natural with these high

volumes of soil. The operators therefore

ensure that the throughput is as

high as possible and the measurement

conditions are always fulfilled, by

super vising the process with the help

of several technical means (see Figure

3) and intervening when necessary.

NUKEM experts are available on

short notice if the solution of more

complicated problems is demanded

(e.g. with the belt movement

auto matisation systems, or for radiological

questions). Via ­telephone and

remote connection to the measurement

PC, problems can be solved in an

easy and efficient way, which has

proven to be very valuably over the

operation period. This allows the quick

solution of most issues and a fast

resumption of operation, without

requiring the presence of an expert

on site in many of the cases, saving

valuable cost and time.

| Fig. 4.

Results of regular QA measurements with a point source (blue), compared to the reference source activity

(orange, with uncertainty area). Remark: Since smaller activities can regularly occur in practice by errors in

positioning of the source on the belt, those values are expected and clear cases have been excluded (red marks).

The reliability of the radiological

results is regularly monitored by an

expert for gamma spectrometry. The

automated spectrum evaluation has

always proven to give reliable results

during tests and operation, so that

there is no need for a sophisticated

on-site supervision of the measurements.

Instead, the results are

regularly reviewed by experts to check

the correctness and consistency of

evaluation, and to identify possible

| Fig. 5.

Typical signature during review of mass measurements over three days (separated red). The material

portions (blue) show stable masses with certain spread and expected inter ruptions (start/stop, pauses,

etc.). A performed change in the material transport on the third day is clearly visible.

| Fig. 6.

Typical example for measurement results from material with specific activity in the region 1-10 Bq/g

U-tot. (Blue: activity results, red: limit of detection.) [Scale shown here in corresponding units of U235.]

DECOMMISSIONING AND WASTE MANAGEMENT 419

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DECOMMISSIONING AND WASTE MANAGEMENT 420

| Fig. 7.

Exemplary display of the measurement results from one working shift during review: A set of

uncontaminated material (activity below detection limit) is followed by material with different

magnitudes of activity, which are sorted above or below the legal limit value, including a safety margin.

| Fig. 8.

Specific activities of a material set, which has been sorted to Free Release. (Data description as in

Figure 7.) The specific activity can be seen reliably below the legal limit, including the safety margin.

changes in the measurement conditions

(e.g. changes of background,

hardware issues of the detectors,

unexpected effects, etc.).

These checks typically consist of:

p Remotely connected inspection of

the measurement hardware

p Check of measurement software

and its history files

p Examination of the regular

measurements (background and

reference source) for unexpected

changes (also performed qualitatively

by the operators)

p Review of the recorded measurement

results

p Manual inspection of raw

measurement data and evaluation

files

p Regular synchronization to

independent data storages, also

­allowing efficient access to the

data for analysis

In addition to these regular checks,

the correct functionality of the system

is checked on site at least every 2 days

by measurement of a reference point

source. These measurements are

required to correctly identify and

provide the activity value of the

source, by which correct energy and

efficiency calibration as well as the

stability of the system can easily be

­assessed by the operators. An example

of these measurements is shown in

Figure 4, which clearly demonstrates

a stable behavior over the measurement

period.

3 Regular review of

measurement results

An important part of the data review

is the graphical display of the

measurement results, by which the

correct system behavior can be

observed and unintended effects can

be noticed. Examples for the typical

signatures of mass and activity

measurement and are shown in

Figure 5 and Figure 6. In the latter

display the quality of the results can

be easily verified by check that the

limits of detection remain at stable

values and that the found activities

results show a steady and plausible

behavior (except deviation from

known irregularities during measurement

operation).

Figure 7 shows a typical example

of change from material with no

measurable U235 activity to such with

activity in the vicinity of the legal limit

value for U-235, which corresponds

to 1 Bq/g Utot. The latter can be

­identified by FREMES to be below

or above the limit and is sorted accordingly.

An example for sorted free

release material in Figure 8, shows

that it has a specific activity

­significantly below the limit and can

be forwarded to release.

4 Summary

The FREMES system now has operated

reliably for over 2 years on the site

in Dessel. It has successfully characterized,

sorted and released over

34,000 tons of material, corresponding

to over one million single

gamma measurements. The radiological

measurement itself is performed

with high throughputs up to

13 tons/hour fully automated under

supervision of operators on site, who

not need to be measurement experts.

The employed numerous quality

assurance procedures thereby guarantee

the reliability with a minimum

effort by experts themselves. The

remediation of different issues (e.g.

irregularities or hardware problems

with material transport equipment)

as well as radiological review of the

data can be performed efficiently by

remote access in many cases.

This allows the system to

accomplish 100 % radiological

characterization of material in a

cost- effective way also in large

amounts and with high throughput

demands, and thereby minimizes the

material to be classified as non-free

release material as far as possible.

Authors

Dr. Christoph Klein

christoph.klein@

nukemtechnologies.de

Dr. Marina Sokcic-Kostic

Felix Langer

NUKEM Technologies Engineering

Services GmbH

Industriestr. 13

63755 Alzenau, Germany

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Errichtung einer Zerlegehalle

für Groß komponenten am Standort des

Entsorgungswerkes für Nuklearanlagen

in Lubmin

Andreas Fuchs und Bernhard Olm

1 Einleitung Für die Konditionierung von radioaktiven Reststoffen und Abfällen, welche

derzeit in den Lagerhallen des Zwischenlagers Nord (ZLN) der EWN GmbH zwischengelagert

werden, beabsichtigt die EWN die Errichtung und den Betrieb einer Zerlegehalle (ZLH) für

Großkomponenten (Reaktordruckgefäße (RDG) und Dampferzeuger (DE), RDG-Einbauten, etc.)

am Standort Lubmin/Rubenow. Aufbau und Auslegungsmerkmale sowie in der ZLH geplante

Tätigkeiten werden vorgestellt.

2 Aufbau der Zerlegehalle

Am Standort Lubmin/Rubenow der

EWN Entsorgungswerk für Nuklearanlagen

GmbH (EWN) befinden sich

im östlichen Bereich des Betriebsgeländes

die Gebäude Zentrale Aktive

Werkstatt (ZAW) und Zentrale Dekontaminations-

und Wasseraufbereitungsanlage

(ZDW).

Für den Betrieb der Einrichtungen

der ZAW und ZDW ist der Umgang mit

radioaktiven Stoffen, insbesondere

die Bearbeitung und Konditionierung

von radioaktiven Reststoffen und

­Abwässern aus den Kernkraftwerken

Greifswald (KGR) und Rheinsberg

(KKR), sowie aus anderen kerntechnischen

Anlagen mit Leichtwasserreaktoren

genehmigt.

Die Möglichkeiten der Bearbeitung

und Konditionierung sind nicht für

alle radioaktiven Reststoffe und

­Abfälle, welche derzeit überwiegend

in den Lagerhallen des Zwischen lagers

Nord – Abfalllager (ZLN-AL) der EWN

GmbH zwischengelagert werden, ausreichend.

Dies betrifft insbesondere

Großkomponenten, z. B. Reaktodruckgefäße

(RDG) und Dampf er zeuger

(DE), sowie Komponenten mit hohem

radiologischem Gefährdungspotential

z. B. RDG-Einbauten in Abschirm- und

Transportvorrichtungen oder in nicht

endlagerfähigen Behältern.

Für die Konditionierung dieser

­radioaktiven Reststoffe und Abfälle,

inklusive endlagerfähiger Verpackung

der Abfälle, beabsichtigt die EWN

die Errichtung und den Betrieb

einer Zerlegehalle (ZLH) für Großkomponenten

am Standort Lubmin/

Rubenow.

Die ZLH soll als Anbau an die

vorhandenen Gebäude südlich der

ZAW und östlich der ZDW errichtet

und betrieben werden. Die ZLH wird

baulich und technologisch mit der

ZDW und der ZAW zu einem Reststoffbearbeitungs-

und Abfallbehandlungszentrum

verbunden und bildet

damit einen zusammenhängenden

Gebäudekomplex.

In der Abbildung 1 ist die Lage der

ZLH zu den angrenzenden Gebäuden

ZAW (Zentrale Aktive Werkstatt) und

ZDW (Zentrale Dekontaminationsund

Wasseraufbereitungsanlage) ersichtlich.

Abbildung 2 ist eine

­Visu­alisierung der Zerlegehalle mit

den angrenzenden Gebäuden ZAW

und ZDW.

Die ZLH besteht im Wesentlichen

aus einem Zerlegebereich mit vorgelagerter

Schleuse, einer Bereit stellungshalle

mit vorgelagerter Schleuse,

der Personenschleuse (Umkleidebereiche

inaktiv und aktiv) mit Kontrollbereichszugang,

Sanitär räumen, einem

Aufenthaltsraum, ­Büros und den

Lüftungszentralen sowie sonstigen

Infra­strukturräumen und Verkehrswegen.

In die Bereit stellungshalle ist

ein Raum zur ­Aufbewahrung von Ölen

und Fetten integriert. Die Abbildung 3

| Abb. 1.

Lageplan der Zerlegehalle.

Planned entry for

DECOMMISSIONING AND WASTE MANAGEMENT 421

| Abb. 2.

3D-Ansicht der Zerlegehalle mit angrenzender ZAW/ZDW.

| Abb. 3.

Schnitt durch den Zerlegebereich mit vorgelagerter Schleuse.

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 422

| Abb. 4.

Schnitt durch die Bereitstellungshalle und die darüber befindliche Personenschleuse.

zeigt einen Schnitt durch die Zerlegehalle

im Bereich der Schleuse sowie

den Zerlegebereich und die bestehende

ZAW.

Abbildung 4 zeigt einen Schnitt

durch die Bereitstellungshalle mit

der darüber befindlichen Personenschleuse.

Die Errichtung und der Betrieb

der ZLH wurde als Änderungs­genehmigung

nach § 7 Strahlenschutzverordnung

(StrlSchV) für den bestehenden

Komplex ZAW/ZDW

beantragt.

3 Ausbaustufen und

geplante Tätigkeiten

In der ZLH werden die im ZLN-AL

lagernden Dampferzeuger und die aktivierten

Komponenten, die beim

Rückbau der Kernkraftwerke Greifswald

(KGR) und Rheinsberg (KKR)

ausgebaut wurden, bearbeitet und

konditioniert.

Bei der Bearbeitung radioaktiver

Reststoffe wird das Ziel verfolgt, den

sicheren Umgang mit radioaktiven

Stoffen zu gewährleisten und weiterhin

die größtmöglichen Massen unter

Beachtung des ALARA-Prinzips und

Abwägung des Aufwand-/Nutzenprinzips

der uneingeschränkten Freigabe

oder der zweckgerichteten Freigabe

nach StrlSchV zuzuführen.

Die bei der Bearbeitung anfallenden

radioaktiven Abfälle werden in endlagerfähige

Behälter verpackt.

Folgende Großkomponenten und

aktivierte Komponenten sind zu bearbeiten:

p Großkomponenten KGR und KKR

p Dampferzeuger (DE) der Blöcke

1, 2, 3 und 4

p Reaktordruckgefäße (RDG) der

Blöcke 1, 2 und 5 des KGR

p Reaktordruckbehälter (RDB) des

KKR

p RDG der Blöcke 3 und 4

des KGR, jeweils mit den

RDG-Einbauten Reaktorschacht

(RS) und Schachtboden (SB)

p RDG-Einbauten Schutzrohrblock

(SRB), Kassettenkorb

(KK), RS und SB des Blockes 5

des KGR

p RDG-Einbauten SRB und KK

der Blöcke 3, 4 und 5 des KGR

p Aktivierte Komponententeile KGR

und KKR

p Bodenplatten der RDG-Einbauten

SRB und KK der Blöcke 1 und

2 des KGR

p Obere Platten der RDG-Einbauten

SB der Blöcke 1 und 2

des KGR

p Segmente der Ringwasserbehälter

(RWB) der Blöcke 1, 2,

3 und 4 des KGR

p Corebauteile (CBT) des KGR

p CBT des KKR

Des Weiteren müssen nachfolgende,

bereits vorzerlegte und in Stahl- und

Betoncontainern verpackte, aktivierte

Abfälle ggf. nach vorheriger Nachzerlegung

in endlagerfähige Behälter

umverpackt werden:

p die RDB-Einbauten SRB, KK, RS

und SB der Blöcke 1 und 2

p die RDB-Einbauten Druckgestell

(DST), Brennelementekorb (BEK),

RS und SB des KKR

p CBT aus der Entsorgung der

Schachtläger für aktivierte Betriebsabfälle

des KGR

p sowie sonstige aktivierte Abfälle

Die Reststoffbehandlung und Abfallkonditionierung

soll in Arbeits­paketen

erfolgen. Die verschiedenen Arbeitspakete

werden – je nach Arbeitsfortschritt

und Planungsstand –

in Ausbaustufen zusammengefasst.

Zwischen der Abarbeitung der verschiedenen

Arbeitspakete müssen

i. d. R. die Gerätschaften zur Reststoffbehandlung

und Abfallkonditionierung

den zur Zerlegung vorgesehenen

Großkomponenten und

Teilen angepasst werden.

Für die ZLH sind aus technologischer

und radiologischer Sicht

vier Ausbaustufen vorgesehen. Die

erste Ausbaustufe – die Dekontamination,

Zerlegung und Verpackung

der Dampferzeuger – wird detailliert

geplant und ist Antragsgegenstand.

Zum jetzigen Zeitpunkt werden

auch die späteren Ausbaustufen 2

bis 4 in einem Detaillierungsgrad

beschrieben, der es erlaubt, wesentliche,

für die Planung des Gebäudes erforderliche

Auslegungsdaten abzuleiten

und darüber hinaus die gene relle

Machbarkeit der zukünf tigen Hantierungsvorgänge

sowie der Zerlege- und

Verpackungsprozesse zu bewerten.

Die detaillierte Planung der Ausbaustufen

2 bis 4 erfolgt jeweils rechtzeitig

vor dem geplanten Beginn der

Tätigkeiten der jeweiligen Ausbaustufe

anhand der dann möglicherweise

neu zur Verfügung stehenden

Hantierungs- und Zerlegetechnik. Die

Ausbaustufen 2 bis 4 werden über

das atomrechtliche Aufsichtsverfahren

unter Berücksichtigung der

zum Zeitpunkt der Antragsstellung

gültigen Gesetzeslage bei der Aufsichtsbehörde

beantragt.

Nachfolgend sind die vorgesehenen

Ausbaustufen aufgelistet:

p Ausbaustufe 1:

Trockenzerlegung (DE)

p Ausbaustufe 2/1:

Trockenzerlegung – Erweiterung

(RWB)

p Ausbaustufe 2/2:

Nass-/Trockenzerlegung (RDG/

RDB, teilweise mit Einbauten)

p Ausbaustufe 3:

Umverpackung der aktivierten

­Abfälle aus Zwischenlagerbehältern

in Endlagercontainer

p Ausbaustufe 4/1:

Trockenzerlegung (RDG-

Einbauten Block 5 KGR)

p Ausbaustufe 4/2: Nasszerlegung

(RDG-Einbauten KGR und CBT

KGR und KKR)

4 Dekontamination

und Zerlegung der

Dampferzeuger

in der 1. Ausbaustufe

In der ersten Ausbaustufe der ZLH

ist geplant, dass die 21 im ZLN-AL

lagernden Dampferzeuger (DE) dekontaminiert,

zerlegt und die entstandenen

Reststoffe dem Stoffkreislauf

(Rezyklierung) zugeführt

werden. Die entstehenden bzw.

­verbleibenden radioaktiven Abfälle

werden in endlagerfähige Behälter

verpackt.

Die DE wurden unter Demontagebedingungen

aus den Blöcken 1-4

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­entnommen und befinden sich als

Großkomponenten in der Halle 7 des

ZLN-AL. Die DE sind entleert und alle

Öffnungen sind mit Stahlblechen verschlossen.

Weiterhin sind die DE an

der berührbaren Oberfläche durch

eine Oberflächenbeschichtung bzw.

Dekontamination frei von abwischbarer

Kontamination.

Das Gesamtgewicht jedes DE

­beträgt 157 Mg. Die max. Gesamtaktivität

pro DE liegt bei 1,1 E+11 Bq

(Bezugszeitpunkt 01/2021).

Vor der Zerlegung der DE werden

die Nadelrohre im Zerlegebereich

primärseitig dekontaminiert (siehe

Abbildung 5), um das oben genannte

Ziel der Rezyklierung der unverschlossenen

Nadelrohre zu erreichen.

Nach erfolgreicher Dekontamination

wird der DE auf den Zerlegeplatz

verbracht, um diesen in liegender

Position zu zerlegen. Der DE wird

eingerüstet, um die Zugänglichkeit zu

gewährleisten (siehe Abbildung 6).

Die Zerlegung kann in einem separaten,

lufttechnisch abgeschlossenen

Zerlegebereich erfolgen. Als Zerlegeverfahren

kommen u. a. Seilsäge,

mechanische Trennwerkzeuge wie

Winkelschleifer, Abbruchhammer,

Stichsäge und thermische Trennwerkzeuge

wie Autogenbrenner zur Anwendung.

| Abb. 5.

Dampferzeuger im Zerlegebereich während der Dekontamination.

| Abb. 6.

Dampferzeuger auf dem Zerlegeplatz.

DECOMMISSIONING AND WASTE MANAGEMENT 423

5 Wesentliche Merkmale

der weiteren

Ausbaustufen

Ab Ausbaustufe 2 wird eine Verpackungsstation

eingerichtet, die es

ermöglicht, die zerlegten Komponenten

endlagergerecht in Konrad-

Container zu verpacken.

Dabei werden zunächst die Ringwasserbehältersegmente

nachzerlegt

und verpackt.

Diese befinden sich in 20’-Containern

(52 Stück). Die RWB-Segmente

befinden sich in liegender Position

auf einem Tragrahmen im Container.

Die Masse der einzelnen RWB-

Segmente variiert zwischen ca. 10

und 22 Mg.

Die Anlieferung der 20’-Container

in die ZLH erfolgt mit einer Transporteinheit

bestehend z. B. aus Zugmaschine

und Plattenwagen (siehe

Abbildung 7). Die Transporteinheit

fährt über die vorgelagerte Schleuse

in den Zerlegebereich, wo der

20’-Container mit dem Brückenkran

abgeladen wird. Nachdem die

Transporteinheit den Zerlegebereich

verlassen hat, wird der Containerdeckel

abgenommen. Das RWB-Segment

wird mit dem Brückenkran und einer

Traverse entladen und auf dem

| Abb. 7.

Antransport des Containers.

| Abb. 8.

Zerlegung RDG Block 5.

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

DECOMMISSIONING AND WASTE MANAGEMENT 424

| Abb. 9.

Nasszerlegung: geöffneter RDG mit Peripherie.

vorbereiteten Zerlegeplatz abgelegt.

Anschließend wird der 20’-Container

verschlossen und aus dem Zerlegebereich

herausgefahren.

Die Zerlegung der RWB-Segmente

soll aus jetziger Sicht in liegender

Position erfolgen. Die Zerlegung wird

an der Luft in einem separaten lufttechnisch

abgeschlossenen Zerlegebereich

(Einhausung) mit lokalen

­Abschirmungen und Interventionsmöglichkeiten

sowie Absaugung erfolgen.

Als Zerlegeverfahren kommen

u. a. die Seilsäge, mechanische Trennwerkzeuge

wie Winkelschleifer, Abbruchhammer

und thermische Trennwerkzeuge

wie Autogenbrenner zur

Anwendung. Zur Minimierung der

endzulagernden Abfallmenge werden

ggf. zerlegte Teile in der ZAW nachzerlegt

und dekontaminiert, damit

eine Freimessung nach den gültigen

Freigabekriterien erfolgen kann.

Die Teile, die nicht freigemessen

werden können, werden in endlagerfähige

Behälter verpackt und

ggf. entsprechend den Anforderungen

für das Endlager Konrad in der

Bereit stellungshalle mit Beton vergossen.

Darauffolgend werden – ebenfalls

im Rahmen der 2. Ausbaustufe – sämtliche

im ZLN gelagerte RDG/RDB

zerlegt.

Zunächst ist beabsichtigt, das

Reaktordruckgefäß Block 5 in liegender

Position trocken zu zerlegen

(Ab bil dung 8). Die Zerlegung wird an

der Luft in einem separaten lufttechnisch

abgeschlossenen Zerlege bereich

(Einhausung) mit lokalen Abschirmungen

und Interventionsmöglichkeiten

erfolgen. Als Zerlegeverfahren

kommen u. a. die Bandsäge, die Seilsäge,

mechanische Trennwerkzeuge

(fernhantiert oder fernbedient) und

thermische Trennwerkzeuge wie

Auto­genbrenner (fernhantiert oder

fern­bedient) zur Anwendung.

Die RDG der Blöcke 3 und 4 mit

ihren Einbauten RS und SB sind aufgrund

ihrer Masse und ihrer geometrischen

Gegebenheiten auslegungsbestimmend

für die Abmessungen des

Zerlegebereiches, für die Kranhakenbzw.

Gebäudehöhe sowie für die

statische Belastung. Sie sind somit

abdeckend für die Betrachtung der

Großkomponenten bezüglich der

maximal zu handhabenden Massen

und der maximal zu handhabenden

Geometrien.

Die RDG der Blöcke 3 und 4

wurden jeweils im Ganzen aus ihrer

Einbaulage entfernt und in liegender

Position ins ZLN-AL verbracht.

Zuvor wurde ein ca. 5,5 m langer

| Abb. 10.

Umverpackung mit Abschirmglocke.

Abschirmzylinder im Core-Bereich

außen am RDG montiert. Anschließend

wurden die RDG-Einbauten RS

und SB in das RDG eingesetzt. Abschließend

wurde jeweils eine Abschirmplatte

mit Lastanschlagpunkt

(Lastaufnahme­öse) unter Verwendung

der vorhandenen Gewindesacklochbohrungen

am RDG-Flansch verschraubt.

Für diese RDG wurde ein Konzept

entwickelt, bei dem vorgesehen ist,

die Behälter in der Grube des Zerlegebereiches

aufzurichten, mit Wasser zu

füllen und die enthaltenen Einbauten

fernbedient unter Wasser zu zerlegen

(siehe Abbildung 9)

In Ausbaustufe 3 ist im Wesentlichen

ein Umverpacken von zerlegten

und in Stahl- oder Betoncontainern

verpackten Teilen in endlagerfähige

Behälter vorgesehen.

Es befinden sich Einsatzkörbe/

Sammelbehälter mit aktivierten bzw.

kontaminierten Abfällen in Stahloder

Betoncontainern im Abfalllager

des Zwischenlagers Nord. Bei den

­Abfällen handelt es sich u. a. um

­Reaktoreinbauten der Blöcke 1 und 2

des KGR bzw. des Rheinsberger

Reaktors.

Diese Abfälle befinden sich in ­

Stahl- oder Betoncontainern, welche

keine Endlager-Zulassung besitzen.

Deshalb müssen diese in endlagerfähige

Behälter umverpackt werden.

Auf Grund der radiologischen Randbedingungen

erfolgt die Um verpackung

fernbedient. Hierfür wird die im Übergang

zwischen Zerlege bereich und

­Bereitstellungshalle eingebaute Verpackungsstation

aus der Ausbaustufe 2

weitergenutzt. Zusätzlich wird im

Zerlegebereich eine Umverpackungsstation

errichtet. In Abbildung 10

sind die Umverpackungs station und

die Verpackungsstation konzeptionell

dargestellt.

Decommissioning and Waste Management

Construction of a Dismantling Hall for Large Components at Entsorgungswerk für Nuklearanlagen in Lubmin ı Andreas Fuchs and Bernhard Olm


atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Abb. 11.

Nasszerlegebecken – Einrichtung als Trockenzerlegeplatz.

| Abb. 12.

Absetzen eines Kassettenkorbes im Nasszerlegeplatz.

vorgesehen, die Inbetriebnahme

2023.

Die derzeitige Terminplanung

sieht die Verarbeitung der im ZLN

gelagerten Großkomponenten in der

Zerlegehalle bis 2059 vor.

Mit der Errichtung der Zerlegehalle

für Großkomponenten am

Standort des Entsorgungswerkes für

Nuklearanlagen in Lubmin wird eine

wesentliche Voraussetzung dafür

geschaffen, dass sämtliche im

­Zwischenlager Nord (ZLN) befindlichen

Reststoffe behandelt und in

endlagerfähige Verpackungen überführt

werden können.

Autoren

Andreas Fuchs

andreas.fuchs@steag.com

Steag Energy Services GmbH

Rüttenscheider Straße 1–3

45128 Essen, Germany

Bernhard Olm

bernhard.olm@ewn-gmbh.de

EWN Entsorgungswerk

für Nuklearanlagen GmbH

Latzower Straße 1,

17509 Rubenow, Germany

DECOMMISSIONING AND WASTE MANAGEMENT 425

| Abb. 13.

Baufeld nach Fertigstellung der Sauberkeitsschicht im Mai 2020.

In Ausbaustufe 4 wird das für die

Grube im Zerlegebereich geplante

Becken eingebaut und im ersten

Schritt als Trockenzerlegeplatz genutzt

(siehe Abbildung 11).

Zunächst ist die Zerlegung von in

sogenannten Abschirm- und Transportvorrichtungen

(ATV) verpackten

Reaktoreinbauten vorgesehen.

Nach Einbau der Nasszerlegetechnik

als Nasszerlegeplatz erfolgt die

Unterwasserzerlegung von hochaktivierten

RDB-Einbauten und CBT,

wie in Abbildung 12 dargestellt.

6 Aktueller Stand

des Projektes

Die Errichtung der Zerlegehalle hat

mit der Herstellung der Tiefgründung

auf ca. 330 Großbohrpfählen auf

Basis einer vorgezogenen Teilbaugenehmigung

im November 2018

begonnen (siehe Abbildung 13). Die

letzten Pfähle wurden im Januar 2020

fertig gestellt.

Der Hochbau beginnt mit der

Herstellung der Bodenplatte im

Sommer 2020, die Fertigstellung

des Gebäudes ist bis Ende 2022

Decommissioning and Waste Management

Construction of a Dismantling Hall for Large Components at Entsorgungswerk für Nuklearanlagen in Lubmin ı Andreas Fuchs and Bernhard Olm


atw Vol. 65 (2020) | Issue 8/9 ı August/September

426

ENVIRONMENT AND SAFETY

Planned entry for

Current Procedure for Determining

Release Parameters for a Plane Crash

on a Nuclear Facility in the Context

of Accident Analyses

Steffen Böhlke and Henrik Niegoth

The possible accidents in nuclear

facilities are examined in an accident

analysis according to §13 (1) (6a)

StrlSchG /L-2/. The determination of

the range of assumed incidents is

performed on a case-by-case basis

taking into account location- and

­facility-specific conditions and using

the definitions contained in the

Incident Guidelines /L-3/, in the

various Nuclear Waste Management

Commission (= Entsorgungskommission

(ESK)) guidelines /L-4/, /L-5/

and /L-6/, and in the licensing constraints

and conditions. A distinction

must be made between mal functions,

design basis accidents, and beyonddesign-basis

events. The design basis

accidents are differen tiated into the

categories “internal impacts” and

“ external impacts”. The external

­impacts in turn are classified as

natural or man-made impacts.

Due to the low probability of

occurrence, the incident “plane crash”

normally is included among the

beyond-design-basis events for which

measures to reduce damage have to be

considered. Such measures then are

adequate if the radiological effects do

not necessitate any relevant disaster

protection measures in accordance

with /L-7/.

In recent years, STEAG Energy

Services has established a licensable

standard procedure for accident

analyses for new-build nuclear

facilities or facilities undergoing

modernization and refurbishment,

inside and outside Germany, which is

generally approved by experts and

licensing authorities. For this procedure

the “plane crash” incident is

always a special case since the release

parameters cannot be directly taken

from generally applicable bodies of

regulations, standards or codes.

1 Introduction For protection against major safety-related incidents in nuclear facilities,

for example a plane crash on a radioactive waste storage facility or a conditioning plant, provision

has to be made in the planning for structural, or other technical safety measures against potential

accidents in order to limit the release of radioactive substances in the environment

of the facility. The planning limits contained in § 104 of the “Radiation Protection Ordinance”

(= Strahlenschutzverordnung (StrlSchV) /L-1/), “Limitation of exposure due to accidents”,

in conjunction with § 194 StrlSchV/L-1/, are to be taken as basis for this.

In the following, the general procedure

for determining radiological

impacts is explained taking the

example of a plane crash on a storage

building for radioactive waste. In

addition to the determination of the

load impacts occurring in case of a

plane crash, it is necessary to

­determine the nuclide-­specific activity

inventory of the waste packages and

the respective release parameters

and to perform the cor responding

dispersion and dose cal culations in

order to evaluate the facility’s radiological

impact on its environment.

The method can also be adapted to

conditioning plants for radioactive

waste, i.e., facilities for treatment of

residual materials.

2 Plane crash scenario

The hypothetical scenario refers to a

fully fueled military aircraft crashing

at high speed into a storage building

containing waste packages, causing a

fire within the facility because of the

kerosene carried by the aircraft. The

appropriate load assumptions for a

plane crash, e.g. details of the impact

load-time diagram or the impact

area and impact angle, are described

in the Reactor Safety Commission

(= Reaktor-Sicherheitskommission

(RSK)) guidelines /L-8/.

The sequence of a worst-case plane

crash into a storage building can be

described as follows: As conservative

assumption, the military aircraft

crashes vertically into the roof of the

storage building. Upon penetrating

the roof of the building, it punches a

piece of concrete debris out of the

roof. At the same time, as parts break

off and kerosene leaks out, the body of

the aircraft loses mass, the kinetic

energy of which, however, does not

contribute to the pattern of damage.

The rest of the aircraft is decelerated

and, together with the concrete

debris, falls with reduced velocity

onto the waste packages stored below,

inside the storage building. If a

­statistical verification is not available,

additionally the collapse of the

collapse of further parts of the roof is

assumed. The beams of the roof

structure next to the area of impact

are also affected by the plane crash. A

conservative assumption is that thereby

the roof structure will collapse

across the entire width of the building

damaging the waste packages below.

In the affected stacks of packages,

the kinetic energy is consumed proportionately

and, depending on the

mass of the impacting pieces of debris

and the masses of the packages in the

stack, is distributed among the waste

packages through impact analysis.

Mechanical loading can occur in the

process, acting, on the one hand, on

waste packages stored directly under

the area of impact and, on the other

hand, on neighboring waste packages,

as a result of falling parts of the roof.

Waste packages located only partially

within one of the two areas subjected

to loading are considered to be fully

impacted. This covers the action of

­additionally flying debris.

Furthermore, a fire has to be

ex pected to occur subsequent to

the crash due to the kerosene carried

on board the military aircraft.

Making allowance for the debris

­covering the floor after the impact,

the depth of the resulting kerosene

lake is ascertained through geometric

analysis in order to determine the

maximum duration of a fire. Within

the scenario the fire is assumed to

burn for at least 30 minutes. As a

­consequence of the kerosene fire

on the floor of the entire storage

Environment and Safety

Current Procedure for Determining Release Parameters for a Plane Crash on a Nuclear Facility in the Context of Accident Analyses

ı Steffen Böhlke and Henrik Niegoth


atw Vol. 65 (2020) | Issue 8/9 ı August/September

facility, a part of the waste packages

additionally are subjected to thermal

loading.

3 Nuclide vector and

activity inventory

In addition to the properties of the

waste packages themselves, par ticularly

the radiological characteristics

of the radioactive wastes, for example

the nuclide composition or the activity

inventory, are important input data

for determining the radiological

effects subsequent to an accident.

A nuclide composition that covers

the occurring radioactive wastes can

be estimated on the basis of analyses

of the operation and post-operation of

the nuclear power plant at the site of

the planned storage building, or on

the basis of experience from the

dismantling of parts of other nuclear

power plants. For calculation of the

release fractions, the nuclides are

­classified particularly on the basis of

their volatility under the assumed

load impacts.

To determine the activity inventory

one can rely on empirical values or

concrete information from databases

of existent nuclear facilities. In the

case of new-build facilities, conservative

assumptions are made with the

aid of radiological planning guides.

The activity inventory of the waste

packages also can be deduced from

dose rate criteria.

4 Release fractions

Extensive investigations of release

parameters and concrete information

about release fractions can be found in

the Konrad Transport Study /L-9/,

together with the corrections in the

corresponding more detailed study

/L-10/ and verification /L11/, as well

as in the ESK statement on the stress

test for storage facilities for low- and

intermediate-level radioactive waste,

stationary facilities for the con ditioning

of low- and intermediate-level

radioactive waste, disposal facilities

for radioactive waste in Germany

/L-12/. These studies are suitable in

themselves for determining release

Impact speed

fractions of radioactive residues

and wastes resulting from the load

impacts of a plane crash.

In the following, the term Konrad

Transport Study is understood to

include both the more detailed study

and the verification. The calculation

options of the Konrad Transport

Study /L-9/ and the ESK statement

on the stress test /L-12/ can be

applied independently of each other

and also lead to different release

fractions.

4.1 Konrad Transport Study

In the Konrad Transport Study /L-9/,

release fractions are determined

experimentally for certain package

and waste types and different load

cases. In addition, release fractions

are extended or generalized for

further package types and load cases

on the basis of empirical and physical

dependencies.

In conformity with the conditioning

and packaging requirements

for the Konrad repository in Germany,

in /L-9/ accident loads of transport

vehicles and transport casks are

­classified in nine severity categories

(= Belastungsklassen (BK)) and possible

waste package combinations in

eight waste package groups (= Abfallgebindegruppen

(AGG)), adequately

covering the range of possible package

and load case combinations. Solely

containers approved for transport and

storage in the Konrad final repository

are taken into account, and solely load

cases which occur on the basis of the

specified modes of transport are

considered (transport by road and

rail). In the severity categories, both

mechanical and thermal loads are

taken into account for the release.

The severity categories are listed in

the following Table 1.

The classification of the mechanical

loads according to the Konrad

Transport Study /L-9/ goes to the

maximum of a load equivalent of an

impact of the package at a speed of

110 km/h. The debris resulting from a

plane crash will have substantially

higher speeds, well in excess of

Duration and temperature of fire

No fire 30 min at 800 °C 60 min at 800 °C

0 – 35 km/h BK 1 BK 2 BK 3

36 – 80 km/h BK 4 BK 5 BK 6

80 – 110 km/h BK 7 BK 8 BK 9

| Tab. 1.

Severity categories used in the Konrad Transport Study.

110 km/h, and larger masses. The

­severity category definitions initially

are inadequate for analyzing the load

impacts resulting from a plane crash.

The high energy input in the packages

as a result of the plane crash therefore

makes it necessary to extrapolate

the mechanically induced release

fractions for the non-respirable particles

with an aerodynamic equivalent

diameter (AED) of > 10 µm. Extrapolation

of the release fractions is

­performed using the specific energy

input into the packages during impact.

The extrapolation formula for

the release fractions is:

Here f 110 and f 80 represent the release

fractions of the corresponding waste

package groups at 80 km/h and

110 km/h, respectively, and v the

velocity of the impacting piece of

debris.

Since the release fractions in the

Konrad Transport Study /L-9/ always

refer to packages with a mass of

11 tons and a volume of 7.4 m 3 , the

release fractions for package types

with a different mass and volume

must be adjusted. The following scales

are to be applied:

For AGG 5 and 7:

For AGG 1, 2, 3, 4 and 6:

where m package mass

V package volume

f release fraction of AGG

for the severity class (BK)

f korr corrected release fraction

For smaller volumes or masses the

release fractions increase. The release

fractions of the particles with AEDs

from 0 to 10 µm and 10 to 100 µm,

­derived from the two studies /L-9/

and /L-10/, scaled, and extrapolated

for the plane crash, are conservatively

added to determine the releases.

Hence, from the individual release

fractions, for every waste package

a total release fraction can be put

together from mechanical and

thermal loads.

ENVIRONMENT AND SAFETY 427

Current Procedure for Determining Release Parameters for a Plane Crash on a Nuclear Facility in the Context of Accident Analyses

Environment and Safety

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

ENVIRONMENT AND SAFETY 428

Package type Thermal impact Large-area mechanical impact Punctiform mechanical impact

Combustible

raw waste

| Tab. 2.

Release fractions in accordance with the ESK stress test.

Combustible

conditioned

waste

Noncombustible

waste

Raw waste,

of which 50%

respirable

Conditioned

waste, of which

10% respirable

Raw waste,

of which 50%

respirable

Conditioned

waste, of which

10% respirable

Cast iron cask 2,00E-05 2,00E-05 2,00E-05 0 0 0 0

Concrete cask 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03

Konrad IV container 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03

20-ft. container 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03

200-l drums 5,00E-01 4,00E-03 5,00E-04 1,00E-02 4,00E-04 1,50E-01 6,00E-03

4.2 ESK stress test

To establish the release fractions for a

plane crash, it is also possible to use

the release fractions cited in the ESK

statement on the stress test /L-12/.

Typical damage patterns were defined

by ESK for this purpose, covering

all types of serious impacts on the

facilities under investigation, including

the waste packages they

contain. In /L-12/ these typical

damage patterns are assumed and the

resistance of the facilities against

these damage patterns is evaluated.

In regard to the typical damage

patterns of a pattern crash, the following

types of impacts have to be considered:

p Thermal impacts due to a longer

lasting fire (escape of fuel and

burnup in the area of the radioactive

wastes)

p Mechanical impacts on waste

packages, taking into account the

difference in energy input:

p Punctiform mechanical impact

(impact of an engine shaft on

packages)

p Large-area mechanical impact

(roof truss falls on packages)

In /L-12/, for the scenarios of the

possible impacts, release fractions are

compiled for cast iron and concrete

casks as well as Konrad containers,

20-foot containers and 200-liter steel

drums as functions of the degree of

conditioning (raw waste, conditioned

waste) and the combustibility of the

inventory of the waste packages. The

basis for /L-12/ is, inter alia, the

­Konrad Transport Study /L-9/, which

was already used for the mechanical

impacts.

The relevant release fractions are

shown in the following Table 2:

5 Determination

of source term

The activity released into the environment

of the facility (= source term)

is determined as the product of the

­nuclide-specific activities of the

affected waste packages and the

respective release fractions of the

mechanical-thermal impacts. Due

account has to be taken of the

number of affected waste packages

which have been subjected to

mechanical and thermal loading as

a result of the plane crash.

Scenario-specific restraining ­effects

possibly also need to be considered.

This can be, for example, debris of

the aircraft or the concrete roof, which

contribute in a certain way to retain

the radioactive inventory in the

environment of the facility.

6 Dispersion and

dose calculations

To determine the exposure in the

environment of the facility, dispersion

and dose calculations are made.

The definitive basis for calculations

is the incident calculation bases

(= Störfall­berechnungsgrundlagen

(SBG)) /L-13/. The dose rate coefficients

from the compilation of dose

coefficients for external and internal

radiation exposure – Part I and Part II

– published in the Bundesanzeiger

(German Federal Government’s

­Bulletin) of 23 July 2001 /L-14/, are

used for the calculations.

| Fig. 1.

Sample-dose-distributions in the environment of a potential local release.

Environment and Safety

Current Procedure for Determining Release Parameters for a Plane Crash on a Nuclear Facility in the Context of Accident Analyses

ı Steffen Böhlke and Henrik Niegoth


atw Vol. 65 (2020) | Issue 8/9 ı August/September

As result, the effective doses in a

calendar year for the different age

groups and different exposure pathways

are determined at the most

unfavorable point of exposure in

the environment of the facility. The

extent to which the intervention

reference level of 100 mSv assessed

for the plane crash is exhausted is

established and evaluated according

to /L-7/.

A graphic description of a sample

dose distribution in the environment

of a potential local release is provided

by the following illustrations (see

Figure 1).

7 Summary

By means of the

p scenario definition of the beyond-­

design-basis event “plane crash”,

p the determination of nuclide

vectors and activity inventories,

and

p the determination of release

fractions and source terms,

described in this article, a standard

procedure approved by authorities

and the consulted experts has been

established which enables committed

doses following a plane crash on

existing facilities for the dismantling

of nuclear facilities, or such facilities

still to be built, to be determined

in a comprehensible and verifiable

manner, without incurring approval

risks.

The described methods also can

be adapted to determine committed

doses of other less intensive scenarios

involving smaller load impacts

(e.g. the falling of packages, or

earthquakes).

References

/L-1/

/L-2/

/L-3/

/L-4/

/L-5/

/L-6/

/L-7/

/L-8/

/L-9/

Verordnung zum Schutz vor der schädlichen Wirkung

ionisierender Strahlung (Strahlenschutzverordnung –

StrlSchV) vom 29. November 2018 (BGBl. I S. 2034, 2036),

die durch Artikel 1 der Verordnung vom 27. März 2020

(BGBl. I S. 748) geändert worden ist.

Gesetz zum Schutz vor der schädlichen Wirkung

ionisierender Strahlung (Strahlenschutzgesetz – StrlSchG)

vom 27. Juni 2017 (BGBl. I S. 1966), das zuletzt durch

Artikel 248 der Verordnung vom 19. Juni 2020 (BGBl. I S.

1328) geändert worden ist

Leitlinien zur Beurteilung der Auslegung von Kernkraftwerken

mit Druckwasserreaktoren gegen Störfälle im

Sinne des § 28 Abs. 3 der Strahlenschutzverordnung –

Störfall-Leitlinien - vom 18. Oktober 1983

ESK-Empfehlung vom 10.06.2013: „ESK-Leitlinien für die

Zwischenlagerung von radioaktiven Abfällen mit

vernachlässigbarer Wärmeentwicklung“

ESK-Empfehlung vom 10.06.2013: „Leitlinien für die

trockene Zwischenlagerung bestrahlter Brennelemente

und Wärme entwickelnder radioaktiver Abfälle in

Behältern“

ESK-Empfehlung vom 16.03.2015: „Leitlinien zur

Stilllegung kerntechnischer Anlagen“

Verordnung zur Festlegung von Dosiswerten für frühe

Notfallschutzmaßnahmen (Notfall-Dosiswerte-Verordnung

– NDWV) (SSK) vom 29. November 2018 (BGBl. I S. 2034,

2172)

RSK-Leitlinien für Druckwasserreaktoren,

Ursprungsfassung (3. Ausgabe vom 14. Oktober 1981)

mit Änderungen vom 15.11.1996

Transportstudie Konrad, Sicherheitsanalyse zur

Beförderung radioaktiver Abfälle zum Endlager Konrad,

Gesellschaft für Anlagen- und Reaktorsicherheit mbH

(GRS), 2009, mit Corrigendum vom April 2010

/L-10/ GRS: Vertiefung und Ergänzung ausgewählter Aspekte der

Abfalltransportrisiko-analyse für die Standortregion der

Schachtanlage Konrad, Abschlussbericht zum Vorhaben

3607R02600, Arbeitspaket 1, Teilaufgaben 11-14, Bericht

GRS-A-3684 vom Februar 2013

/L-11/ Richter, C., Forell, B., Sentuc, F.-N.: Überprüfung des

unfallbedingten Freisetzungsverhaltens bei der

Beförderung radioaktiver Stoffe, Abschlussbericht,

Arbeitspaket 3, GRS - 482, Oktober 2017

/L-12/ ESK-Stellungnahme vom 11.07.2013: „ESK-Stresstest

für Anlagen und Einrichtungen der Ver- und Entsorgung

in Deutschland, Teil 2: Lager für schwach- und

mittelradioaktive Abfälle, stationäre Einrichtungen zur

Konditionierung schwach- und mittelradioaktiver Abfälle,

Endlager für radioaktive Abfälle“

/L-13/ BMU: Störfallberechnungsgrundlagen, Stand Oktober

1983, mit Kapitel 4 “Berechnung der Strahlenexposition”

gemäß § 49 StrlSchV, Stand September 2003

/L-14/ Bekanntmachung der Dosiskoeffizienten zur Berechnung

der Strahlenexposition vom 23. Juli 2001, Bundesanzeiger

Verlag

Authors

Dr. Steffen Böhlke

Head of Licensing & Nuclear

Calculation

steffen.boehlke@steag.com

ENVIRONMENT AND SAFETY 429

Henrik Niegoth

Licensing & Nuclear Calculations

STEAG Energy Services GmbH

Rüttenscheider Str. 1-3

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Current Procedure for Determining Release Parameters for a Plane Crash on a Nuclear Facility in the Context of Accident Analyses

Environment and Safety

ı Steffen Böhlke and Henrik Niegoth


atw Vol. 65 (2020) | Issue 8/9 ı August/September

430

RESEARCH AND INNOVATION

Evaluation of MACST Strategies for

Extended Loss of AC Electric Power

Event in OPR1000 Nuclear Power Plants

Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim

1 Introduction In the nuclear power plant (NPP), station black out (SBO) is initiated by simultaneous loss

of offsite power (LOOP) and operational failure of both emergency diesel generators (EDGs). In such cases, the primary

operator action is required to recover alternative alternating current (AAC) power by manually operating an AAC diesel

generator (AAC DG). If the AAC DG is also unavailable, the plant remains inoperable from all AC power recovery over

the long term. After all the extended loss of AC power (ELAP) event occurs.

The ELAP event has been considered

as one of the consequent accidents

initiated by natural disasters. In this

event, only active systems powered by

direct current (DC) from batteries and

the passive accident coping measures

such as turbine driven auxiliary feed

water pumps (TD-AFWPs) play an

essential role. So, the functions,

operational range and capacity of the

passive systems could be significant

factors during the event [1]. The

­response strategy of the ELAP event

can thus be very limited.

The component cooling water

system (CCWS) and charging pumps

providing seal injection water to the

reactor coolant pumps (RCPs) are also

not available during the ELAP event.

Stoppage of seal injection flow can

cause the inflow of coolant into

the seal cartridge, which results in

exposing the RCP seals at a high

temperature. Maintaining the RCP

seals at a high temperature can

degrade seal materials and increase

leak rates. In such conditions, the

continuous loss of the inventory could

occur in the reactor coolant system

(RCS) through a RCP seal leakage and

it could bring about core uncovery

and damage.

Early cooldown and depressurization

of the RCS could minimize the

inventory loss through the RCP seals

and provide rapid RCS make-up

through the injection of borated water

from the safety injection tanks (SITs).

According to the coping strategy of

the ELAP event, RCS cooldown and

depressurization are conducted via

main steam release to the atmosphere

through the main steam safety valves

(MSSVs) and/or atmospheric dump

valves (ADVs) of the main steam

supply system. The RCS conditions

will decrease to the actuating pressure

of the SITs and then reach the entry

| Fig. 1.

OPR1000 NPP nodalization.

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Evaluation of MACST Strategies for Extended Loss of AC Electric Power Event in OPR1000 Nuclear Power Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Event no. Event Set value

10

1 Initiating event SBO

2 Reactor trip

3 RCP trip

4 Turbine trip

5 MSSVs open SG pressure: 8.618 MPa

6 Fail to operate AAC DG ELAP

7 Start of a TD-AFWPs operation SG level: 23.6 %

8 Completing load shedding

9 ADVs open Cooling rate: 50 °F/hr

| Tab. 1.

Sequence of ELAP event.

Completing connection

of a 1 MW mobile generator

condition of the shutdown cooling

system (SCS). Available AFWPs will

continuously supply water flow to

steam generators (SGs) to make-up

for steam release.

Operators will regulate ADVs to

control the amount of the steam

release and the RCS cooldown rate as

necessary. Because all AC power is not

available, ADVs should be operated

by local manual control during the

ELAP event. If the ADV local control is

not available, the RCS will not be

depressurized and reach the injection

pressure of SITs. In such cases, the

back-up strategy should be necessary

to compensate for RCS inventory loss

to avoid core uncovery and damage,

because installed equipment has

­certain limitations under the ELAP

conditions.

In this study, we present effects of

the RCP seal leakage on the safety

capabilities of the NPP during the

ELAP event. A comparative study is

conducted according to whether the

RCP seal leakage occurs. We also

examine the feasibility of the back-up

strategy using mobile facilities by

­assuming the worst in the ELAP event.

The target plants are the OPR1000

NPP and other conditions for the

target scenario are adopted from

coping strategies presented in the

Stress Test Report [2]. To analyze the

thermal hydrodynamic behavior of

the plant, RELAP5 Mod 3.3 is used

[3]. The nodalization diagram of the

OPR1000 NPP is shown in Figure 1.

2 Effects of RCS seal

leakage

This section aims to provide effects of

the RCS seal leakage on the NPP

­systems at an early stage in the ELAP

event. The comparative case study is

performed for two different cases

whether the RCP seal leakage occurs.

Table 1 shows a sequence of the ELAP

event until completion of connecting a

1 MW mobile generator to cope with

total loss of AC power according to the

stress test guideline of Nuclear Safety

and Security Commission (NSSC) [4].

A RCP seal leakage coincides with

loss of all AC power, leading to

­stoppage of seal injection flow to

RCPs from charging pumps. The initial

seal leak rate is assumed to be 25 gpm

(1.58 l/s) per one RCP [5]. The

RCP seal leaks will be completely

stopped by activating a charging

pump supplying seal injection flow

­after recovering AC power by connecting

of the 1 MW mobile generator.

All conditions for the analysis are

| Fig. 2.

Pressurizer pressures.

adopted from coping strategies

presented in the Ref. [2].

Two transient cases are analyzed in

this section. For Case 1, we consider

an initial seal leakage rate of 25 gpm

(1.58 l/s) per one RCP. Case 2 assumes

that a RCP seal leakage does

not occur during the ELAP event.

Figures 2 and 3 compare pressure and

water level changes of the pressurizer,

respectively. Transient behavior of

two cases shows obvious differences

depending on the RCP seal leakage. In

Case 1, the pressurizer pressure and

water level are rapidly reduced from

the beginning of the event due to the

continuous loss of the RCS inventory

by the seal leakage. The water level of

the pressurizer in Case 1 is completely

empty within one hour. On the other

hand, the pressure and water level of

the pressurizer in Case 2 are initially

maintained and start to decrease after

ADVs open. The water level of the

pressurizer is totally exhausted after

three hours.

Figure 4 shows temperature

changes of hot legs for both cases. The

results of Case 1 and Case 2 are similar

in overall trend. For the first two

hours, hot legs temperature is maintained,

but starts to decrease after the

plant cooling begins. The cooling rate

RESEARCH AND INNOVATION 431

| Fig. 3.

Pressurizer water levels.

| Fig. 4.

Hot legs temperatures.

Research and Innovation

Evaluation of MACST Strategies for Extended Loss of AC Electric Power Event in OPR1000 Nuclear Power Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


atw Vol. 65 (2020) | Issue 8/9 ı August/September

RESEARCH AND INNOVATION 432

| Fig. 5.

RCP seal leak rates.

| Fig. 6.

Core collapsed water levels.

| Fig. 7.

Steam generator pressures.

| Fig. 8.

Pressurizer pressure (Phase 1).

is 50 oF/hr (27.8 oC/hr) as described

in Table 1. Variations of the RCP seal

leak rate for one RCP are shown in

Figure 5. In Case 1, the seal leak rate

continues to decrease as pressure of

the primary system is reduced. On the

contrary, the seal leakage does not

occur in Case 2.

The changes of core collapsed

water level for both cases are depicted

in Figure 6. In Case 1, the abrupt drop

of the core collapsed water level is

observed at the early stage due to the

continuous loss of the inventory

through RCP seals. On the other hand,

the core collapsed water level of

Case 2 starts to decrease three hours

later than that of Case 1. In both cases,

however, the core collapsed water

level does not reach the top of the

active core. Thus, core degradation

and melting due to the core uncovery

do not occur during the event. When

the pressurizer pressure decreases to

the set point of the SIT actuation, the

core collapsed water level starts to

recover as shown in Figure 6.

The pressure behavior of SGs

during the event period shows no

meaningful differences between two

cases as shown in Figure 7. In both

cases, the pressure of SGs is maintained

during the first two hours by

opening MSSVs and starts to decrease

after ADVs open.

3 MACST strategy

In previous section, we paid attention

to an effect of the RCP seal leakage.

The RCP seal leakage had a great

effect on the primary pressure and

Phase no.

Phase 1

(0 ~ 8 hr)

Phase 2

(8 ~ 72 hr)

Phase 3

(72 hr ~ )

| Tab. 2.

MACST Strategy.

Coping strategy

Installed equipment

MACST facilities

supplementing installed

equipment

All on/off-site

equipment

inventory conditions. The abrupt drop

of the core collapsed water level was

observed from the early stage due to

rapid depletion of the pressurizer

­water level. We identified that depressurization

of the RCS can reduce the

inventory loss through the RCP seals

and provide the injection of borated

water from the SITs at an early stage.

If the operation of the ADVs is failed,

the RCS will not be depressurized and

reach the injection pressure of SITs.

The U.S. Nuclear Regulatory

Committee (NRC) has developed

­flexible mitigation (FLEX) strategy

for increasing defense-in-depth for

beyond-design-basis events such as

the ELAP and loss of ultimate heat

sink (LUHS) events [6-10]. The

OPR1000 NPPs have also been introduced

the multi-barrier accident

­coping strategy (MACST) to prevent

the severe consequences from

such events. Coping strategies are

established to keep the pressure

boundaries of the RCS, appropriately

cool down the reactor core for

avoiding fuel damage, and maintain

the integrity of the containment

building in the transient conditions

with the events of loss of safety

functions. There are three phases

depending on coping strategies as

shown in Table 2.

In Phase 1, the plant copes with

only the installed equipment, such as

DC batteries and natural circulation

cooldown through SGs by the available

AFWPs up to 8 hours. And then

the cooldown and depressurization of

the RCS are performed by MSSVs

and/or ADVs of the main steam supply

system. In Phase 2, the MACST

facilities are completely connected

and supplement functions of installed

Research and Innovation

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

Phase Time Event

Phase 1

Phase 2

0 sec Initiating event

0 sec Reactor trip

0 sec RCP trip

0 sec Turbine trip

0 sec Start of RCP seal leakages

8 sec MSSVs open

10 min Fail to operate AAC DG

17 min Start of AFWPs operation

30 min Completing load shedding

2 hr Failure of RCS cooldown using ADVs

8 hr Completing connection of HP mobile pump

8 hr 15 min Activating HP mobile pump

8 hr 30 min Preparing connection of 3.2 MW mobile generator

72 hr Completing connection of 3.2 MW mobile generator

| Tab. 3.

Sequence of ELAP event during Phase 1 and Phase 2.

equipment for supplying AC power to

some safety systems or alternative

RCS make-up to avoid core uncovery

and damage. From 72 hours after the

event initiation, the plant enters the

final phase with maintaining Phase 2

strategy and all available on/offsite

facilities are in service.

As stated above, one of the main

concerns of the coping strategy of

the ELAP event is the RCS make-up

strategy due to the continuous inventory

loss by the RCP seal leakage. In

the MACST for the ELAP event, early

cooldown and depressurization of the

RCS should be conducted in Phase 1

for reducing the inventory loss

and providing rapid borated water

injection from the SITs as described in

Section 2. Operators locally control

ADVs for depressurizing the primary

system by releasing the steam.

If the ADV local control and connecting

the mobile generator are

failed in both Phase 1 and 2, the RCS

inventory will continuously decrease

and the RCS make-up by SITs and

CPs will not be available. Therefore,

the back-up strategy for the RCS

make-up should be necessary to avoid

core uncovery and damage. The highpressure

(HP) mobile pump, with a

nominal flow rate of 40 gpm at

1500 psig, will be equipped for primary

inventory make-up and boration

during Phase 2 in case of a failure

of the RCS depressurization, SIT

injection, and CP operation.

4 Feasibility study

on high-pressure

mobile pump for MACST

strategy

In this section, we assume a failure of

ADV local control and connection

of the mobile generator to create a

situation using the HP mobile pump

for RCS make-up during Phase 2. It is

assumed that the HP mobile pump is

successfully connected at 8 hours

after initiating the event and starts

operation in 15 minutes after the

connection. Table 3 shows a major

­sequence of the ELAP event in Phase 1

and Phase 2 to comply with the

MACST strategy.

Transient behavior of pressurizer

pressure during Phase 1 is shown in

Figure 8. The pressurizer pressure decreases

from the beginning of the

event due to the continuous leakage of

the RCS inventory through the RCP

seals, and then is maintained approximately

9.0 MPa (1290 psig) without

opening ADVs at the end of Phase 1.

The pressure of the primary system

does not sufficiently decrease to the

set point of the SIT actuation, so

borated water is not injected to the

RCS.

The RCP seal leak rate for four

RCPs during Phase 1 are shown in

Figure 9. A tendency of the leak rate

coincides with transient behavior of

pressurizer pressure. The leak rate

starts to oscillate between 6 and 9

gpm after 7 hours because the upstream

of the RCP seal becomes two

phases and the core level is reduced to

the cold leg elevation.

The level in the reactor core is

shown in Figure 10. The core

collapsed water level starts to decrease

from 2.5 hours after initiating the

event due to the continuous loss of the

RCS inventory. Although the core

uncovery does not occur during

Phase 1, the continuous loss of the

inventory occur without any RCS

make-up action. At the end of Phase 1

which is the time of connecting and

activating the HP mobile pump, RCS

pressure is kept at 9.0 MPa (1290 psig)

and RCP seal leak rates show an

oscillatory behavior with an average

of 7 gpm. These conditions are consistent

with the design specifications

of the HP mobile pump.

To control reactivity and make up

RCS inventory, Phase 2 strategy using

MACST facilities should be performed

within 8 hours after the event. Figure

11 shows transient behavior of

pressurizer pressure during Phase 1

and Phase 2. In Phase 2, pressurizer

pressure is kept at approximately

RESEARCH AND INNOVATION 433

| Fig. 9.

RCP seal leak rate (Phase 1).

| Fig. 10.

Core collapsed water level (Phase 1).

Research and Innovation

Evaluation of MACST Strategies for Extended Loss of AC Electric Power Event in OPR1000 Nuclear Power Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


atw Vol. 65 (2020) | Issue 8/9 ı August/September

RESEARCH AND INNOVATION 434

| Fig. 11.

Pressurizer pressure (Phase 1 + Phase 2).

9.2 MPa (1320 psig) and the HP mobile

pump continues to function. The

core collapsed water level are shown

in Figure 12. Since the charging rate

(approximately 2.51 kg/s) for RCS

inventory make-up is larger than the

rate of inventory loss (approximately

2.40 kg/s) through RCP seals,

the inventory of RCS is gradually

re covered. When taking into consideration

that the initial flow rate of

the RCP seal leak is conservatively

estimated at 25 gpm, the HP mobile

pump can give the entire satisfaction

as the back-up strategy for RCS

make-up.

5 Conclusion

In this study, we particularly paid

attention to an effect of the RCP seal

leakage causing the continuous loss

of the RCS inventory on the safety capabilities

of the NPP. Two different

cases depending on consideration of

the RCP seal leakage were compared.

The RCP seal leakage had a great

effect on the primary pressure and

inventory conditions. The primary

pressure rapidly decreased at the

beginning of the event when considering

the RCP seal leakage. The

abrupt drop of the core collapsed

water level was also observed from

the early stage due to rapid depletion

of the pressurizer water level. However,

the core uncovery did not occur

and the core collapsed water level

starts to recover after reaching the

actuating pressure of the SITs by early

cooldown and depressurization of the

RCS. The secondary side conditions

also presented no meaningful differences

between two cases when

assuming the leak rate of 25 gpm

(1.58 l/s) per one RCP.

This study also examined the

­feasibility of the MACST strategy for

inventory make-up using the HP

mobile pump having a capacity of

40 gpm at 1500 psig in Phase 2 of the

ELAP event. To verify the strategy, we

assumed failure of RCS cooldown and

depressurization by local control of

ADVs in Phase 1 and a subsequent

failure of the 1 MW mobile generator

in Phase 2. It is confirmed that

the RCS pressure could be kept at

maximum 9.2 MPa (1320 psig) and

the seal leak rate for one RCP was

­approximately 7 gpm at the end of

Phase 1 through the simulation with

RELAP5 code. It is concluded that

these RCS conditions were sufficient

to provide RCS make-up and boration

using the HP mobile pump in Phase 2

and the core water level could be fully

recovered before entering Phase 3.

Acknowledgments

This work was supported by the Korea

Institute of Energy Technology Evaluation

and Planning (KETEP) and the

Ministry of Trade, Industry & Energy

(MOTIE) of the Republic of Korea

(No. 20161510101840).

References

1. S. W. Lee et. al, “Extended Station Black Out Coping

Capabilities of APR1400”, Science and Technology of Nuclear

Installations, Vol. 2014, p. 1-10, 2014.

2. Korea Hydro and Nuclear Power Co. Ltd., “Stress Test Report

for Hanul Unit 3&4”, 2017.

3. Information Systems Laboratories, Inc., “RELAP5/MOD3.3

Code Manual”, 2016.

4. Nuclear Safety and Security Commission, “Stress Test for

Nuclear Power Plants in Long Term Operation”, 2013.

5. J. Hartz et. al, “WCAP-17601-P: Reactor Coolant System

Response to the Extended Loss of AC Power Event for

Westinghouse, Combustion Engineering and Babcock &

Wilcox NSSS Designs, 2012.

6. Nuclear Energy Institute, “Diverse and Flexible Coping

Strategies (FLEX) Implementation Guide”, 2012.

7. D. H. Kim et. Al, “Development of Mitigation Strategy for

Beyond Design Basis External Events for NRC Design

Certification”, Transactions of the Korean Nuclear Society

Autumn Meeting, 2013.

8. United States Nuclear Regulatory Commission, “Byron Station,

Units 1 and 2 – Safety Evaluation Regarding Implementation

of Mitigating Strategies and Reliable Spent Fuel Pool

Instrumentation Related to Orders EA-12-049 and EA-12-051”,

2016.

9. M. M. Rahman and M. B, Shohag, “FLEX Strategy to Cope

with Extended SBO for APR1400”, International Journal of

Engineering Research and Technology (IJERT), Vol. 5, Issue 10,

2016.

10. C. F. Huang et. al, “Analysis of a Postulated ELAP Event in

Maanshan NPP using Trace Code”, TopSafe 2017, 2017.

| Fig. 12.

Core collapsed water level (Phase 1 + Phase 2).

Authors

Bongsik Chu

bongsik.choo@khnp.co.kr

Seyun Kim

Junkyu Song

Minjeong Kim

Chang Hyun Kim

Korea Hydro and Nuclear Power

Co., Ltd.,

Central Research Institute

70, Yuseong-daero

1312-gil, Yuseong-gu

Daejeon, 34101, Korea

Research and Innovation

Evaluation of MACST Strategies for Extended Loss of AC Electric Power Event in OPR1000 Nuclear Power Plants ı Bongsik Chu, Seyun Kim, Junkyu Song, Minjeong Kim and Chang Hyun Kim


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Neutronic Study of CAREM-25

Advanced Small Modular Reactor

Using Monte Carlo Simulation

Saeed Zare Ganjaroodi and Ali Pazirandeh

1 Introduction The trend in development has been towards design certification of small modular reactors,

which are defined as advanced reactors that produce electricity up to 300 MW(e), designed to be built in factories and

shipped to utilities for installation as demand arises. These new factory-built designs aim to reduce lengthy con struction

times while simultaneously increasing quality, thereby minimizing the financing costs associated with nowadays design

projects that span 5–8 years. Small Modular Reactors designs include water-cooled reactors, high-temperature

gas-cooled reactors, as well as liquid metal cooled reactors with fast neutron spectrum. Some of Small Modular Reactors

are to be deployed as multiple-module power plants. Several countries are also pioneering the development and

­application of transportable nuclear power plants, including floating and seabed-based Small Modular Reactors. The

distinct concepts of operations, staffing and security requirements, size of emergency planning zones (EPZs), licensing

process, legal and regulatory framework are the main issues for the Small Modular Reactors deployment. The projected

timelines of readiness for deployment of SMRs generally range from the present to 2025–2030 [1-2].

Central Argentina de Elementos

­Modulares (CAREM-25) is a national

SMR development project based on

LWR technology coordinated by the

Argentina National Atomic Energy

Commission (CNEA) in collaboration

with leading nuclear companies in

­Argentina with the purpose to

develop, design and construct innovative

small nuclear power plants

with high economic competitiveness

and high level of safety. CAREM-25 is

deployed as a prototype to validate

the innovations for the future commercial

version of CAREM that

will generate an electric output of

150-300 MW(e). CAREM-25 is an

integral type PWR based on indirect

steam cycle with distinctive features

that simplify the design and support

the objective of achieving a higher

level of safety. Some of the design

characteristics of CAREM-25 are

integrated primary cooling system,

in-vessel hydraulic control rod drive

mechanisms and safety systems

relying on passive features. Coolant

flow in the primary reactor system

is done by natural circulation.

­CAREM-25 reactor was developed

using domestic technology with at

least 70 % of the components and

­related services required by Argentine

companies [1-3-4].

Although several reports and

essays on various technical aspects of

CAREM-25 small modular reactor

have been studied in recent years but

their Neutronic behavior of the core

in critical situation has not been

discussed in details [3-4-5-6-7-8-9-10-

11-12]. Also, due to the importance of

calculation of the critical parameters

in critical condition, some parameters

such as neutron spectrum, power

peaking factor associated with each

fuel assemblies and the worth of

­control rods are evaluated for the first

time in this paper.

In this work, MCNPX code is used

for the neutronic simulation of the

CAREM-25 reactor core. The reasons

for using this code are different

capabilities of the code to analyze the

Neutronic calculation and to benchmark

the results with references to

show an appropriate consistency.

MCNPX is a general purpose

Monte Carlo radiation transport code

designed to track many particle types

over broad ranges of energies. It can

be used in several transport modes:

neutron only, photon only, electron

only, combined neutron/photon

transport where the photons are

produced by neutron interactions,

neutron/ photon/electron, photon/

electron, or electron/photon [13].

The MCNPX code is using Monte Carlo

method to simulate the geometry and

solve transport equation by tracing

individual particles and recording

some aspects (tallies) of their average

behavior. It does not solve an explicit

transport equation, but rather obtains

answers by simulating individual

particles and recording some tallies of

individual particle average behavior.

Then, the average behavior of the

particles in the physical system is

interpreted as the average behavior of

the simulated particles [14].

This code was used to create some

inputs for the core of CAREM-25

reactor core to calculate and analyze

the Neutronic parameters such as

effective multiplication factor, neutron

flux distribution, axial power

distribution, power peaking factor

related to each fuel assemblies and

the worth of control rods.

2 Materials and Methods

2.1 CAREM-25 Reactor

CAREM-25 has been chosen as the

reference small modular reactor in

this study for Neutronic simulation.

CAREM-25 is an integrated and

self-pressurized reactor that has

primary cooling by natural circulation.

This reactor has some features

that make the reactor extremely

simple and also contribute to a higher

level of safety. CAREM-25 core design

data are shown in Table 1 [1].

The CAREM-25 reactor pressure

vessel (RPV) contains the core, the

steam generators (SG), the whole

primary coolant, and the absorber rod

drive mechanisms. The RPV diameter

is about 3.2 m and the overall length

is about 11 m. CAREM-25 Reactor

­Pressure Vessel is shown in Figure 1

[1-4-5-12].

The core consists of 61 hexagonal

fuel assemblies having 1.4 active

lengths. Each fuel assembly has 108

fuel rods, 18 guide thimbles and 1

instrumentation thimble (Figure 2).

The fuel is enriched Uranium Oxide

[1-15].

CAREM-25 core do not use any

chemical controller such as boric acid.

Core reactivity is controlled by the use

of 8 % gadolinium oxide (Gd 2 O 3 )

mixed with 92 % uranium oxide as

burnable poison in specific fuel rods

in 42 fuel assemblies. Absorbing

­Materials including Silver (80 %),

Cadmium (5 %) and Indium (15 %)

are used into the adjusting and safety

control rods [15-16].

RESEARCH AND INNOVATION 435

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

RESEARCH AND INNOVATION 436

Parameter

Technology developer

Country of origin

Reactor type

Electrical capacity (MW) 27

Value

CNEA

Thermal capacity (MW) 100

Design life (year) 60

Coolant/moderator

Primary circulation

Argentina

System pressure (MPa) (Primary Cycle) 12.25

System pressure (MPa) (Secondary Cycle) 4.7

Main reactivity control mechanism

RPV height (m) 11

RPV diameter (m) 3.2

Coolant temperature, core inlet (°C) 284

Coolant temperature, core outlet (°C) 326

Power conversion process

Passive safety features

Active safety features

Fuel type/assembly array

Fuel rod cladding material

Absorbent pellet

Control rod cladding material

Fuel active length (m) 1.4

Number of fuel assembly 61

Fuel enrichment (%)

Fuel burn-up (GWd/ton)

Fuel cycle (month)

Number of safety trains 2

Emergency safety and Residual heat removal systems

Modules per plant 1

| Tab. 1.

CAREM-25 design parameters.

Integral PWR

Light water

Natural circulation

Only by control rods

Indirect Rankine cycle

Yes

Yes

UO2 pellets/hexagonal

Zry-4

Ag-In-Cd

AISI 316 L

3.1 (prototype)

24 (prototype)

14 (prototype)

Passive

2.2 MCNPX Neutronic

Simulation

MCNPX2.6.0 code is used to model

CAREM-25 core and calculate neutronic

parameters such as effective

multiplication factor, thermal, epithermal

and fast neutron flux distribution,

axial power distribution,

power peaking factor related to each

fuel assembly and the worth of

adjusting control rods in steady state

for fresh fuel. In this code, the KCODE

card is used for critical source calculations,

and 1 million particles with

200 cycles were considered. The axial

coolant temperature changes from

284 (°C) in the core inlet to approximately

326(°C) in the core outlet.

In the calculation, the temperature

of water is considered 305 (°C) as

average temperature via using the

ENDF/B-VI library and 42.C, 51.C,

52.C and 70.C identification databases

codes in MCNPX input file.

3 Results and Discussion

In this work, the neutronic evaluation

of the CAREM- 25 advance small

modular reactor is discussed using

MCNPX code simulation. According to

the calculation of Neutronic parameters

in critical condition, in addition

to analyzing the core in critical

situation, some parameters including,

neutron flux distributions, neutron

spectrum, power distributions, power

peaking factor related to each fuel

assemblies and the worth of control

rods have been discussed for the first

time in this paper. It should be noted

that simulation is done according to

the design criteria in the latest reports

from Argentina.

| Fig. 1.

CAREM-25 reactor pressure vessel and fuel assembly diagram.

Control rods are classified into

several groups as adjusting and safety

control rods. Adjusting control rods are

applied in 19 fuel assembly in ­specific

rods. Safety rods are applied just into

6 fuel assembly. Control ­systems are

used to reactivity control during normal

operation and to produce a sudden

interruption of the nuclear chain reaction

when required [11-12-15-16].

3.1 Criticality calculations

The effective multiplication factor

(k eff ) is calculated 1.04576 with

0.00028 percent error (excess reactivity

is equal to 43.75 (mk)) in

­MCNPX code when 10 % of the adjust

control rods are into the core. Comparison

of results shows an appropriate

consistency with reference. The

small difference in the results is due

to the difference in the use of the

percentage of gadolinium oxide in

the fuel mixture in the simulation by

MCNPX code.

Given that the CAREM-25 reactor

has not been constructed yet, Reactor

design parameters according to the

latest reports indicate that use of 8 %

gadolinium oxide mixed with 92 %

uranium oxide as burnable poison

in specific fuel rods in 42 fuel

assemblies. However, in some reports

there has been a ratio of 7.5 % to

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| Fig. 2.

The configuration of different types of fuel assemblies into the core.

92.5 % for the amount of gadolinium

oxide and fuel [17]. Therefore, due to

the high neutron absorption cross

section of gadolinium, the greater

amount used in this simulation

­reduces the neutron flux and the

effective multiplication factor consequently.

Due to the lack of poisons ( 135 Xe

and 149 Sm) in the beginning of the

reactor operation for archiving

criticality state, both adjust and safety

control rods should be inserted into

the reactor core. By inserting 50 % of

adjusting control rods and 10 % of

safety control rods into the core the

reactor will achieve to critical condition.

3.2 Axial and radial flux

distributions

The axial flux distribution in the

­CAREM-25 small modular reactor

core has been shown in Figure 3.

It should be noted that the energy

intervals are selected according to the

reference from Argentina [19].

The maximum thermal flux is

­related about 57 (cm) from the height

of the core. The inserting of the

control rods into the core can move

the maximum flux height of about

40 % to the end. In Figures 4, 5 and 6

the radial thermal, epithermal and

fast flux distributions are shown in

critical condition using cubic mesh in

MCNPX code.

As the figures show the neutron

flux distribution in this reactor has a

one-thirds symmetry. The reason of

flux drop in some zones is the ­entrance

of control rods and gadolinium oxide

with the high absorption crosssection.

Calculations show flux drop is in

some zones that the control rods and

gadolinium oxide with the high

absorption cross-section are inserted

into the core. Also, the figures

illustrate that the thermal and total

flux decrease due to the insertion of

adjusting control rods (in 19 fuel

assemblies) into the core during the

| Fig. 4.

The radial thermal flux distribution per one neutron in critical condition.

| Fig. 5.

The radial epithermal flux distribution per one neutron in critical condition.

| Fig. 6.

The radial fast flux distribution per one neutron in critical condition.

| Fig. 3.

Axial flux distribution in critical condition.

expected operating transient. The

­remarkable decreases of thermal flux

have occurred in fuel assemblies,

which includes control rods and

RESEARCH AND INNOVATION 437

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RESEARCH AND INNOVATION 438

| Fig. 7.

Neutron spectrum in the CAREM-25 reactor core.

gadolinium oxide. The insertion of the

adjusting control rod clusters into the

core changes thermal flux peaks with

respect to both of the place and the

magnitude. By inserting 50 and

100 percent of control rods into the

core, the central maximum thermal

flux decrease remarkably.

The neutron spectrum in the

­CAREM-25 small modular reactor

core by MCNPX code has been shown

in Figure 7.

Figures 8 and 9 illustrate the effect

of gadolinium oxide mixed with

­uranium oxide as fuel on fission and

absorption cross-section in the core.

According to Figures 8 and 9

gadolinium oxide with high absorption

cross-section can effect on

the fission rate in the core as a burnable

poison. Gadolinium oxide mixed

with uranium oxide as fuel can reduce

the fission cross-section in order of ten

times considerably.

3.3 Worth of adjust control

rods

From the theoretical point of view, the

integral worth of control rod is the

total reactivity along the control rod.

On the other hand, the differential

worth is the reactivity in each unit of

the length of the control rod. In Figure

10, the integral and differential worth

of adjusting control rods are plotted in

of the core. According to the MCNPX

code calculation, the total reactivity

value of the adjusting control rods is

almost 130.74 mk.

| Fig. 8.

UO 2 (3.1%) neutron cross-section.

3.4 Power Peaking Factor

Calculation

Power Peaking Factor of each fuel

assembly indicates a factor that determines

the amount of power produced

in each fuel assembly. This coefficient

is defined as follows:

| Fig. 9.

UO 2 (3.1%) + Gd 2 O 3 neutron cross-section.

| Fig. 10.

Integral and differential worth of adjust control rods.

The axial power and power peaking

factor (PPF) distribution in the core by

MCNPX code has been shown in

Figures 11.

The maximum power distribution

in the core is related about 57 (cm)

from the height of the core. The

inserting of the control rods into the

core can move the maximum axial

power height of about 40 % to the

end. It should be noted that axial flux

and power distribution will reach the

maximum value at the same axial

height.

Calculated Power Peaking Factor

(PPF) of each fuel assembly by

MCNPX code has been demonstrated

in Figure 12.

The power peaking factors of the

FAs are calculated using the MCNPX

code. Since the coolant temperature

and density considerably affect the

neutron moderation and fission

cross-section of materials, the height

of the core is divided into 10 equally

spaced zones to calculate the power in

each zone.

The maximum calculated power

peaking factor for fuel assembly is

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

| Fig. 11.

Axial power distribution in the core.

| Fig. 12.

Power Peaking Factor (PPF) of fuel assemblies.

calculated 1.526 in the hottest fuel

assembly. The maximum power

peaking factors are related to central

fuel assemblies and some fuel assemblies

consist of 3.1% enrichment

fuel without any gadolinium oxide.

4 Conclusions

The purpose of this study is the

Neutronic evaluation of the generation

IV reactors CAREM-25 small

modular reactor using Mote carlo

simulation. Regarding some unique

feathers, CAREM-25 is chosen as the

reference small modular reactor in

this study for simulation. CAREM-25

is an integrated and self-pressurized

reactor that has primary cooling by

natural circulation. This reactor has

some features that greatly make the

reactor simply the reactor and also

contribute to a higher level of safety.

Considering the importance of calculation

of the parameters in critical

condition, and the reactor has not

constructed yet, Neutronic simulation

is done according to the latest design

parameters from Argentina. Then, in

addition to the Neutronic evaluation,

some parameters such as neutron

spectrum, power peaking factor

related to each fuel assemblies and

the worth of control rods discussed

for the first time in this study. Results

show that the reactor core has

­approximately 43.75 (mk) excess

­reactivity by inserting 10 % of adjusting

control rods into the core in

MCNPX code. The reason for this high

excess reactivity of the core is the long

cycle of this reactor over 14 months.

The excess reactivity of fresh fuel

is appropriately compensated using

gadolinium oxide as a burnable poison

mixed with uranium dioxide in some

fuel rods in 24 fuel assemblies. Due to

the lack of poisons ( 135 Xe and 149 Sm)

in the beginning of the reactor

operation for archiving criticality

state, both adjust and safety control

rods should be inserted into the

­reactor core. By inserting 50 % of

­adjusting control rods and 10 % of

safety control rods into the core the

CAREM-25 reactor core will reach to

critical condition. The inserting of the

control rods into the core can move

the maximum flux height of about

40% to the end. Also, near control

rods and burnable poisons regions

due to the absorbers materials, the

neutron flux dropped sharply. It

should be noted that the total

reactivity value of the adjust control

rods is almost 130.74 (mk). The

maximum calculated axial power

peaking factor for fuel assemblies is

1.526 in the hottest fuel assembly. The

maximum axial power peaking factor

is related to central fuel assemblies

and some fuel assemblies consist

of 3.1 % enrichment fuel without

gadolinium oxide.

References

[1] International Atomic Energy Agency, 2014. Advances in

Small Modular Reactor Technology Developments, A

Supplement to: IAEA Advanced Reactors Information System

(ARIS). IAEA, Vienna.

[2] International Atomic Energy Agency, 2006. Status of

Innovative Small and Medium Sized Reactor Designs 2005

Reactors with Conventional Refueling Schemes.

IAEATECDOC-1485. IAEA, Vienna.

[3] Ishida, M., 2000. Development of New Nuclear Power Plant

in Argentina, Advisory Group Meeting on Optimizing

Technology, Safety and Economics of Water Cooled Reactors

(Vienna, Austria).

[4] Ishida, M., et al., 2001. CAREM Project Development

Activities”. International Seminar on Status and Prospects for

Small and Medium Size Reactors (Cairo, Egypt).

[5] CNEA & INVAP, 2000. CAREM-25-informe Consolidado.

[6] Gomez, S., 2000. Development Activities on Advanced LWR

Designs in Argentina, Technical Committee Meeting on

Performance of Operating and Advanced Light Water Reactor

Designs (Munich, Germany).

[7] Delmastro, D., 2000. Thermal-hydraulic Aspects of CAREM

Reactor, IAEA TCM on Natural Circulation Data and Methods

for Innovative Nuclear Power Plant Design (Vienna, Austria).

[8] Delmastro, D., Mazzi, R., Santecchia, A., Ishida, V., Gomez, S.,

Gomez de Soler, S., Ramilo, L., 2002. CAREM: An Advanced

Integrated PWR’. In IAEA, Small and Medium Sized Reactors:

Status and Prospects. IAEA-CSP-14/P, pp. 224e231.

[9] Mazzi, R., Santecchia, A., Ishida, V., Delmastro, D., Gomez, S.,

Gomez de Soler, S., Ramilo, L., 2002. CAREM Project

Development. In IAEA, Small and Medium Sized Reactors:

Status and Prospects. IAEA-CSP-14/P, pp. 232e243.

[10] Reyes, J., 2005. Integral System Experiment Scaling

Methodology. Annex 11, Natural Circulation in Water Cooled

Nuclear Power Plants Phenomena, Models, and

Methodology for System Reliability Assessments.

IAEA TECDOC 1474.

[11] Boado Magan, H., Delmastro, D.F., Markiewicz, M., Lopasso,

E., Diez, F., Gimenez, M., Rauschert, A., Halpert, S., Chocron,

M., Dezzutti, J.C., Pirani, H., Balbi, C., Fittipaldi, A., Schlamp,

M., Murmis, G.M., Lis, H., 2011. CAREM Project Status,

Science and Technology of Nuclear Installations. Article ID

140373.

[12] Boado Magana, H., Delmastrob, D.F., Markiewiczb, M.,

Lopassob, E., Diez, F., Gim_enezb, M., Rauschertb, A.,

Halperta, S., Chocr_onc, M., Dezzuttic, J.C., Pirani, H.,

Balbi, C., 2012. CAREM Prototype Construction and Licensing

Status. IAEA-CN-164e5S01.

[13] Pelowitz, D.B., 2008. MCNPXTM User’s Manual Version 2.6.0.

LOS ALAMOS NATIONAL LABORATORY.

[14] Liu, B., Lv, X., Zhao, W., Wang, K., Tu, J., Ouyang, X., 2010.

The comparison of MCNP perturbation technique with MCNP

difference method in critical calculation. Nucl. Eng. Des. 240,

2005e2010.

[15] Villarino, E., Hergenreder, D., Matzkin, S., 2012. Neutronic-

Core Performance of CAREM-25 Reactor. INVAP, Argentina.

[16] Diego Ferraro, 2009. Calculo de exposicion de estructuras

interiores recipinte de presion del CAREM-25 mediante

MCNP. Instituto Balseiro Universidad Nacional de Cuyo

Comision Nacional de Energia Atomica. San Carlod de

Bariloche Argentina.

[17] S. Tashakor, E. Zarifi, M. Naminazari. 2017. Neutronic

simulation of CAREM-25 small modular reactor. Progress in

Nuclear Energy 99 (2017) 185e195.

Authors

Saeed Zare Ganjaroodi

Ali Pazirandeh

Islamic Azad University

Engineering Science and Research

Branch

Shodada Hesarak blvd,

Daneshgah Square,

Sattari Highway,

Tehran, Iran

RESEARCH AND INNOVATION 439

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

440

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September 2020

30 Jahre | 1990

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01. Dr.-Ing. Hans-Georg Willschütz, Springe

Oktober 2020

25 Jahre | 1995

16. Jonathan C. J. Schade, Krefeld

35 Jahre | 1985

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Operating results 2019

In 2019 the seven (7) German nuclear power plants generated 75.10

billion kilowatt hours (kWh) of electricity gross. At the end of 2019

the Philippsburg 2 nuclear power plant ceased commercial operation

due to the revision of the German Atomic Energy Act in the political

aftermath of the accidents in Fukushima, Japan, in 2011. Seven

nuclear power plants with an electric gross output of 10,013 MWe

were in operation during the year 2019.

All seven nuclear power plants in operation in 2019 achieved

results with a gross production greater than 10 billion kilowatt hours,

one power plant, The Isar 2 unit even produced more than 12 billion

kilowatt hours.

Additionally the Isar 2 unit achieved one of the world’s ten best

production results in 2019 (“Top Ten”, sixth place). At the end of

2019, 449 reactor units were in operation in 31 countries worldwide

and 54 were under construction in 16 countries. The share of nuclear

power in world electricity production was around 11 %. German

nuclear power plants have been occupying top spots in electricity

production for decades thus providing an impressive demonstration

of their efficiency, availability and reliability.

The Taishan-1 nuclear power plant in China (capacity: 1,750 MWe

gross, 1,660 MWe net, reactor type: EPR, the most powerful nuclear

power plant worldwide and the most powerful single power plant

worldwide) achieved the world record in electricity production in

2019 with appr. 13 billion kilowatt hours.

Worldwide, 41 nuclear power plant units achieved production

­results of more than 10 billion kilowatt hours net in the year 2019.

Additionally German nuclear power plants are leading with their

lifetime electricity production. The Brokdorf, Emsland, Grohnde,

Isar 2 and Philipsburg 2 nuclear power plant have produced more

than 350 billion kilowatt hours since their first criticality.

441

REPORT

Operating results of nuclear power plants in Germany 2018 and 2019

Nuclear power plant Rated power Gross electricity

generation

in MWh

Availability

factor*

in %

Energy availability

factor**

in %

gross

in MWe

net

in MWe

2018 2019 2018 2019 2018 2019

Brokdorf KBR 1,480 1,410 10,375,751 10,153,213 90.60 87.69 84.72 82.34

Emsland KKE 1,406 1,335 11,495,686 10,781,232 94.78 89.20 94.67 89.12

Grohnde KWG 1,430 1,360 10,946,635 10,700,632 92.82 90.10 91.61 89.80

Gundremmingen KRB C 1,344 1,288 10,361,862 10,381,798 90.41 89.20 89.85 88.50

Isar KKI 2 1,485 1,410 12,127,490 12,036,656 95.46 95.95 95.24 95.68

Neckarwestheim GKN II 1,400 1,310 9,703,700 10,411,410 81.35 94.03 81.00 88.00

Philippsburg KKP 2 1,468 1,402 10,993,639 10,606,307 90.63 89.63 90.47 89.31

Total 10,013 9,515 76,004,763 75,071,247 90.85 90.82 89.60 88.86

* Availability factor (time availability factor) kt = tN/tV: The time availability factor kt

is the quotient of available time of a plant (tV) and the reference period (tN).

The time availability factor is a degree for the deployability of a power plant.

** Energy availability factor kW = WV/WN: The energy availability factor kW is the quotient of available

energy of a plant (WV ) and the nominal energy (WN). The nominal energy WN is the product

of nominal capacity and reference period. This variable is used as a reference variable (100 % value)

for availability considerations. The available energy WV is the energy which can be generated

in the reference period due to the technical and operational condition of the plant.

Energy availability factors in excess of 100 % are thus impossible, as opposed to energy utilisation.

*** Inclusive of round up/down, rated power in 2019.

**** The Gundremmingen nuclear power plant (KRB B) was permanently shutdown on 31 December 2017

due to the revision of the German Atomic Energy Act in 2011.

All data in this report as of 31 March 2020

Report

Operating results 2019


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442

Brokdorf

REPORT

Operating sequence in 2019

100

90

80

70

60

50

40

30

20

10

Electrical output in %

January February March April May June July August September October November December

In 2019, the Brokdorf nuclear power plant (KBR) was connected to p Transformers

the grid for a total of 7,682 operating hours with an availability factor replacement

of 82.3 %. The gross generation for the year under review was p Building ZB.9

10,153,212 MWh. In 2019, the thermal reactor output was again

­limited to a maximum of 95 % with a coolant temperature reduced

by 3 K due to the specifications of ME 02/2017 “Increased oxide layer

thickness on fuel rod cladding tubes of fuel elements”.

On 13 April 2019, the plant was disconnected from the grid by

triggering “Manual-TUSA” due to increasing vibrations at the turbine-­

generator set. After completion of the inspection work, the unit was

0

reconnected to the grid on 22 April 2019.

In the period from 17 to 22 November 2019, the plant was shut

down to the “subcritical cold” state due to several findings on the

100

­baffle in the cooling water return structure.

100

80

60

40

20

80

Planned shutdowns

On 7 June 2019, the plant was shut down for the 31 st refuelling and

60

annual inspection.

The inspection and maintenance included the following priorities:

40

p Reactor

Full core discharge

Replacement of 60 fresh fuel elements

20

Inspection of fuel elements,

0

control elements, throttle bodies.

p Containment Leak rate test.

p Main coolant Ring exchange electric motor

pump YD30 Inspection of axial bearings

Replacing the mechanical seal

Positionierung: (axial bearing).

p Main Bezug, coolant links, Inspection untenof the axial bearing.

pump YD40

p Steam generator WS test Steam generator 10/20,

VGB: HKS6K additionally 30 % 30/40.

p Main atw: steam 100 safety 60 Internal 0 0 inspections.

and relief valves station

p Cooling water Work in the pump antechambers of the

secondary cooling water systems

VE10/20 and 30/40

Work in the main cooling water

channels VA40-60.

p Turbine/

Control Inspection / Generator

Generator ND II + III, run-out measurements,

Displacement measurements

Inspection bearing SB14.

X = 20,475 Y = 95,25 B = 173,5 H = 38,2

Exchange CS32 (emergency power supply),

Exchange CS41 (normal power supply).

Fire protection upgrading overflow flaps.

The grid synchronisation took place on 9 July 2019 at 08:17 h after

31.1 days.

Compared to planning, the start-up date was delayed by 4.6 days.

The inspection extension is mainly due to the additional inspections

of the heating tube plugs on steam generators 30 and 40.

Unplanned shutdowns and reactor/turbine trip

On 13 April 2019, the turbine was disconnected from the grid by

triggering “manual turbine trip” due to increasing vibrations at the

turbine-generator set. After completion of the inspection work, the

unit was reconnected with the grid on 22 April 2019.

In the period from 17 to 22 November 2019, the plant was shut

down to “subcritical cold” state due to several findings on the baffle

in the cooling water return structure.

Power reductions above 10 % and longer than for 24 h

Load reductions for the implementation of the grid-supporting

power control as well as redispatch on demand of the control centre

were carried out.

WANO Review/Technical Support Mission

The WANO Peer Review 2019 was conducted in the period 15 to

26 July 2019 in the form of an “Optimized Peer Review”, i.e.

­shortened to 2 weeks. In the run-up to the WANO Peer Review 2019,

a Crew Performance Observation (CPO) was successfully completed

for the first time by the KBR shift team on the simulator of a German

nuclear power plant with pressurised water reactor in the period

from 6 to 9 May 2019.

In summary, WANO confirmed a very good result for KBR.

Delivery of fuel elements

During the reporting year 28 fuel elements were delivered and

stored.

Waste management status

By the end of the year 2019, 33 loaded CASTOR © cask were located

at the Brokdorf on-site intermediate storage.

Report

Operating results 2019


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Operating data

Review period 2019

REPORT

Plant operator: PreussenElektra GmbH

Shareholder/Owner: PreussenElektra GmbH (80 %),

Vattenfall Europe Nuclear Energy GmbH (20 %)

Plant name: Kernkraftwerk Brokdorf (KBR)

Address: PreussenElektra GmbH, Kernkraftwerk Brokdorf,

25576 Brokdorf, Germany

Phone: +49 4829 752560

Web: www.preussenelektra.de

100

90

80

70

84

Availability factor in %

Capacity factor in %

92

93

93

93

90

78

First synchronisation: 10-14-1986

Date of commercial operation: 12-22-1986

Design electrical rating (gross):

1,480 MW

Design electrical rating (net):

1,410 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

60

50

40

44

The following operating results were achieved:

Operating period, reactor:

7,682 h

Gross electrical energy generated in 2019:

10,153,212 MWh

Net electrical energy generated in 2019:

9,635,834 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2019:

360,721,021 MWh

Net electrical energy generated since

first synchronisation until 12-31-2019:

342,884,965 MWh

Availability factor in 2019: 87.69 %

Availability factor since

date of commercial operation: 89.78 %

Capacity factor 2019: 82.34 %

Capacity factor since

date of commercial operation: 85.94 %

Downtime

(schedule and forced) in 2019: 12.31 %

Number of reactor scrams 2019: 0

30

20

10

0

10

9

8

84

2012

93

2013

93

2014

93

2015

93

2016

52

2017

Collective radiation dose of own

and outside personnel in Sv

91

2018

88

2019

Licensed annual emission limits in 2019:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.0 · 10 15 Bq

6.0 · 10 9 Bq

5.55 · 10 10 Bq

Proportion of licensed annual emission limits

for radioactive materials in 2019 for:

Emission of noble gases with plant exhaust air: 0.075 %

Emission of iodine-131 with plant exhaust air: 0.087 %

Emission of nuclear fission and activation products

with plant waste water (excluding tritium): 0.0005 %

Collective dose:

0.158 Sv

7

6

5

4

3

2

1

0

0.13

2012

0.22

2013

0.17

2014

0.14

2015

0.14

2016

0.13

2017

0.14 0.16

2018 2019

Report

Operating results 2019


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444

Emsland

REPORT

Operating sequence in 2019

Electrical output in %

100

90

80

70

60

50

40

30

20

10

0

January February March April May June July August September October November December

Apart from the 39.4 days refuelling outage the Emsland nuclear

power plant (KKE) had been operating uninterrupted and mainly at

full load during the review period 2019. Producing a gross power

generation of 10,781,232 MWh with a capacity factor of 89.12 % the

power plant achieved a very good operating result.

Planned shutdowns

32 rd refuelling and 31 rd overall maintenance outage.

The annual outage was scheduled for the period 17 May to 26 June.

The outage took 39.4 days from breaker to breaker including an

outage prolongation (18 days) due to replacement of the generator.

In addition to the replacement of 40 fuel elements the following

100

major maintenance and inspection activities were carried out:

80

p Inspection of core and reactor pressure vessel internals.

p Inspection of a reactor coolant pump.

60

p Inspection of pressurizer valves.

p Eddy current test on steam generator tubes.

40

p Containment leak rate test.

p Pressure test on different coolers and tanks.

20

p Inspection on main condensate pump.

p0

Maintenance works on different transformers.

p Different automatic non-destructive examinations.

Unplanned shutdowns and reactor/turbine trip

None.

Power reductions above 10 % and longer than for 24 h

16 April to 17 May: 17 st Stretch-out operation.

Peer Reviews

From 9 September 2019 to 20 September 2019 a WANO team of

10 experienced nuclear professionals from 7 different countries,

conducted an optimized Peer Review (PR) at KKE NPP.

In summary, the WANO exit report concludes that KKE completed

the 2019 Peer Review with a very good result.

Delivery of fuel elements

28 Uranium-fuel elements were delivered.

Waste management status

No CASTOR © cask loading was carried out during the review period

2019.

At the end of the year 47 loaded casks were stored in the local

interim storage facility.

Positionierung:

Bezug, links, unten

X = 20,475 Y = 95,25 B = 173,5 H = 38,2

VGB: HKS6K 30 %

atw: 100 60 0 0

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Operating data

Review period 2019

REPORT

Plant operator: Kernkraftwerke Lippe-Ems GmbH

Shareholder/Owner: RWE Power AG (87,5 %),

PreussenElektra GmbH (12,5 %)

Plant name: Kernkraftwerk Emsland (KKE)

Address: Kernkraftwerk Emsland,

Am Hilgenberg, 49811 Lingen, Germany

Phone: +49 591 806-1612

Web: www.rwe.com

100

90

80

95

Availability factor in %

Capacity factor in %

95

95

91

94

93

95

89

70

First synchronisation: 04-19-1988

Date of commercial operation: 06-20-1988

Design electrical rating (gross):

1,406 MW

Design electrical rating (net):

1,335 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

60

50

40

The following operating results were achieved:

Operating period, reactor:

7,821 h

Gross electrical energy generated in 2019:

10,781,232 MWh

Net electrical energy generated in 2019:

10,237,093 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2019:

357,600,201 MWh

Net electrical energy generated since

first synchronisation until 12-31-2019:

339,066,997 MWh

Availability factor in 2019: 89.20 %

Availability factor since

date of commercial operation: 93.91 %

Capacity factor 2019: 89.12 %

Capacity factor since

date of commercial operation: 93.77 %

Downtime

(schedule and forced) in 2019: 10.80 %

Number of reactor scrams 2019: 0

30

20

10

0

10

9

8

95

2012

95

2013

95

2014

91

2015

94

2016

93

2017

Collective radiation dose of own

and outside personnel in Sv

95

2018

89

2019

Licensed annual emission limits in 2019:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

(incl. H-3 and C-14)

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.0 · 10 15 Bq

5.0 · 10 9 Bq

3.7 · 10 10 Bq

7

6

5

Proportion of licensed annual emission limits

for radioactive materials in 2019 for:

Emission of noble gases with plant exhaust air: 0.016 %

Emission of iodine-131 with plant exhaust air: 0.0 %

(incl. H-3 and C-14)

Emission of nuclear fission and activation products

with plant waste water (excluding tritium): 0.00 %

Collective dose:

0.067 Sv

4

3

2

1

0

0.09

2012

0.08

2013

0.06

2014

0.10

2015

0.05

2016

0.09

2017

0.06 0.07

2018 2019

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Grohnde

REPORT

Operating sequence in 2019

100

90

80

70

60

50

40

30

20

10

Electrical output in %

January February March April May June July August September October November December

During the 2019 reporting year, the Grohnde nuclear power plant

was scheduled for a 36-day overhaul with refueling from grif and

80

achieved a time availability of 90.1 %. Gross generation amounted to

10,700,632 MWh.

60

Compared to the planned 26 days, the inspection was extended

by 250 hours. Delays in the inspection of the drive rods, the

40

­replacement of the LVD lance G10, a conspicuous closing behaviour

of the DH spray valve YP10 S233 and contamination of the turbine oil

20

were the main reasons for the revision delay.

Due to a high Weser water temperature, the output was reduced

0

to 447 MW on June 26 and then increased again to full capacity.

According to the specifications of the load distribution system,

26 load reductions were made in 2019 over a total of 228 hours and

100

156 mains and 70 primary controls for a total of 4,307 hours.

100

After completion of the work on bracing the RPV head and closing

the primary circuit (RKL hydraulically sealed), a limited availability

of a core instrumentation lance (internal neutron measuring system

for the determination and monitoring of the power distribution

density) and 3 Fuel element exit temperature measurements, caused

by a defective plug connection, were determined. The defective core

instrumentation lance was replaced.

Unplanned shutdowns and reactor/turbine trip

None.

Power reductions above 10 % and longer than for 24 h

In the months January, February, April, October and December load

following operation due to requirements of the load distributor.

80

Planned shutdowns

21 April to 27 May: 36 th refuelling and plant inspection with

60

maintenance.

As planned, the Grohnde nuclear power plant was shut down on

40

April 21 for revision and the 36 th refuelling. The main planned work

of the inspection and maintenance were:

20

p Unloading and loading with the insertion

0

of 52 fresh fuel elements.

p Full inspection on 19 fuel elements.

p Eddy current testing on 32 control rods.

p Visual inspection on 15 throttle bodies.

p YD10 D001 Overhaul axial bearing.

p YD30 Positionierung:

D001 Motor conversion.

p YD20 Bezug, D001 Replacing links, unten the mechanical seal.

p YD40 D001 Replacing the mechanical seal &WS test

of the pump shaft.

p Start-up VGB: test HKS6K of the BE 30 centring % pins of the UKG and the OKG.

p VA01 atw: + VA02 100 Inspection 60 0 0and cleaning of cooling water sections.

p TF20 B001 Cleaning the nuclear intercooler.

p TF20 S013/S014 Replacing the screws on the quick-closing flaps.

p TH20 Inspection of secondary shut-offs with pipe freezing.

p Work and inspections in the redundancies with the main focus

of activities in the main redundancy 2/6 (maintenance work on

valves and actuators and tests on tanks, batteries and

electrotechnical branches).

X = 20,475 Y = 95,25 B = 173,5 H = 38,2

Delivery of fuel elements

In March 2019 the delivery of 44 U-/U-Gd fuel elements of the

company Westinghouse took place.

Waste management status

No CASTOR © V/19 containers were loaded in 2019.

The interim storage facility with 34 stored CASTOR © V/19 casks

was handed over to Bundesgesellschaft für Zwischenlagerung mbH

(BGZ).

General points/management systems

In September 2019, the surveillance audit of the quality ­management

system (ISO 9001) and the recertification of the environmental

management system (ISO 14001) and the occupational health

and safety management system (OHSAS 18001) were successfully

completed.

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Operating data

Review period 2019

REPORT

Plant operator: Gemeinschaftskernkraftwerk Grohnde GmbH & Co. OHG

Shareholder/Owner: PreussenElektra GmbH (83,3 %),

Stadtwerke Bielefeld (16,7 %)

Plant name: Kernkraftwerk Grohnde (KWG)

Address: Gemeinschaftskernkraftwerk Grohnde GmbH & Co. OHG,

P.O. bx 12 30, 31857 Emmerthal, Germany

Phone: +49 5155 67-1

E-mail: kwg-kraftwerksleitung@preussenelektra.de

Web: www.preussenelektra.de

100

90

80

70

95

Availability factor in %

Capacity factor in %

89

84

89

73

82

92

90

First synchronisation: 09-05-1984

Date of commercial operation: 02-01-1985

Design electrical rating (gross):

1,430 MW

Design electrical rating (net):

1,360 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

60

50

40

The following operating results were achieved:

Operating period, reactor:

7,889 h

Gross electrical energy generated in 2019:

10,700,632 MWh

Net electrical energy generated in 2019:

10,113,330 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2019:

388,274,835 MWh

Net electrical energy generated since

first synchronisation until 12-31-2019:

367,082,606 MWh

Availability factor in 2019: 90.10 %

Availability factor since

date of commercial operation: 91.70 %

Capacity factor 2019: 89.80 %

Capacity factor since

date of commercial operation: 91.30 %

Downtime

(schedule and forced) in 2019: 9.90 %

Number of reactor scrams 2019: 0

30

20

10

0

10

9

8

95

2012

90

2013

84

2014

89

2015

75

2016

86

2017

Collective radiation dose of own

and outside personnel in Sv

93

2018

90

2019

Licensed annual emission limits in 2019:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

9.0 · 10 14 Bq

7.5 · 10 9 Bq

5.55 · 10 10 Bq

Proportion of licensed annual emission limits

for radioactive materials in 2019 for:

Emission of noble gases with plant exhaust air: 0.019 %

Emission of iodine-131 with plant exhaust air: 0.000 %

Emission of nuclear fission and activation products

with plant waste water (excluding tritium): 0.000 %

Collective dose:

0.261 Sv

7

6

5

4

3

2

1

0

0.27

2012

0.54

2013

0.25

2014

0.31

2015

0.52

2016

0.23

2017

0.12 0.26

2018 2019

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Gundremmingen C

REPORT

Operating sequence in 2019

Electrical output in %

100

90

80

70

60

50

40

30

20

10

0

January February March April May June July August September October November December

In the review year 2019, unit C of Gundremmingen (KRB C) nuclear

power plant was operated at full load without any major restrictions

except for one planned outage for refuelling.

From 4 April to 21 April unit C was in stretch out operation.

During the shutdown a total of 152 fuel elements were unloaded and

replaced with 112 fresh and 40 (2 MOX) partially spent fuel elements.

During the outage all safety relevant workings were monitored

by the relevant nuclear controlling authority, the Bavarian State

­Ministry of the Environment and Consumer Protection (StMUV), and

consulted authorized experts. The inspection of the technical systems

with regard to safety and reliability confirmed the excellent ­condition

of the plant.

A gross total of 10,381,798 MWh of electricity was produced.

100

Planned shutdowns

80

21 April to 29 May 2019: 33 th refuelling outage and 21 th overall

maintenance inspection.

60

The following major activities were carried out:

p Refuelling and sipping of all fuel elements inside the core; result:

40

two defective fuel elements.

p20

Visual inspection and non-destructive testing of reactor pressure

vessel stud bolts and internals.

p0

Inspection of main isolation valves of main steam and safety and

relief valves.

p Emptying of redundancies 1 and 3 for preventive measures on

valves and tanks.

p Inner inspection and pressure tests on high pressure preheater

strings Positionierung:

and reactor water clean-up system.

p Extensive Bezug, non-destructive links, unten testing of pipes and tanks.

p Inspection of two emergency diesel generators.

p Precautionary replacement of 10 kV power cables.

X = 20,475 Y = 95,25 B = 173,5 H = 38,2

VGB: HKS6K 30 %

atw: 100 60 0 0

Unplanned shutdowns and reactor/turbine trip

29 to 31 May: Manual scram due to safety valve staying in open

position during a period test, subsequently exchange of one of his

pilot valves.

Power reductions above 10 % and longer than for 24 h

24 to 25 February: Periodic tests.

4 to 21 April: Stretch-out-operation.

4 to 5 August: period tests and maintenance work.

1 to 4 December: period tests, change of the control rod traversing

order , leak detection in turbine condenser and maintenance work.

Delivery of fuel elements

In 2019, no fresh fuel elements were delivered.

Waste management status

In 2019, a total of 9 CASTOR © casks were loaded. Thus, at the end of

2019, 69 CASTOR © casks with each 52 spent fuel elements out of

units B and C are stored in the local interim storage.

General points

In the year 2019, the recertification of the environmental ­ management

system (ISO 14001) and energy management system (ISO

50001) were successfully carried out.

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Operating data

Review period 2019

REPORT

Plant operator: Kernkraftwerk Gundremmingen GmbH

Shareholder/Owner: RWE Power AG (75 %),

PreussenElektra GmbH (25 %)

Plant name: Kernkraftwerk Gundremmingen C (KRB C)

Address: Kernkraftwerk Gundremmingen GmbH,

Dr.-August-Weckesser-Straße 1, 89355 Gundremmingen, Germany

Phone: +49 8224 78-1

E-mail: kontakt@kkw-gundremmingen.de

Web: www.kkw-gundremmingen.de

100

90

80

70

91

Availability factor in %

Capacity factor in %

89

90

90

86

86

90

89

First synchronisation: 11-02-1984

Date of commercial operation: 01-18-1985

Design electrical rating (gross):

1,344 MW

Design electrical rating (net):

1,288 MW

Reactor type:

BWR

Supplier:

Siemens/KWU,

Hochtief

The following operating results were achieved:

Operating period, reactor:

7,810 h

Gross electrical energy generated in 2019:

10,381,798 MWh

Net electrical energy generated in 2019:

9,900,234 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2019:

341,323,552 MWh

Net electrical energy generated since

first synchronisation until 12-31-2019:

325,082,303 MWh

Availability factor in 2019: 89.20 %

Availability factor since

date of commercial operation: 89.20 %

Capacity factor 2019: 88.50 %

Capacity factor since

date of commercial operation: 87.60 %

Downtime

(schedule and forced) in 2019: 10.80 %

Number of reactor scrams 2019: 1

Licensed annual emission limits in 2019

(values added up for Units B and C, site emission):

Emission of noble gases with plant exhaust air:

1.85 · 10 15 Bq

Emission of iodine-131 with plant exhaust air:

2.20 · 10 10 Bq

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.10 · 10 11 Bq

Proportion of licensed annual emission limits for radioactive

materials in 2019 for (values added up for Units B and C):

Emission of noble gases with plant exhaust air: 0.48 %

Emission of iodine-131 with plant exhaust air: 0.40 %

Emission of nuclear fission and activation products

with plant waste water (excluding tritium): 0.16 %

Collective dose:

0.79 Sv

60

50

40

30

20

10

0

10

9

8

7

6

5

4

3

2

1

92

2012

0.78

90

2013

Collective radiation dose of own

and outside personnel in Sv

1.36

90

2014

1.14

90

2015

1.49

86

2016

0.84

88

2017

0.89

90

2018

0.55

89

2019

0.79

0

2012

2013

2014

2015

2016

2017

2018 2019

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Isar 2

REPORT

Operating sequence in 2019

100

90

80

70

60

50

40

30

20

10

Electrical output in %

January February March April May June July August September October November December

With a gross electricity generation of 12,036,656 MWh and a

­availability of 95.68 %, unit 2 of Isar (KKI 2) nuclear power plant

achieved an excellent operating result in 2019. The unit also made an

important contribution to grid stability through increased load

sequence and control operation, which, however, reduced the net

electricity supply that can be generated by 384,950,000 MWh,

corresponding to 11.4 full operation days. The highest generator

­active power was reached on 20.01.2019 and amounted to 1520 MW.

Planned shutdowns

The fuel element replacement with plant inspection and maintenance

took place from 13 July 2019 to 27 July 2019 with a duration of

14.8 days. During the inspection and maintenance, 48 new fuel

100

elements were inserted.

80

Unplanned shutdowns and reactor/turbine trip

None.

60

Power reductions above 10 % and longer than for 24 h

40

None.

20

Safety Reviews

20

0

and 22 Februar: Management evaluation KKI.

6 March: Review of operations by the management

of PreussenElektra GmbH.

11 to 15 March:­­ ­ ­Re-certification audit by DNV GL

Business Assurance Zertifizierung & Umweltgutachter

GmbH according to DIN EN ISO

Positionierung:

Bezug, links, 9001/14001, untenBS OHSAS 18001 and EMAS.

12 March­­ ­ Inspection in accordance with §16 of the Major

and 30 April:­ ­ ­Accidents Ordinance (Störfall Verordnung) –

VGB: HKS6K Fire 30 Protection % and Immission Control.

3 and atw: 4 July: 100 60 Internal 0 0audit

“Measurement and test

equipment monitoring” in KKI.

7 August:­­ ­ Management system status meeting.

8, 18 October Management system audit - Part 1 in the KKI.

and 13 November:

10 October: 2 nd operational review (half-year review) by

the management of PreussenElektra GmbH.

29 October:­­ ­ Audit “Fuel element handling and

and 4/5 November fuel element disposal“.

4/5 December: Plant inspection “Integrated

Management System.

X = 20,475 Y = 95,25 B = 173,5 H = 38,2

Delivery of fuel elements

In the year under review, 40 uranium fuel elements were delivered

from Westinghouse. The dry storage facility contains 16 uranium fuel

elements in stock.

Waste management status

In 2019, no fuel elements were stored in the BELLA on-site interim

storage facility.

Of the storage and transport casks stored in the on-site interim

storage facility, 26 CASTOR © V/19 casks as well as 7 TN24E have to

be assigned to KKI 2.

The interim storage facility was taken over by Bundesgesellschaft

für Zwischenlagerung mbH (BGZ) on 1 January 2019.

General points

Due to increased oil temperatures at the engine of the main coolant

pump JEB30 AP001, a precautionary bearing oil change was carried

out. For this purpose, the plant output was reduced on 10 August

2019 from 07:00 hrs to the minimum load point of 360 MW with a

gradient of 30 MW/min.

After the scheduled completion of the exchange and start of the

main coolant pump, full load operation was reached again around

6:34 pm.

An internal emergency exercise was conducted on 18 November2019.

The exercise started within the normal working hours. The scenario

assumed was an earthquake close to the site, which caused the

failure of the external main electricity supply and the cooling water

supply of the emergency diesel generators as well as leakages in the

storage pool cooling water pipe. In addition, the failure of various

communication systems was assumed in the course of the exercise.

The emergency exercise was completed professionally and

purposefully by a highly motivated team.

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Operating data

Review period 2019

REPORT

Plant operator: PreussenElektra GmbH

Shareholder/Owner: PreussenElektra GmbH (75 %),

Stadtwerke München GmbH (25 %)

Plant name: Kernkraftwerk Isar 2 (KKI 2)

Address: PreussenElektra GmbH, Kernkraftwerk Isar,

Postfach 11 26, 84049 Essenbach, Germany

Phone: +49 8702 38-2465

Web: www.preussenelektra.de

100

90

80

94

Availability factor in %

Capacity factor in %

94

90

89

96

91

95

96

70

First synchronisation: 01-22-1988

Date of commercial operation: 04-09-1988

Design electrical rating (gross):

1,485 MW

Design electrical rating (net):

1,410 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

60

50

40

The following operating results were achieved:

Operating period, reactor:

8,405 h

Gross electrical energy generated in 2019:

12,036,656 MWh

Net electrical energy generated in 2019:

11,375,505 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2019:

365,762,469 MWh

Net electrical energy generated since

first synchronisation until 12-31-2019:

345,352,094 MWh

Availability factor in 2019: 95.95 %

Availability factor since

date of commercial operation: 93.36 %

Capacity factor 2019: 95.68 %

Capacity factor since

date of commercial operation: 92.48 %

Downtime

(schedule and forced) in 2019: 4.05 %

Number of reactor scrams 2019: 0

30

20

10

0

10

9

8

94

2012

96

2013

95

2014

89

2015

96

2016

92

2017

Collective radiation dose of own

and outside personnel in Sv

95

2018

96

2019

Licensed annual emission limits in 2019:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.1 · 10 15 Bq

1.1 · 10 10 Bq

5.5 · 10 10 Bq

Proportion of licensed annual emission limits

for radioactive materials in 2019 for:

Emission of noble gases with plant exhaust air: 0.127 %

Emission of iodine-131 with plant exhaust air:

< limit of detection

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

< limit of detection

Collective dose:

0.047 Sv

7

6

5

4

3

2

1

0

0.14

2012

0.08

2013

0.09

2014

0.25

2015

0.06

2016

0.14

2017

0.06 0.05

2018 2019

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Neckarwestheim II

REPORT

Operating sequence in 2019

100

90

80

70

60

50

40

30

20

10

Electrical output in %

January February March April May June July August September October November December

In 100 2019, the Neckarwestheim II nuclear power plant (GKN II)

generated a gross output of 10,411,400 MWh. Net electrical

­generation 80 was 9,758,339 MWh, of which 9,371,600 MWh were

­supplied to the public three-phase grid and 1,039,800 MWh to the

static 60 converter system of Deutsche Bahn AG. The plant was

­connected to the grid for 7,699.8 hours. This results in a time

­utilization 40 of 87.90 %.

Since

20

the three-phase alternating current machine was

­commissioned, 340,241,584 MWh gross and 318,174,476 MWh net

0

have been generated.

Planned shutdowns

100

9 August to 22 September: 36 th fuel element replacement and annual

inspection with maintenance.

80

The inspection and maintenance included the following priorities:

p Fuel element replacement with the use of 40 new fuel elements.

60

p Eddy current tests of the heating tubes of all 4 steam generators.

p Secondary tube sheet inspection on all 4 steam generators.

40

p Major overhaul of a primary-side safety valve on the

pressuriser JEF10.

20

p Pressure test of the heat exchangers in the volume control

0

system.

p Leak rate check of the containment.

p Partial major overhaul of the main feed water pump LAC30 and

the main condensate pump LCB10.

p Major overhaul of main steam valves at LBA10 and LBA40.

p Maintenance Positionierung:

activities on transformers and on

both Bezug, grid connections. links, unten

p Maintenance on the switchgear and on mechanical components

in the main redundancy 2/6.

X = 20,475 Y = 95,25 B = 173,5 H = 38,2

VGB: HKS6K 30 %

atw: 100 60 0 0

Unplanned shutdowns and reactor/turbine trip

3 to 22 September: Unplanned extension of the maintenance.

Power reductions above 10 % and longer than for 24 h

13 June to 9 August: Stretch-out operation

January to April and October to December: Load sequence ­operation.

Integrated management system (IMS) EnKK

The Integrated Management System (IMS) of EnBW Kernkraft

GmbH (EnKK) according to KTA 1402 with the partial systems

for nuclear safety (SMS), quality management (QMS/QSÜ),

­occupational safety management (AMS) as well as environmental

and energy management (UMS, EnMS) was also continuously

­further developed in 2019. The scope and content of the respective

process descriptions were gradually adapted to the various internal

requirements and the related approval-relevant specifications.

The completeness and effectiveness (conformity) of the processoriented

IMS, including the quality management measures, were

confirmed by appropriate internal audits as well as by a several-day

inspection by the assessor (ESN) and the supervisory authority at the

GKN and KKP sites.

The modular and demand-oriented structure of the IMS according to

KTA 1402 also enables continuous and efficient adaptation to the

site-specific requirements in operation/post-operation in subsequent

years. Another important focus will be the gradual integration of

dismantling aspects into the IMS in order to exploit synergy effects.

Waste management status

In 2019, 5 CASTOR © V/19 casks were loaded with 69 GKN I and 2

GKN II fuel elements and transported to the Neckarwestheim on-site

interim storage facility. At the end of 2019, 716 GKN II fuel elements

(dry storage, wet storage and reactor) and 49 GKN I fuel elements

(wet storage) were thus in the GKN II facility. Since 1 January 2019,

the on-site interim storage facility in Neckarwestheim has been

operated by the federally owned Gesellschaft für Zwischenlagerung

(BGZ). This is the implementation of the “Act on the Reorganisation

of Responsibility in Nuclear Waste Management”.

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Operating data

Review period 2019

REPORT

Plant operator: EnBW Kernkraft GmbH (EnKK)

Shareholder/Owner: EnBW Erneuerbare und Konventionelle

Erzeugung AG (98,45 %), ZEAG Energie AG, Deutsche Bahn AG,

Kernkraftwerk Obrigheim GmbH

Plant name: Kernkraftwerk Neckarwestheim II (GKN II)

Address: EnBW Kernkraft GmbH, Kernkraftwerk Neckarwestheim,

Im Steinbruch, 74382 Neckarwestheim, Germany

Phone: +49 7133 13-0

E-mail: poststelle-gkn@kk.enbw.com

Web: www.enbw.com

100

90

80

70

92

Availability factor in %

Capacity factor in %

90

93

93

94

89

81

88

First synchronisation: 01-03-1989

Date of commercial operation: 04-15-1989

Design electrical rating (gross):

1,400 MW

Design electrical rating (net):

1,310 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

60

50

40

The following operating results were achieved:

Operating period, reactor:

7,706 h

Gross electrical energy generated in 2019:

10,411,400 MWh

Net electrical energy generated in 2019:

9,758,339 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2019:

340,241,584 MWh

Net electrical energy generated since

first synchronisation until 12-31-2019:

318,174,476 MWh

Availability factor in 2019: 94.03 %

Availability factor since

date of commercial operation: 92.92 %

Capacity factor 2019: 88.00 %

Capacity factor since

date of commercial operation: 92.55 %

Downtime

(schedule and forced) in 2019: 5.87 %

Number of reactor scrams 2019: 0

30

20

10

0

10

9

8

92

2012

90

2013

93

2014

93

2015

95

2016

89

2017

Collective radiation dose of own

and outside personnel in Sv

81

2018

94

2019

Licensed annual emission limits in 2019:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.0 · 10 15 Bq

1.1 · 10 10 Bq

6.0 · 10 10 Bq

Proportion of licensed annual emission limits

for radioactive materials in 2019 for:

Emission of noble gases with plant exhaust air: 0.0098 %

Emission of iodine-131 with plant exhaust air:

< limit of detection

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

< limit of detection

Collective dose:

0.096 Sv

7

6

5

4

3

2

1

0

0.13

2012

0.08

2013

0.10

2014

0.12

2015

0.08

2016

0.15

2017

0.12 0.10

2018 2019

Report

Operating results 2019


atw Vol. 65 (2020) | Issue 8/9 ı August/September

454

Philippsburg 2

REPORT

Operating sequence in 2019

100

90

80

70

60

50

40

30

20

10

Electrical output in %

January February March April May June July August September October November December

In 100 the reporting year 2019 the nuclear power plant block

­Philippsburg 2 (KKP 2) generated a gross output of 10,993,639 MWh.

The

80

net electrical power generation consisted of 10,323,151 MWh.

The plant was 7,939 h on the grid. This corresponds to a availabilty

60

factor of 90.63 %.

Since the commissioning of the plant 366,161,155 MWh gross and

40

347,076,473 MWh net were generated

20

Planned shutdowns

11 May to 15 June: 33 nd refuelling and annual major inspection.

0

Major inspection work carried out:

p Inspection of one of the three main feed pumps.

p

100

Eddy current testing of two of the four steam generators.

p Leak test of reactor containment.

p 80 Inspection of the main cooling water system.

p Engine replacement on two of six main cooling water pumps.

p Maintenance work on individual emergency power generators.

60

Unplanned 40 shutdowns and reactor/turbine trip

18 August: Turbine trip (TUSA) via the criterion “high condenser

pressure”. 20

Power 0 reductions above 10 % and longer than for 24 h

15 March to 11 May: Stretch-out operation

26 July to 24 August: Reduction of heat input into the Rhine and

compliance with the permissible outlet temperature.

15 October to 2 November: Reduction of heat input into the Rhine

and compliance Positionierung:

with the permissible outlet temperature.

8 November Bezug, to 3 links, December: unten Reduction of heat input into the Rhine

and compliance with the permissible outlet temperature.

X = 20,475 Y = 95,25 B = 173,5 H = 38,2

VGB: HKS6K 30 %

atw: 100 60 0 0

Integrated management system (IMS) EnKK

(NPP P, GKN, KWO)

The integrated management system (IMS) of the EnBW Kernkraft

GmbH (EnKK) with its partial system for nuclear safety (SMS),

quality management (QMS/QSÜ) as well as environmental

and energy management (UMS, EnMS, Umwelt- und Energiemanagementsystem)

were also in 2019 continuously further

developed. Scope and content of each process descriptions were

gradually adapted to the different internal requirements and related

approval criteria. Besides the confirmation of conformity for the

IMS, the recertification of the EnKK energy management system

(EnMS, Energiemanagementsystem) according to DIN EN ISO 50001

took place in 2019 to improve energy efficiency. The certificate

was thus exten ded by three years.

The completeness and effectiveness of the process-oriented IMS,

including the quality management measures, were confirmed by

appropriate internal audits as well as by a several-day inspection by

the expert (ESN) and the supervisory authority at the GKN and KKP

sites.

The modular and demand-oriented structure of the IMS according

to KTA1402 also enables continuous and efficient adaptation to the

site-specific requirements in operation/post-operation in subsequent

years. Another important focus will be the gradual integration of

dismantling aspects into the IMS in order to exploit synergy effects.

Waste management status

During the year 2019 in total 2 transportation and storage casks of

type CASTOR © V/19 were stored in the on-site intermediate storage.

Altogether 33 loaded CASTOR © V/19 and 29 loaded CASTOR ©

V/25 casks were at the on-site intermediate storage.

Report

Operating results 2019


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455

Operating data

Review period 2019

REPORT

Plant operator: EnBW Kernkraft GmbH (EnKK)

Shareholder/Owner: EnBW Erneuerbare und Konventionelle

Erzeugung AG (98,45 %), ZEAG Energie AG, Deutsche Bahn AG,

Kernkraftwerk Obrigheim GmbH

Plant name: Kernkraftwerk Philippsburg 2 (KKP 2)

Address: EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg,

P.O. box 11 40, 76652 Philippsburg, Germany

Phone: +49 7256 95-0

E-mail: Poststelle-kkp@kk.enbw.com

Web: www.enbw.com

100

90

80

70

86

Availability factor in %

Capacity factor in %

73

82

90

82

90

89

First synchronisation: 12-17-1984

Date of commercial operation: 04-18-1985

Design electrical rating (gross):

1,468 MW

Design electrical rating (net):

1,402 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

60

50

40

63

The following operating results were achieved:

Operating period, reactor:

7,865 h

Gross electrical energy generated in 2019:

10,606,307 MWh

Net electrical energy generated in 2019:

9,963,117 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2019:

376,767,462 MWh

Net electrical energy generated since

first synchronisation until 12-31-2019:

357,039,589 MWh

Availability factor in 2019: 89.63 %

Availability factor since

date of commercial operation: 88.78 %

Capacity factor 2019: 89.31 %

Capacity factor since

date of commercial operation: 88.51 %

Downtime

(schedule and forced) in 2019: 10.37 %

Number of reactor scrams 2019: 0

30

20

10

0

10

9

8

86

2012

73

2013

82

2014

91

2015

82

2016

63

2017

Collective radiation dose of own

and outside personnel in Sv

91

2018

90

2019

Licensed annual emission limits in 2019:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.1 · 10 15 Bq

1.1 · 10 10 Bq

5.5 · 10 10 Bq

Proportion of licensed annual emission limits

for radioactive materials in 2019 for:

Emission of noble gases with plant exhaust air: 0.11 %

Emission of iodine-131 with plant exhaust air: 0.0002 %

Emission of nuclear fission and activation products

with plant waste water (excluding tritium): 0.06 %

Collective dose:

0.072 Sv

7

6

5

4

3

2

1

0

0.22

2012

0.16

2013

0.14

2014

0.15

2015

0.18

2016

0.07

2017

0.12 0.07

2018 2019

Report

Operating results 2019


atw Vol. 65 (2020) | Issue 8/9 ı August/September

456

NEWS

Top

Safe start-up of Unit 1 of

Barakah Nuclear Energy Plant

successfully achieved

p Start-up is major step in process

for upcoming generation of

emissions-free electricity

p Process undertaken in line with

regulatory requirements and

highest international standards for

nuclear quality and safety

(enec) On 1 August 2020, the Emirates

Nuclear Energy Corporation (ENEC)

announced that its operating and

maintenance subsidiary, Nawah

Energy Company (Nawah) has

successfully started up Unit 1 of the

Barakah Nuclear Energy Plant, located

in the Al Dhafrah Region of Abu Dhabi,

United Arab Emirates (UAE). This step

is the most historic milestone to date in

the delivery of the UAE Peaceful

Nuclear Energy Program, as part of

the process towards generating clean

electricity for the Nation for at least

the next 60 years.

Since receipt of the Operating

­License from the Federal Authority

for Nuclear Regulations (FANR) in

February 2020, and the completion of

fuel assembly loading in March 2020,

Nawah, the Joint Venture nuclear

operations and maintenance subsidiary

of ENEC and the Korea Electric

Power Corporation (KEPCO), has

been safely progressing through a

comprehensive testing program, prior

to successfully completing the start-up

of the first nuclear energy reactor of

the Barakah plant.

The start-up of Unit 1 marks the

first time that the reactor safely produces

heat, which is used to create

steam, turning a turbine to generate

electricity. Nawah’s qualified and

licensed team of nuclear operators

focus on safely controlling the process

and controlling the power output of

the reactor. After several weeks and

conducting numerous safety tests, Unit

1 will be ready to connect to the UAE’s

| United Arab Emirate: Barakah site during construction

electricity grid, delivering the first

megawatts of clean electricity to the

homes and businesses of the Nation.

Testing has been undertaken with the

continued oversight of the UAE’s

­independent nuclear regulator, FANR,

and follows the World Asso­ciation of

Nuclear Operator’s (WANO) completion

of a Pre Start-up Review (PSUR) in

January 2020, prior to receipt of the

Operating License, which ensures Unit

1 is aligned with international best

practice in the nuclear energy industry.

H.E. Mohamed Ibrahim Al

­Hammadi, Chief Executive Officer

of ENEC, said: “Today is a truly historic

moment for the UAE. It is the

culmination of more than a decade of

vision, strategic planning and robust

program management. Despite the

recent global challenges, our team has

demonstrated outstanding resilience

and commitment to the safe delivery

of Unit 1. We are now another step

closer to achieving our goal of supplying

up to a quarter of our Nation’s

electricity needs and powering its

future growth with safe, reliable, and

emissions-free electricity.

“Through the realization of the

vision of our Leadership, the Barakah

Nuclear Energy Plant has become an

engine of growth for the Nation. It will

deliver 25 % of the UAE’s electricity

with zero carbon emissions while also

supporting economic diversification

by creating thousands of high-value

jobs through the establishment of

a sustainable local nuclear energy

industry and supply chain. We are

grateful to the Leadership for their

continuous support in making this

remarkable achievement happen,

along with the support of our UAE

stakeholders and Korean partners,

and congratulate everyone involved

in the Program on this landmark

occasion.”

Once the unit is connected to the

grid, the nuclear operators will

continue with a process of gradually

raising the power levels, known as

Power Ascension Testing (PAT).

Throughout, the systems of Unit 1 are

continuously monitored and tested

as the unit proceeds towards full

electricity production in line with all

regulatory requirements and the

highest international standards of

safety, quality and security. Once the

process is completed over the course

of a number of months, the plant will

deliver abundant baseload electricity

at full capacity to power the growth

and prosperity of the UAE for decades

to come.

Commenting on this key milestone

in UAE nuclear energy operations, Eng.

Ali Al Hammadi, Chief Executive Officer

of Nawah, said: “The start-up of

Unit 1 is a significant milestone for

Nawah Energy Company as we fulfill

our mandate to operate and maintain

the plant in accordance with the

highest international standards of safety

and quality. The dedication of our

people as well as our close collaboration

with our Korean partners and cooperation

with numerous international

expert organizations has enabled this

accomplishment. This ­reflects our

commitment to upholding the highest

safety, quality and operational transparency

standards through out the entire

commissioning and startup process

by leveraging the expertise of the global

nuclear industry.

“I am especially proud of our

­talented UAE National engineers and

nuclear professionals who contributed

to the construction of Unit 1, as

well as the UAE National Senior

Reactor Operators and Reactor Operators

who have been certified to safely

operate the plant, alongside our international

experts, to ensure the safe

and sustainable operations of the unit

for decades to come.” concluded Eng.

Ali Al Hammadi.

The UAE is the first country in the

Arab World, and the 33 rd nation

globally, to develop a nuclear energy

plant to generate safe, clean, and reliable

baseload electricity. The Barakah

plant is significantly contributing to

the UAE’s efforts to move towards the

electrification of its energy sector, and

the decarbonization of electricity production.

When fully operational, the

plant will produce 5.6 gigawatts of

electricity while preventing the release

of more than 21 million tons of carbon

emissions every year, equivalent to the

removal of 3.2 million cars from the

Nation’s roads annually.

The UAE Peaceful Nuclear Energy

Program commenced in 2009, ENEC

has worked closely with international

nuclear bodies, including the

News


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Operating Results April 2020

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

457

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

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

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

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

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

KKG Gösgen 7) PWR CH 1060 1010 720 761 249 3 087 541 325 203 776 100.00 100.00 99.99 99.94 99.74 100.34

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

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

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

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

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

Temelin B1 4) PWR CZ 1080 1030 0 0 1 756 742 123 671 555 0 54.74 0 54.22 0 55.93

Temelin B2 PWR CZ 1080 1030 720 782 217 3 223 320 120 705 938 100.00 100.00 100.00 100.00 100.41 102.62

Doel 1 2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0

Doel 2 2) PWR BE 454 433 0 0 0 136 335 470 0 0 0 0 0 0

Doel 3 PWR BE 1056 1006 720 775 809 3 139 030 266 250 680 100.00 100.00 100.00 100.00 101.65 101.96

Doel 4 PWR BE 1084 1033 720 788 060 3 186 503 272 824 778 100.00 100.00 99.26 99.70 99.26 99.63

Tihange 1 2) PWR BE 1009 962 0 0 0 307 547 424 0 0 0 0 0 0

Tihange 2 PWR BE 1055 1008 720 754 589 3 042 824 261 097 343 100.00 100.00 99.98 99.87 100.34 100.33

Tihange 3 PWR BE 1089 1038 720 778 353 3 142 871 283 705 448 100.00 100.00 100.00 99.97 99.91 100.03

NEWS

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Since Month Year Month Year Month Year

commissioning

KBR Brokdorf DWR 1480 1410 720 951 530 3 708 586 364 429 609 100.00 100.00 93.80 94.13 89.00 85.93

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

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

KRB C Gundremmingen 3) SWR 1344 1288 590 746 240 3 295 647 344 619 199 81.99 86.09 77.15 84.34 76.49 83.84

KKI-2 Isar DWR 1485 1410 720 972 470 4 115 090 369 877 559 100.00 100.00 99.98 99.99 90.37 95.04

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

International Atomic Energy Agency

(IAEA), and WANO, in line with the

robust regulatory framework of FANR.

To date, more than 255 inspections

have been undertaken by FANR to

ensure the Barakah plant and its

people and processes meet the highest

standards of nuclear quality and

safety. These national reviews have

been supported by more than 40 assessments

and peer reviews by the

IAEA and WANO.

ENEC recently announced the

construction completion of Unit 2,

with operational readiness preparations

now underway by Nawah.

Construction of Units 3 and 4 of the

Barakah Nuclear Energy Plant is in the

final stages, with the overall construction

completion of the four units now

standing at 94 %.

| www.enec.gov.ae (201121401)

World

IAEA Launches Initiative to

Help Prevent Future Pandemics

(iaea) The Director General of the

­International Atomic Energy Agency

(IAEA), Rafael Mariano Grossi,

launched an initiative today to

strengthen global preparedness for future

pandemics like COVID-19. The

project, called ZODIAC, builds on

the IAEA’s experience in assisting

countries in the use of nuclear and

nuclear-derived techniques for the

rapid detection of pathogens that

cause transboundary animal diseases,

including ones that spread to humans.

These zoonotic diseases kill around

2.7 million people every year.

The IAEA Zoonotic Disease Integrated

Action (ZODIAC) project will

establish a global network to help

national laboratories in monitoring,

surveillance, early detection and control

of animal and zoonotic diseases

such as COVID-19, Ebola, avian influenza

and Zika. ZODIAC is based on

the technical, scientific and laboratory

capacity of the IAEA and its partners

and the Agency’s mechanisms to

quickly deliver equipment and knowhow

to countries.

The aim is to make the world

better prepared for future outbreaks.

“Member States will have access

to equipment, technology packages,

expertise, guidance and training.

Decision-makers will receive up-todate,

user-friendly information that

will enable them to act quickly,”

Mr Grossi told a meeting of the IAEA

Board of Governors.

Mr Grossi said COVID-19 had

exposed problems related to virus

*)

Net-based values

(Czech and Swiss

nuclear power

plants gross-based)

1)

Refueling

2)

Inspection

3)

Repair

4)

Stretch-outoperation

5)

Stretch-inoperation

6)

Hereof traction supply

7)

Incl. steam supply

BWR: Boiling

Water Reactor

PWR: Pressurised

Water Reactor

Source: VGB

News


atw Vol. 65 (2020) | Issue 8/9 ı August/September

Operating Results May 2020

458

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

NEWS

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

OL2 Olkiluoto 1) BWR FI 910 880 476 414 256 3 088 454 262 452 539 63.97 92.65 60.56 91.76 60.52 92.05

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

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

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

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

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

Dukovany B1 PWR CZ 500 473 744 368 085 1 819 513 117 703 696 100.00 100.00 100.00 99.92 98.95 99.78

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

Dukovany B3 PWR CZ 500 473 744 354 979 639 874 110 891 610 100.00 36.33 96.72 35.39 95.42 35.09

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

Temelin B1 PWR CZ 1080 1030 408 399 882 2 156 624 124 071 437 54.84 54.76 50.08 53.37 49.67 54.65

Temelin B2 PWR CZ 1080 1030 744 800 038 4 023 358 121 505 976 100.00 100.00 99.84 99.97 99.38 101.96

Doel 1 2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0

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

Doel 3 PWR BE 1056 1006 744 798 061 3 937 091 267 048 741 100.00 100.00 100.00 100.00 101.19 101.80

Doel 4 PWR BE 1084 1033 744 808 168 3 994 671 273 632 946 100.00 100.00 99.40 99.64 98.57 99.41

Tihange 1 2) PWR BE 1009 962 0 0 0 307 547 424 0 0 0 0 0 0

Tihange 2 PWR BE 1055 1008 744 772 511 3 815 335 261 869 854 100.00 100.00 99.96 99.89 99.38 100.13

Tihange 3 PWR BE 1089 1038 744 799 791 3 942 662 284 505 238 100.00 100.00 100.00 99.97 99.27 99.87

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Since Month Year Month Year Month Year

commissioning

KBR Brokdorf DWR 1480 1410 744 1 001 313 4 709 899 365 430 922 100.00 100.00 94.33 94.17 90.70 86.90

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

KWG Grohnde 2) DWR 1430 1360 188 243 479 3 508 965 391 783 810 98.32 87.21 96.87 87.08 22.69 66.85

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

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

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

detection capabilities in many countries,

as well as a need for

better communication between health

institutions around the world. While

the IAEA has been doing important

work to help countries in these areas,

such as through the provision of

­COVID-19 tests, he said it was “essential

to pull these diverse strands

together into a coherent and comprehensive

framework of assistance”.

Nuclear-derived techniques, such

as tests using real time reverse

transcription–polymerase chain reaction

(RT-PCR), are important tools in

the detection and characterization

of viruses. The IAEA is providing

emergency assistance to some 120

countries in the use of such tests to

rapidly detect COVID-19.

Zoonotic diseases are caused by

bacteria, parasites, fungi or viruses

that originate in animals and can be

transmitted to humans. Many of these

diseases are treatable if medication is

available, such as E. coli- and brucella

bacterial infections. But others

have the potential to severely affect

­humans, such as Ebola, SARS and

COVID-19.

ZODIAC builds on the experience

of VETLAB, a network of veterinary

laboratories in Africa and Asia that

was originally set up by the Food

and Agriculture Organization of the

­United Nations (FAO) and the IAEA to

combat the cattle disease rinderpest.

VETLAB now supports countries in

the early detection of several zoonotic

and animal diseases, such as African

swine fever and pest des petit ruminants

(PPR).

“About 70 per cent of all diseases in

humans come from animals,” said

Gerrit Viljoen, Head of the Animal

Production and Health Section of

the Joint FAO/IAEA Programme for

Nuclear Techniques in Food and

­Agriculture.

ZODIAC aims to help veterinary

and public health officials identify

these diseases before they spread.

“We have seen an increase in the number

of zoonotic epidemics in the last

decades: first Ebola, then Zika, and

now COVID-19. It’s important to monitor

what is in the animal kingdom –

both wildlife and livestock – and to

act quickly on those findings before

the pathogens jump to humans,”

Mr ­Viljoen said.

Following the One Health concept

for a multidisciplinary collaborative

approach between human and animal

health authorities and specialists,

­ZODIAC will benefit from the unique

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

joint FAO/IAEA laboratories and from

partners such as the World Health Organization

(WHO) and the World Organisation

for Animal Health (OIE).

“We have a unique capacity to provide

laboratory support and guidance

to countries,” said Mr Viljoen, adding

that ZODIAC will, for example, provide

technical know-how and advice

to laboratories on test performance

and assist authorities in the interpretation

of results and in devising containment

measures.

ZODIAC will also support R&D

activities for novel technologies and

methodologies for early detection and

surveillance. Under the project, the

IAEA will enhance its capacities to

host scientists and fellows from

Member States at its Seibersdorf

­laboratories outside Vienna and to

carry out research on immunological,

molecular, nuclear and isotopic tests,

as well as in the use of irradiation to

develop vaccines against diseases

such as avian influenza.

| www.iaea.org (201711216)

Research

TUM FRMII: Improved welding

process for turbine parts

(tum) Scientists at the Heinz Maier-

Leibnitz Zentrum (MLZ) use imaging

techniques with neutrons to study

specially bonded steel components.

With their results they can improve

the welding process for oil and gas

pipelines and turbines.

The “transient liquid-phase

bonding” (TLPB) process is used for

joining metallic systems where

standard welding methods cannot be

used. For example in the repair of gas

pipelines. By using special materials,

the TLPB process achieves very good

mechanical properties.

Researchers at MLZ have now

taken a closer look at this type of

welded joint. With their method they

can non-destructively test the quality

of TLPB joints. In doing so, they

discover harmful inclusions at the

fusion joint. This is particularly important

for the oil and gas industry, where

the TLPB method is used particularly

often. It is also used in the manufacture

of turbine blades in jet engines.

Neutrons make the joining

seam visible

Scientists Dr. Nicolás Di Luozzo and

Dr. Marcelo Fontana from the Universidad

de Buenos Aires (UBA) together

with MLZ scientist Dr. Michael Schulz

(Technical University Munich) have

investigated TLP-bonded steel components

with boron-alloyed foils.

For this purpose they used neutron radiography

and tomography at the instrument

ANTARES.

The neutrons make the “seam”

between the metal parts, the microstructure

of the compound and the

subsequent base metal, visible. “With

existing microscopic methods it is not

possible to quantify the boron concentrations

in a steel matrix,” Di Luozzo

explains, “quite contrary to neutron

radiography. The spatial resolution

achieved by ANTARES is remarkable.”

Researchers identify flaws in the

welding process

Within the joint, Di Luozzo and his

colleagues identify two areas: they

distinguish where all solidification

phases are completed and where they

are not. In case of incomplete solidification,

the scientists find borides in

addition to ferrite. This is an indicator

of weak points in the joint: the TLPB

process did not succeed perfectly at

these points and the joint is defective.

Using neutron tomography, the

scientists were able to reconstruct the

size, quantity and exact location of

these weak points in three-dimensional

form. Here they still have to

improve the welding process in order

to obtain a perfect seam.

Non-destructive testing thanks

to neutrons

In their experiment, the scientists

showed that neutrons are suitable for

making borides visible on welding

seams. This helps to assess the quality

of TLPB welded joints. “The borides at

the joint have an extremely negative

influence on its mechanical properties,”

explains Di Luozzo. In the worst

case, this means that the weld is brittle

and breaks, and the joined components

no longer hold together.

In theory, the TLPB process enables

very good mechanical properties. “It is

possible for joined materials to achieve

properties comparable to those of the

base material,” emphasizes Di Luozzo.

The product is therefore as strong as if

it were made in one piece. That's why

the process is so important: the

neutrons can detect defective spots

in the welded end product – and in

the case of an aircraft engine, this is

vital for passengers' survival.

Original publication:

Di Luozzo, N., Schulz, M. & Fontana,

M. Imaging of boron distribution

in steel with neutron radiography

and tomography. J Mater Sci 55,

7927–7937 (2020). DOI: 10.1007/

s10853-020-04556-z

| www.frm2.tum.de (202321930)

Company News

Framatome and Lockheed

Martin join forces to provide

additional solution for US nuclear

plant instrumentation

and control

(framatome) Framatome and Lockheed

Martin recently signed a teaming

agreement that will integrate Lockheed

Martin’s Discrete Logic Solving

System (DLSS) into proven Framatome

instrumentation and control (I&C)

nuclear plant modernization solutions.

This additional analog solution combines

the companies’ technologies and

supports the safety and reliability of

nuclear power plants.

The I&C system serves as part of

the plant’s “central nervous system.”

It provides operators with critical

information on plant operation,

allows them to control various plant

safety systems during routine operations

and automatically protects the

reactor if needed.

“With Lockheed Martin’s Discrete

Logic Solving System now part of our

I&C portfolio, U.S. customers will have

access to additional analog-based

safety solution options for upgrading

their existing equipment,” said Clayton

Scott, senior vice president global

sales and deputy for the I&C Business

Unit at Framatome. “While Fram atome

focuses on helping our U.S. customers

transition to digital I&C systems, it’s

important that we con tinue to serve

our non-digital customers with safe

and reliable solutions.”

DLSS is one of Lockheed Martin’s

non-digital I&C solutions offering

effective applications for nuclear

systems that currently employ analog

technology. It supports operational

measures of nuclear plant systems

by monitoring, calculating and activating

plant interfaces.

“Lockheed Martin is proud to

partner with Framatome in support of

incorporating DLSS on future nuclear

energy I&C modernization efforts,”

said John Pericci, Lockheed Martin

Nuclear Systems & Solutions program

director. “For more than 60 years, we

have provided safety critical nuclear

I&C systems to commercial and U.S.

government customers, enhancing

operation and reliability in the

industry.”

459

NEWS

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460

NEWS

In 2019, Framatome and Lockheed

Martin partnered to complete the first

installation and site acceptance testing

of a new Engineered Safety

­Features Actuation System (ESFAS)

replacement that included DLSS. A

second installation at the same plant

took place earlier this year. Both

systems are fully operational and

meeting all requirements.

| www.framatome.com (201711204)

Westinghouse program

awarded £10m from

UK Government advanced

modular reactor project

(westinghouse) Westinghouse Electric

Company today announced their

Lead-cooled Fast Reactor (LFR) program

has successfully progressed to

Phase 2 of the UK Government’s

Department for Business, Energy

and Industrial Strategy’s (BEIS)

­Advanced Modular Reactor (AMR)

Feasibility and Development project,

receiving £10m ($12.5m) in funding

from the BEIS Energy Innovation

Portfolio. The funding will help

to advance nuclear technology

through innovation in order to deliver

reliable, clean energy for future

generations.

As part of Phase 2, Westinghouse, in

collaboration with industry, research

centres and academic partners, will

utilise the funding to undertake applied

research and development activities.

The award will be used to demonstrate

LFR components and accelerate the

development of high-temperature

materials, advanced manufacturing

technologies and modular construction

strategies for the LFR.

“Our progression to Phase 2 builds

on our eighty-year history in the

UK as a Strategic National Asset,” said

Patrick Fragman, Westinghouse president

and chief executive officer. “This

is the perfect combination of reducing

the cost of electricity and maintaining

a leading edge of science, research

and innovation for the UK.”

The Westinghouse LFR, a 450 MWe-­

class Generation IV reactor design, has

the potential to have a transformative

effect on the cost and market flexibility

of new nuclear. The key features of

the Westinghouse LFR include a

­simplified design, flexible operations

and fuel cycle capabilities, zero CO 2

emissions, walk-away safety features

and modular assembly. The Westinghouse

LFR will also achieve a competitive

Levelised Cost of Electricity

(LCoE) to ensure economic competitiveness

in the most challenging

global electricity markets.

As part of the AMR project, some of

the development facilities will be

established at the Clean Energy

­Technology Park at Springfields. The

Clean Energy Technology Park is

contributing towards the UK’s Net

Zero ambitions by leveraging the

existing strategic national asset,

Springfields, to support innovation

and collaborative partnerships, whilst

providing opportunities for bringing

highly-skilled jobs to the North West

of England and significant economic

benefits to the UK.

Westinghouse will deliver the

Phase 2 program in collaboration with

Ansaldo Nucleare and ENEA, in

addition to Bangor University, Frazer-

Nash Consultancy, Jacobs, National

Nuclear Laboratory (NNL), Nuclear

Advanced Manufacturing Research

Centre (NAMRC), the University

of Cambridge, the University of

­Manchester and Vacuum Process

­Engineering, Inc. (VPE).

| www.westinghousenuclear.com

(201711200)

Orano acquires KSE and

strengthens its position

in industrial maintenance

(orano) Orano announces the acquisition

of the company KSB Service

Energie (KSE) and of its subsidiaries

KSB Service Cotumer (KSC) and the

Société de Travaux et d’Ingénierie

Industrielle (STII) from the German

group KSB, a global player in the

manufacturing of industrial pumps

and valves.

This acquisition, effective as of July

1st, is part of Orano's strategy to

develop its service activities, in

particular in the area of industrial

maintenance. KSE and its subsidiaries

KSC and STII are recognized for the

role they play in providing services

to the French nuclear fleet and to

the industry, whether it be carrying

out interventions on valve systems,

mechanical maintenance on rotating

machines or boilermaking services

(anchor points, supports, piping, etc.).

With this acquisition, Orano

enhances its service offer with new

specialized resources which complement

the nuclear maintenance

activities in which the group is already

present. More than 250 employees

and the industrial capacities of

the three entities will be joining

forces with Orano's “Dismantling

and Services” unit, whose 1,600 employees

already work on the French

nuclear fleet on a daily basis in

industrial logistics, site support and

maintenance.

Philippe Knoche, Chief Executive

Officer of Orano, declared: “I ­welcome

our new colleagues from KSE, STII

and KSC to the Orano group. Their

arrival is concrete evidence of our

ambition to develop in the area of

service activities and of our confidence

in the future of nuclear, for

which needs in maintenance will

remain strong over the years to come.”

Alain Vandercruyssen, Senior

­Executive Vice President in charge of

Orano’s “Dismantling and Services”

unit, added: “with this transaction,

Orano reinforces its expertise and its

qualifications to achieve the critical

mass necessary to become a major

player in maintenance for EDF. The

contribution that KSE and its subsidiaries

bring to the table also consolidates

our presence with nuclear

operators on international markets.”

| www.orano.group (202330732)

People

Daniel Oehr takes over from

Dr Hannes Wimmer

as Chairman of the

GNS Management Board

(gns) After nine very successful years

as Chairman of the Management

Board of GNS Gesellschaft für

Nuklear- Service mbH, Dr Hannes

Wimmer (56) will leave GNS at the

end of this year by best mutual agreement

with the shareholders.

“On behalf of all shareholders, I

would like to thank Hannes Wimmer

for his great commitment, especially in

the internationalisation and repositioning

of GNS over the past years.

With the spin-off of the interim storage

and final disposal activities and the

successful acquisition of Höfer &

Bechtel as well as Eisenwerk Bassum,

the management team with Mr.

Wimmer has succeeded in developing

GNS into a renowned cask manufacturer

and service provider in the field

of nuclear waste management,” says

Dr. Guido Knott, Chairman of the

Supervisory Board of GNS.

Wimmer's successor will be

Daniel Oehr (43), currently Head

of Controlling and Performance

Management at GNS shareholder

Preussen Elektra GmbH, who has been

familiar with GNS for many years.

Daniel Oehr will join the management

board on November 1, 2020 and will

take over the tasks of Dr. Wimmer as

of December 1, 2020, in particular as

chairman of the management board

and CEO.

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atw Vol. 65 (2020) | Issue 8/9 ı August/September

Uranium

Prize range: Spot market [USD*/lb(US) U 3 O 8 ]

140.00

120.00

) 1

Uranium prize range: Spot market [USD*/lb(US) U 3 O 8 ]

140.00

120.00

) 1

461

100.00

100.00

80.00

60.00

40.00

20.00

0.00

1980

Jan. 2009

Yearly average prices in real USD, base: US prices (1982 to1984) *

Jan. 2010

1985

Jan. 2011

* Actual nominal USD prices, not real prices referring to a base year. Year

1990

Jan. 2012

Jan. 2013

1995

Jan. 2014

2000

Jan. 2015

Jan. 2016

2005

Jan. 2017

2010

Jan. 2018

Jan. 2019

2015

Jan. 2020

2020

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020

Year

* Actual nominal USD prices, not real prices referring to a base year. Year

) 1 ) 1

prices, * Actual nominal USD not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020

| Uranium spot market prices from 1980 to 2020 and from 2009 to 2020. The price range is shown.

In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.

Separative work: Spot market price range [USD*/kg UTA]

Conversion: Spot conversion price range [USD*/kgU]

180.00

26.00

24.00

160.00

22.00

140.00

120.00

100.00

80.00

60.00

40.00

20.00

0.00

Jan. 2021

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020

80.00

60.00

40.00

20.00

0.00

20.00

18.00

16.00

14.00

12.00

10.00

Jan. 2009

8.00

6.00

4.00

2.00

0.00

Jan. 2009

Jan. 2010

Jan. 2010

Jan. 2011

Jan. 2011

Jan. 2012

Jan. 2012

Jan. 2013

Jan. 2013

Jan. 2014

Jan. 2014

* Actual nominal USD prices, not real prices referring to a base year. Year

Jan. 2015

Jan. 2015

Jan. 2016

Jan. 2016

Jan. 2017

Jan. 2017

Jan. 2018

Jan. 2018

Jan. 2019

Jan. 2019

Jan. 2020

Jan. 2020

Jan. 2021

Jan. 2021

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020

NEWS

| Separative work and conversion market price ranges from 2009 to 2020. The price range is shown.

)1

In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps.

* Actual nominal USD prices, not real prices referring to a base year

Sources: Energy Intelligence, Nukem; Bilder/Figures: atw 2020

“The shareholders are pleased to

have won Daniel Oehr for this

challenging task. With him in the

lead, GNS will be further developed

into a quality and customer-oriented

service provider with increasing

business in third markets.”

In addition, Sascha Bechtel (49)

will, at the request of the shareholders,

support the GNS management

in the repositioning of GNS and

will be responsible for LLW-ILW

residues and waste and disposal projects.

He will continue to per