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atw - International Journal for Nuclear Power | 11/12.2019

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

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

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nucmag.com<br />

2019<br />

<strong>11</strong>/12<br />

Per<strong>for</strong>mance Shaping<br />

Factors <strong>for</strong> Human Error<br />

Reduction<br />

Decommissioning & Dismantling<br />

of the Rossendorf Research<br />

Reactor RFR<br />

First On-site Demonstration<br />

of Laser- based Decontamination<br />

Technology in Germany<br />

ISSN · 1431-5254<br />

24.– €<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

TMI and Lessons Learned – Afterwards<br />

and <strong>for</strong> the Future<br />

No technical development is perfect from the outset, each involves specific risks. A responsible approach to<br />

technology there<strong>for</strong>e requires a responsible and <strong>for</strong>ward-looking approach to its risks, the best possible protection and,<br />

if necessary, further development. This also means that there are no such things as absolutely perfect technology and<br />

absolute safety.<br />

With a view to criticism of technologies, technology sceptics<br />

are provided with a com<strong>for</strong>table approach <strong>for</strong> transporting<br />

their respective concerns by anniversaries. One such event<br />

is the 1979 Three-Mile-Island (TMI 2) accident.<br />

On 28 March 1979, shortly after 4 a.m., the feed water<br />

supply <strong>for</strong> the steam generators in unit 2 of the Three Mile<br />

Island (TMI 2) nuclear power plant near Harrisburg,<br />

Pennsylvania, USA, failed. Such a malfunction was taken<br />

into account in the operating procedures with a view to<br />

the future and can be easily dealt with by the plant and<br />

operators. The event is one of the design cases. But what<br />

began as a normal incident later developed into the most<br />

serious accident to date in a Western nuclear power plant<br />

due to a combination of technical defects, human error<br />

and organisational inadequacies.<br />

The technical aspects of the TMI-2 accident have been<br />

extensively reviewed and can be traced in detail in the<br />

literature: After the failure of the feed-water supply, the<br />

safety systems of the nuclear power plant reacted as<br />

designed. However, due to valves in the emergency feedwater<br />

system that were closed manually on site, the steam<br />

generators were not supplied with water and a link in the<br />

chain <strong>for</strong> the residual heat dissipation of the reactor, which<br />

was switched off but still supplies heat due to the decay<br />

heat, was missing. This error was corrected 8 minutes<br />

after the start of the fault. Subsequently, the increase in<br />

pressure in the primary circuit caused the safety relief<br />

valve on the pressure holder to open, the escape of steam<br />

from the primary circuit and the intended pressure<br />

reduction. However, the opened blow-off valve did not<br />

close as designed. This caused large quantities of coolant<br />

to escape from the primary circuit – a loss of coolant<br />

accident “ LOCA” had occurred. Only 2 hours later it was<br />

detected by the operators and corrected by closing other<br />

existing shut-off valves. After further misinterpretations<br />

about the condition of the reactor core and the primary<br />

circuit, the core was no longer cooled by water and the<br />

exothermic reaction between water vapour and the<br />

cladding tube material of the fuel elements then set on,<br />

releasing additional heat. About half of the fuel rod mass<br />

in the core melted or was severely damaged. Core masses<br />

shifted and partially reached the bottom of the reactor<br />

pressure vessel where they solidified. After it had been<br />

recognised that a core melt was in full swing, measures <strong>for</strong><br />

core cooling were taken and it was possible to ensure the<br />

heat removal of the TMI-2 core again in the long term.<br />

Of the radioactive substances released in the plant, a<br />

small part was released into the environment. Triggered<br />

by fears of the additional radiation exposure associated<br />

with the release of a calculated average of 0.02 mSv – i.e.<br />

about 1 % of the natural annual radiation exposure – the<br />

Pennsylvania Department of Health kept a register <strong>for</strong><br />

18 years containing the data of more than 30,000 persons<br />

who had lived within a five-mile radius of the nuclear<br />

power plant at the time of the accident. This state register<br />

was closed in June 1997 after no unusual health<br />

developments had been identified. More than a dozen<br />

comprehensive studies on the physical health effects of the<br />

accident did not provide evidence of an abnormally high<br />

number of cancer cases in the region around the nuclear<br />

power plant years after the event.<br />

Despite all the shortcomings and shortcomings that<br />

led to the TMI-2 accident, the basic safety concept of<br />

Western nuclear power plants that had been established at<br />

an early stage was confirmed: The amount of radioactive<br />

substances released from the plant was low and there were<br />

no fatalities or injuries. On the other hand, the accident<br />

sequence underlines the suitability of successively<br />

staggered safety barriers, which, taking even serious<br />

accidents into account, aims at the confinement of the<br />

radioactive substances in the plant itself: The reactor<br />

building of TMI-2 withstood, the containment remained<br />

intact.<br />

In addition, research and development <strong>for</strong> reactor<br />

safety further minimised the risks in an international<br />

context during the past 40 years: Expressed in figures,<br />

even <strong>for</strong> running plants in Germany and elsewhere, a<br />

safety level was achieved, which is 100 times higher than<br />

the original international reference level. The reactor<br />

disaster at Fukushima in 20<strong>11</strong> will not change this point<br />

of view either, as other <strong>for</strong>eseeable risks and specific<br />

boundary conditions were the causes.<br />

Safety is and remains an international task <strong>for</strong> all<br />

concerned. <strong>Nuclear</strong> safety and the maintenance and<br />

promotion of competences, especially when they are<br />

available with excellent know-how in research and<br />

development and are internationally recognised – as in<br />

Germany – are part of the overall social and political<br />

responsibility at the same level as other goals of<br />

environmental protection. Today's and future safety and<br />

the promotion of safety culture and technology, also <strong>for</strong><br />

nuclear technology, cannot and must not be the subject<br />

of restricted or restrictive action. Research and<br />

development should always be allowed to advance in<br />

the spirit of freedom of science and technology.<br />

Christopher Weßelmann<br />

– Editor in Chief –<br />

507<br />

EDITORIAL<br />

Editorial<br />

TMI and Lessons Learned – Afterwards and <strong>for</strong> the Future


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

508<br />

EDITORIAL<br />

TMI und Lessons Learned – danach<br />

und für die Zukunft<br />

Keine technische Entwicklung ist von vornherein perfekt, jede birgt spezifische Risiken. Ein verantwortungsvoller<br />

Umgang mit Technik er<strong>for</strong>dert daher einen verantwortungsvollen und vorausblickenden Umgang mit ihren<br />

Risiken, bestmöglichen Schutz und ggf. auch spätere Weiterentwicklung. Dies bedeutet jedoch auch, dass es eine<br />

absolut perfekte Technik nicht gibt, ebenso keine absolute Sicherheit.<br />

Mit Blick auf Kritik an Technologien liefern Jahrestage gerne<br />

Technologieskeptikern einen kom<strong>for</strong>tablen Ansatz, ihre<br />

jeweiligen Anliegen zu transportieren. Ein solches Ereignis<br />

ist auch der Three-Mile-Island (TMI 2)-Unfall von 1979.<br />

Am 28. März 1979, kurz nach 4 Uhr morgens, fiel im<br />

Block 2 des US-amerikanischen Kernkraftwerks Three Mile<br />

Island (TMI 2) nahe Harrisburg, Pennsylvania, die<br />

Speisewasser versorgung der Dampferzeuger aus. Eine<br />

solche Störung ist in den Betriebsabläufen vorausblickend<br />

berücksichtigt und kann von Anlage und Operateuren ohne<br />

Weiteres bewältigt werden. Das Ereignis gehört zu den<br />

Auslegungsfällen. Doch was als normale Störung begann,<br />

entwickelte sich im weiteren Verlauf aufgrund einer Kombination<br />

aus technischen Mängeln, mensch lichen Fehlern und<br />

organisatorischen Unzulänglichkeiten zum bis dahin schwerwiegendsten<br />

Unfall in einem west lichen Kernkraftwerk.<br />

Die technischen Aspekte zum TMI-2-Unfall sind ausgiebig<br />

aufgearbeitet und können in der Literatur detailliert<br />

nachvollzogen werden: Nach dem Ausfall der Speisewasserversorgung<br />

reagierten die Sicherheitssysteme des Kernkraftwerks<br />

auslegungsgemäß. Aufgrund versehentlich<br />

manuell Vor-Ort geschlossener Ventile im Notspeisewassersystem<br />

fehlte aber die Bespeisung der Dampferzeuger und<br />

damit ein Glied in der Kette für die Nachwärmeabfuhr des<br />

zwar abgeschalteten aber durch die Nachzerfalls wärme<br />

immer noch Wärme liefernden Reaktors. Dieser Fehler<br />

wurde 8 Minuten nach Störungsbeginn korrigiert. In<br />

weiterer Folge kam es durch den Druckanstieg im Primärkreislauf<br />

zum Öffnen des für diesen Fall vorgesehenen<br />

Sicherheits-Abblaseventils am Druckhalter, dem Entweichen<br />

von Dampf aus dem Primärkreislauf und der<br />

beabsichtigten Druckabsenkung. Das geöffnete Abblaseventil<br />

schloss aber folgend nicht auslegungsgemäß. Über<br />

diesen Weg entwichen große Mengen Kühlmittel aus dem<br />

Primärkreislauf – ein Loss of coolant accident „LOCA“ war<br />

eingetreten. Erst 2 Stunden später wurde er durch die<br />

Operateure erkannt und durch Schließen weiterer vorhandener<br />

Absperrarmaturen korrigiert. Nach weiteren<br />

Fehlinterpretationen über den Zustand von Reaktorkern<br />

und Primärkreislauf war der Kern nicht mehr durch Wasser<br />

gekühlt und die dann einsetzende exotherme Reaktion<br />

zwischen Wasserdampf und dem Hüllrohrmaterial der<br />

Brenn elemente setzte zusätzliche Wärme frei. Etwa die<br />

Hälfte der Brennstabmasse im Kern schmolz bzw. wurde<br />

schwer beschädigt. Kernmassen verlagerten sich und<br />

erreichten teilweise den Boden des Reaktordruckbehälters,<br />

wo sie erstarrten. Nachdem erkannt worden war, dass eine<br />

Kernschmelze in vollem Gange war, wurden Maßnahmen<br />

zur Kernkühlung ergriffen und es gelang, die Wärmeabfuhr<br />

des TMI-2-Kerns wieder langfristig zu gewährleisten.<br />

Von den in der Anlage frei gesetzten radioaktiven<br />

Stoffen gelangte ein geringer Teil in die Umgebung.<br />

Ausgelöst durch die Ängste der mit der Freisetzung<br />

verbundenen zusätzlichen Strahlenbelastung von berechneten<br />

im Mittel 0,02 mSv – also etwa 1 % der natürlichen<br />

jährlichen Strahlenbelastung – führte das Pennsylvania<br />

Department of Health während 18 Jahren ein Register mit<br />

den Daten von mehr als 30.000 Personen, die zum Zeitpunkt<br />

des Unfalls im Umkreis von fünf Meilen um das<br />

Kernkraftwerk gelebt hatten. Dieses staatliche Register<br />

wurde im Juni 1997 geschlossen, nachdem keine ungewöhnlichen<br />

Entwicklungen bei der Gesundheit festgestellt<br />

worden waren. Mehr als ein Dutzend Studien über die<br />

Auswirkungen des Unfalls auf die physische Gesundheit<br />

gaben auch Jahre nach dem Ereignis keine Hinweise auf<br />

eine abnormal hohe Zahl von Krebsfällen in der Region um<br />

das Kernkraftwerk.<br />

Trotz aller Unzulänglichkeiten und Fehler die zum TMI-<br />

2-Unfall führten, bestätigte sich einerseits das grundlegende<br />

frühzeitig etablierte Sicherheitskonzept westlicher<br />

Kernkraftwerke: Die aus der Anlage freigesetzte Menge an<br />

radioaktiven Stoffen war gering und weder Todesopfer<br />

noch Verletzte waren zu beklagen. Und andererseits<br />

unterstreicht der Unfallablauf den Sinn hintereinander<br />

gestaffelter Sicherheitsbarrieren, die unter Berücksichtigung<br />

auch schwerer Unfälle, den Einschluss der radioaktiven<br />

Stoffe in der Anlage selbst zum Ziel haben: Das<br />

Reaktorgebäude von TMI-2 hielt stand, der Sicherheitsbehälter<br />

blieb intakt.<br />

Zudem haben Forschung und Entwicklung für die<br />

Reaktor sicherheit in den letzten 40 Jahren die Risiken<br />

im internationalen Kontext weiter minimiert: In Zahlen<br />

ausgedrückt wurde selbst für laufende Anlagen in<br />

Deutschland und anderswo ein Sicherheitsniveau erreicht,<br />

das um den Faktor 100 höher liegt, als das ursprüngliche<br />

inter nationale Referenz niveau. Daran ändert auch die<br />

Reaktorhavarie von Fukushima 20<strong>11</strong> nichts, da andere<br />

absehbare Ursachen und spezifische Rand bedingungen<br />

Auslöser waren.<br />

Sicherheit ist und bleibt dabei eine international zu<br />

lebende Aufgabe für alle Beteiligten. Kerntechnische<br />

Sicherheit und der Erhalt und die Förderung von<br />

Kompetenzen, vor allem dann, wenn sie mit exzellentem<br />

Know-how in Forschung und Entwicklung vorhanden und<br />

international anerkannt sind – wie in Deutschland –<br />

gehören auf gleicher Ebene zur gesamtgesellschaftlichen<br />

und politischen Verantwortung wie andere Ziele des<br />

Umweltschutzes. Heutige und künftige Sicherheit und<br />

Förderung von Sicherheitskultur und -technik, auch für<br />

die Kerntechnik, können und dürfen nicht Gegenstand<br />

eingeschränkten oder beschränkenden Handelns sein.<br />

Forschung und Entwicklung sollte immer im Geist der<br />

Freiheit von Wissenschaft und Technik vorgetrieben<br />

werden dürfen.<br />

Christopher Weßelmann<br />

– Chefredakteur –<br />

Editorial<br />

TMI und Lessons Learned – danach und für die Zukunft


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Did you know...?<br />

Growing Popular Support <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> in Belgium<br />

Forum Nucléaire, the association of the nuclear sector in Belgium,<br />

commissioned the 2019 edition of its regular opinion poll<br />

with Kantar TNS. Between 15 July and 6 September 2019<br />

756 Belgians over the age of 16 were interviewed by phone<br />

about their opinion on nuclear power and the nuclear sector<br />

in Belgium. The opinion poll shows a clear increase in support<br />

<strong>for</strong> nuclear power in Belgium compared to the 2017 edition<br />

of the poll, in particular with regard to the security of supply and<br />

the reduction of CO 2 -emissions. Below you can find a selection<br />

of results. The complete poll was published 18 October 2019<br />

on www.<strong>for</strong>umnucleaire.be and is available there in Dutch and<br />

French.<br />

For further details<br />

please contact:<br />

Nicolas Wendler<br />

KernD<br />

Robert-Koch-Platz 4<br />

10<strong>11</strong>5 Berlin<br />

Germany<br />

E-mail: presse@<br />

KernD.de<br />

www.KernD.de<br />

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

509<br />

For our country the production of nuclear energy in Belgium means …<br />

2017<br />

(n=1027)<br />

49 %<br />

13 %<br />

38 %<br />

2019<br />

(n=756)<br />

57 %<br />

More advantages than inconvenients 9 %<br />

Advantages<br />

and inconvenients equally<br />

32 %<br />

More inconvenients than advantages<br />

Are you favourable to keeping nuclear in Belgium <strong>for</strong> the production of electricity?<br />

2017<br />

(n=1027)<br />

30 %<br />

50 %<br />

19 % 2 %<br />

2019<br />

(n=756)<br />

46 %<br />

Yes, in long term too<br />

37 %<br />

Yes, but in short term (2025)<br />

16 %<br />

No<br />

1 %<br />

Don't know<br />

The current legislation <strong>for</strong>esees the closure of all nuclear power plants by 2025.<br />

Do you think that this can be realized without endangering the energy supply? (n=756)<br />

69 %<br />

No<br />

28 %<br />

Yes<br />

3 %<br />

Undecided,<br />

don't know<br />

It is good to replace nuclear power plants with gas fired power plants<br />

even if they emit much more CO 2 . (n=756)<br />

26 %<br />

I don't agree at all<br />

51 %<br />

I rather don't agree<br />

1 %<br />

I don't know<br />

<strong>11</strong> %<br />

I neither agree or disagree<br />

7 %<br />

I rather agree<br />

5 %<br />

I completely agree<br />

Figures in percent. Rounded values.<br />

Source:<br />

Forum Nucléaire<br />

Did you know...?


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

510<br />

Issue <strong>11</strong>/12 | 2019<br />

November/December<br />

CONTENTS<br />

Contents<br />

Editorial<br />

TMI and Lessons Learned – Afterwards<br />

and <strong>for</strong> the Future E/G 507<br />

Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . .509<br />

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

Key to Unlocking Bulgaria’s Belene Project<br />

is Finding Right Financing Package 512<br />

Feature | Environment and Safety<br />

Development of Per<strong>for</strong>mance Shaping Factors <strong>for</strong> Human Error<br />

Reduction during Reactor Decommissioning Activities through<br />

the Task Analysis Method 515<br />

Environment and Safety<br />

Root Causes of the Three Mile Island Accident 521<br />

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

The New Radiation Protection Law –<br />

Protection Against Radon G 525<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient<br />

in a Generic Three-Loops PWR-900 with TRACE Code<br />

Coupled with DAKOTA Uncertainty Analysis 526<br />

Experiment Research on the Insurge Transient Behavior<br />

of Gas-steam Pressurizer under Various Pressure 533<br />

Decommissioning and Waste Management<br />

Decommissioning & Dismantling of the Rossendorf<br />

Research Reactor RFR | Part 1 G 537<br />

First On-site Demonstration of Laser- based Decontamination<br />

Technology in Germany 543<br />

Report<br />

The 25 th Edition of the Frédéric Joliot/Otto Hahn Summer School<br />

on <strong>Nuclear</strong> Reactors “Physics, Fuels and Systems“ 549<br />

Special Topic | A Journey Through 50 Years AMNT<br />

Protection of Man and Environment – <strong>Nuclear</strong> Usage<br />

Outside of Energy Sector G 550<br />

KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .555<br />

Cover:<br />

Korea Kori NPP.<br />

Copyright: ©Korea Kori NPP<br />

News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .556<br />

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

Taking a Leaf out of Greta’s Climate Change Book 562<br />

G<br />

E/G<br />

= German<br />

= English/German<br />

Imprint 542<br />

Insert: INFORUM – Seminarprogramm 2020<br />

Contents


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Feature<br />

Environment and Safety<br />

515 Development of Per<strong>for</strong>mance Shaping<br />

Factors <strong>for</strong> Human Error Reduction<br />

during Reactor Decommissioning Activities<br />

through the Task Analysis Method<br />

5<strong>11</strong><br />

CONTENTS<br />

Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam<br />

Environment and Safety<br />

521 Root Causes of the Three Mile Island Accident<br />

Zoltan R. Rosztoczy<br />

Research and Innovation<br />

526 Evaluation of a Double-Ended Guillotine LBLOCA Transient<br />

in a Generic Three-Loops PWR-900 with TRACE Code<br />

Coupled with DAKOTA Uncertainty Analysis<br />

Andrea Bersano and Fulvio Mascari<br />

Decommissioning and Waste Management<br />

537 Decommissioning & Dismantling<br />

of the Rossendorf Research Reactor RFR | Part 1<br />

Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz<br />

543 First On-site Demonstration<br />

of Laser- based Decontamination Technology in Germany<br />

Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann,<br />

Wolfgang Lippmann and Antonio Hurtado<br />

Contents


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

512<br />

INSIDE NUCLEAR WITH NUCNET<br />

Key to Unlocking Bulgaria’s Belene<br />

Project is Finding Right Financing Package<br />

Last year the government decided to <strong>for</strong>mally revive the nuclear power station project. The process is now<br />

underway to select an investor.<br />

The History of Belene<br />

The project goes back to the mid-1970s when a<br />

two-unit nuclear station at Belene, on the<br />

Danube river in the north of the country, was planned by<br />

Bulgaria’s communist government. Site work began in<br />

1981 and construction of the first Russian-supplied<br />

VVER-1000 began in 1987. This was aborted in 1991 due to<br />

lack of funds and environmental and financial concerns.<br />

The project was revived in 2008, but <strong>for</strong>mally abandoned<br />

in 2012 after failing to find investors and with Sofia facing<br />

pressure from Washington and Brussels to limit its energy<br />

dependence on Russia, which was under contract to build<br />

the facility. The estimated €10bn cost of the project was<br />

too high to be funded by Bulgaria alone. In June 2018, the<br />

government decided to <strong>for</strong>mally revive the project. A call<br />

<strong>for</strong> interest was published and seven companies, three of<br />

which are Bulgarian, applied to invest in the project.<br />

Russia’s state-owned Rosatom, China’s state-owned CNNC<br />

and state-run Korea Hydro and <strong>Nuclear</strong> <strong>Power</strong> Company<br />

all filed applications confirming their interest. Press<br />

reports said that among the three Bulgarian companies<br />

there are two consortia with interests in energy that have<br />

been <strong>for</strong>med specifically <strong>for</strong> the Belene project and a third<br />

company, IPK-UP, which is registered in Kozloduy, but<br />

lacks any registered assets or income. The seventh<br />

company that filed its interest is Karlsruhe-registered<br />

Becktron-Liaz Technical Engineering.<br />

Bulgaria’s <strong>Nuclear</strong> Status<br />

Bulgaria, which joined the EU in 2007, closed four 440-MW<br />

VVER reactors at the Kozloduy nuclear station <strong>for</strong> safety<br />

reasons as part of its accession treaty with Brussels.<br />

Kozloduy has two newer 1,000-MW VVER reactors in commercial<br />

operation, supplying about 34 % of the country’s<br />

electricity. These two units have been renovated with<br />

financing from Euratom, the US ExportImport Bank,<br />

Citibank and Russia’s Roseximbank. Kozloduy-6 recently<br />

received a licence to operate <strong>for</strong> another 10 years until<br />

2029; Kozloduy-5 received a 10-year extension in 2017.<br />

Reviving the Belene Project<br />

The idea to revive the Belene project was born during<br />

negotiations with the European Union on Bulgaria’s<br />

accession. The centrist government at the time said Belene<br />

could replace generation that would be lost with the<br />

retirement of the four older Kozloduy units. In 2005, a<br />

socialist-led government revived Belene and a tendering<br />

procedure <strong>for</strong> a 2000-MW station was approved. A year<br />

later, Bulgaria chose an offer by a proposed consortium<br />

between Atomstroyexport (ASE) of Russia, Areva (now<br />

Framatome) of France and Siemens of Germany <strong>for</strong> the<br />

deployment of two Generation III VVER-1000 PWRs. In<br />

2008, Bulgaria signed a contract with ASE <strong>for</strong> the<br />

design, construction and commissioning of the two units.<br />

However, in 2012 a new government headed by current<br />

prime-minister Boyko Borissov cancelled Belene after<br />

failing to find financing. Russia took Bulgaria to international<br />

arbitration over the cancellation.<br />

Bulgaria again began looking <strong>for</strong> ways to revive Belene<br />

after it lost the arbitration and paid € 600 m in 2016 in<br />

compensation to ASE <strong>for</strong> components which had already<br />

been ordered. The 2016 decision by the Geneva-based<br />

<strong>International</strong> Court of Arbitration of the <strong>International</strong><br />

Chamber of Commerce said Bulgarian state energy<br />

company Nationalna Elektricheska Kompania (NEK) had to<br />

pay € 620 m to ASE and assume ownership of the components.<br />

The price was reduced after interest adjustments.<br />

By the beginning of 2018 Russia had delivered most of the<br />

equipment that was at the centre of the court ruling,<br />

including two reactor pressure vessels, the RPV heads, the<br />

full steam generator sets and the main pipelines. The<br />

equipment is stored at the Belene site.<br />

A Question Of Financing<br />

According to energy minister Temenuzhka Petkova Bulgaria<br />

wants to build the two units within 10 years and at a cost of<br />

up to € 10 bn. However, the exact nature of how the station<br />

will be financed remains unclear. Ms Petkova told parliament<br />

that the government would like state energy company<br />

NEK to have a “blocking stake” – potentially giving it a veto if<br />

necessary – in the project, but only by contributing existing<br />

assets and infrastructure at the Belene site, valued at<br />

€ 1.5 bn. The government’s policy is to attract private investment<br />

<strong>for</strong> the project with no state guarantees or long-term<br />

electricity purchase contracts. Critics have said that would<br />

be difficult given the magnitude of nuclear new-build<br />

projects.<br />

The Need For <strong>Nuclear</strong><br />

Bulgaria says it needs nuclear because it is heavily dependent<br />

on imported primary energy resources and uses the most<br />

electricity relative to GDP in the EU. Bulgaria imports almost<br />

100 % of its oil and gas, 100 % of its nuclear fuel and about<br />

35 to 40 % of its coal. It exports electricity to neighbouring<br />

countries, but the electricity sector is dependent on imports<br />

of primary energy, mainly from Russia. Russia sees Bulgaria<br />

as a transit point into the rest of Europe <strong>for</strong> Russian energy<br />

sources because it bypasses Ukraine.<br />

Bulgaria is under pressure by the EU to reduce carbon<br />

emissions in line with the bloc’s climate ambitions. Bulgaria<br />

generates about 40 % of its electricity from coal and new<br />

nuclear could replace a major proportion of this. The government<br />

has also hinted that new reactors might be needed<br />

to replace the two Kozloduy units, which are scheduled to<br />

be retired in 2027 and 2029. A report by the Bulgarian<br />

Academy of Science said electricity demand in Bulgaria is<br />

expected to increase after 2030 and new nuclear could be<br />

economically viable if construction costs are kept in check.<br />

What Happens Next?<br />

Mr Borissov’s ruling party and its allies command a<br />

com<strong>for</strong>table majority so plans to build new units at Belene<br />

would not appear to pose a political problem. The<br />

opposition socialists also support nuclear energy. Public<br />

support is high, with almost two-thirds of Bulgarians<br />

voting in support of nuclear in a 2013 referendum,<br />

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

Key to Unlocking Bulgaria’s Belene Project is Finding Right Financing Package


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

although the result was not binding because of the low<br />

turnout. The EU and the US have urged Bulgaria to take<br />

stock of its dependence on Russian gas and oil and<br />

encouraged it to seek Western investors <strong>for</strong> Belene. The<br />

key is to find the right financing package. China is said to<br />

be asking <strong>for</strong> financial guarantees from Sofia in return <strong>for</strong><br />

its participation, but Sofia favours the creation of some<br />

<strong>for</strong>m of public- private partnership. Any financing scheme<br />

will need to be approved by the EU’s state aid watchdog.<br />

A risk analysis of the Belene project, prepared by the<br />

Vienna <strong>International</strong> Centre <strong>for</strong> <strong>Nuclear</strong> Competence <strong>for</strong><br />

the Bulgarian Atomic Forum, recommended that the<br />

government reconsider its position of non-participation in<br />

the project. Globally, there is no precedent in which nuclear<br />

projects are implemented without state involvement, the<br />

analysis said. Different <strong>for</strong>ms and levels of participation are<br />

possible, including direct sponsorship, loan guarantees, tax<br />

credits, long-term energy purchase agreements and price<br />

differences, the analysis noted. Another option is to use the<br />

Belene equipment <strong>for</strong> a new unit at Kozloduy, but this<br />

would incur additional costs. Abandoning the project and<br />

selling the equipment as scrap would result in a loss of<br />

about $ 1.5 m. The sale of the project is also a possibility to<br />

companies from Russia or China.<br />

Whoever the project owner and investors are, the<br />

government has said the main contractor <strong>for</strong> the nuclear<br />

part of the project will be Russia’s Rosatom, because of the<br />

nature of the technology already paid <strong>for</strong> by Bulgaria.<br />

Turbine hall equipment and instrumentation and control<br />

systems could be supplied by US-based General Electric<br />

and France’s Framatome. Both companies participated in<br />

the recent tender, but as equipment suppliers rather than<br />

investors.<br />

Author<br />

David Dalton<br />

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

Key to Unlocking Bulgaria’s Belene Project is Finding Right Financing Package


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

CALENDAR 514<br />

Calendar<br />

2020<br />

12.01. – 16.01.2020<br />

<strong>Power</strong> Plant Simulation Conference. Chattanooga,<br />

Tennessee United States, Society <strong>for</strong> Modeling &<br />

Simulation <strong>International</strong>, www.scs.org<br />

13.01. – 14.01.2020<br />

ICNPPS 2020 – 14 th <strong>International</strong> Conference on<br />

<strong>Nuclear</strong> <strong>Power</strong> Plants Safety. Zurich, Switzerland,<br />

Waset, www.waset.org<br />

20.01. – 21.01.2020<br />

6 th Central & Eastern Europe <strong>Nuclear</strong> Industry<br />

Congress 2020. Prague, Czech Republic, SZWGroup,<br />

www.szwgroup.com<br />

22.01.2020<br />

31.03. – 03.04.2020<br />

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

Advances in Thermal Hydraulics. Paris, France,<br />

Société Francaise d’Energie Nucléaire (SFEN),<br />

www.sfen-ath2020.org<br />

19.04. – 24.04.2020<br />

<strong>International</strong> Conference on Individual<br />

Monitoring. Budapest, Hungary, EUROSAFE,<br />

www.eurosafe-<strong>for</strong>um.org<br />

20.04. – 22.04.2020<br />

World <strong>Nuclear</strong> Fuel Cycle 2020. Stockholm,<br />

Sweden, WNA World <strong>Nuclear</strong> Association,<br />

www.world-nuclear.org<br />

06.06. – 12.06.2020<br />

ATALANTE 2020. Montpellier, France, CEA,<br />

www.atalante2020.org<br />

08.06. – 12.06.2020<br />

20 th WCNDT – World Conference on<br />

Non-Destructive Testing. Seoul, Korea, EPRI,<br />

www.wcndt2020.com<br />

15.06. – 19.06.2020<br />

<strong>International</strong> Conference on <strong>Nuclear</strong> Knowledge<br />

Management and Human Resources Development:<br />

Challenges and Opportunities. Moscow,<br />

Russian Federation, IAEA, www.iaea.org<br />

15.06. – 20.07.2020<br />

<strong>Nuclear</strong> Fuel Supply Forum. Washington, D.C., USA,<br />

NEI, www.nei.org<br />

WNU Summer Institute 2020. Japan, World <strong>Nuclear</strong><br />

University, www.world-nuclear-university.org<br />

10.02. – 14.02.2020<br />

01.09. – 04.09.2020<br />

37 th Short Courses on Multiphase Flow. Zurich,<br />

Switzerland, Swiss Federal Institute of Technology<br />

ETH, www.lke.mavt.ethz.ch<br />

IGORR – Standard Cooperation Event in the <strong>International</strong><br />

Group on Research Reactors Conference.<br />

Kazan, Russian Federation, IAEA, www.iaea.org<br />

10.02. – 14.02.2020<br />

09.09. – 10.09.2020<br />

ICONS2020: <strong>International</strong> Conference on <strong>Nuclear</strong><br />

Security. Vienna, Austria, The <strong>International</strong> Atomic<br />

Energy Agency (IAEA), www.iaea.org<br />

02.03. – 06.03.2020<br />

<strong>International</strong> Workshop on Developing a National<br />

Framework <strong>for</strong> Managing the Response to<br />

<strong>Nuclear</strong> Security Events. Madrid, Spain, IAEA,<br />

www.iaea.org<br />

08.03. – 12.03.2020<br />

WM Symposia – WM2019. Phoenix, AZ, USA,<br />

www.wmsym.org<br />

08.03. – 13.03.2020<br />

IYNC2020 – The <strong>International</strong> Youth <strong>Nuclear</strong><br />

Congress. Sydney, Australia, IYNC, www.iync2020.org<br />

18.03. – 20.03.2020<br />

12. Expertentreffen Strahlenschutz. Bayreuth,<br />

Germany, TÜV SÜD, www.tuev-sued.de<br />

25.03. – 27.03.2020<br />

H2020 McSAFE Training Course. Eggenstein-<br />

Leopoldshafen, Germany, Karlsruher Institute für<br />

Technologie (KIT), www.mcsafe-h2020.eu<br />

29.03. – 02.04.2020<br />

PHYSOR2020 — <strong>International</strong> Conference on<br />

Physics of Reactors 2020. Cambridge, United<br />

Kingdom, <strong>Nuclear</strong> Energy Group,<br />

www.physor2020.com<br />

31.03. – 02.04.2020<br />

4 th CORDEL Regional Workshop on<br />

Harmonization to support the Operation and<br />

New Build fo NPPs including SMRs. Lyon, France,<br />

NUGENIA, www.nugenia.org<br />

KERNTECHNIK 2020.<br />

Berlin, Germany, KernD and KTG,<br />

www.kerntechnik.com<br />

05.05. – 06.05.2020<br />

10.05. – 15.05.2020<br />

ICG-EAC Annual Meeting 2020. Helsinki, Finland,<br />

ICG-EAC, www.icg-eac.org<br />

<strong>11</strong>.05. – 15.05.2020<br />

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

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

www.iaea.org<br />

12.05. – 13.05.2020<br />

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

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

www.tuev-sued.de<br />

17.05. – 22.05.2020<br />

BEPU2020, Giardini Naxos. Sicily, Italy, NINE,<br />

www.nineeng.com<br />

18.05. – 22.05.2020<br />

SNA+MC2020 – Joint <strong>International</strong> Conference on<br />

Supercomputing in <strong>Nuclear</strong> Applications + Monte<br />

Carlo 2020, Makuhari Messe. Chiba, Japan, Atomic<br />

Energy Society of Japan, www.snamc2020.jpn.org<br />

20.05. – 22.05.2020<br />

<strong>Nuclear</strong> Energy Assembly. Washington, D.C., USA,<br />

NEI, www.nei.org<br />

31.05. – 03.06.2020<br />

13 th <strong>International</strong> Conference of the Croatian<br />

<strong>Nuclear</strong> Society. Zadar, Croatia, Croatian <strong>Nuclear</strong><br />

Society, www.nuclear-option.org<br />

VGB Congress 2020 – 100 Years VGB. Essen,<br />

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

09.09. – <strong>11</strong>.09.2020<br />

World <strong>Nuclear</strong> Association Symposium 2020.<br />

London, United Kingdom, WNA World <strong>Nuclear</strong><br />

Association, www.world-nuclear.org<br />

16.09. – 18.09.2020<br />

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

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

www.fibiccs.org<br />

16.09. – 18.09.2020<br />

<strong>International</strong> <strong>Nuclear</strong> Reactor Materials<br />

Reliability Conference and Exhibition.<br />

New Orleans, Louisiana, USA, EPRI, www.snetp.eu<br />

28.09. – 01.10.2020<br />

NPC 2020 <strong>International</strong> Conference on <strong>Nuclear</strong><br />

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

Française d’Energie Nucléaire,<br />

www.sfen-npc2020.org<br />

28.09. – 02.10.2020<br />

Jahrestagung 2020 – Fachverband Strahlenschutz<br />

und Entsorgung. Aachen, Germany, Fachverband<br />

für Strahlenschutz, www.fs-ev.org<br />

26.10. – 30.10.2020<br />

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

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

09.<strong>11</strong>. – 13.<strong>11</strong>.2020<br />

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

Improving Radiation Protection in Practice.<br />

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

30.03. – 01.04.2020<br />

INDEX <strong>International</strong> <strong>Nuclear</strong> Digital Experience.<br />

Paris, France, SFEN Société Française d’Energie<br />

Nucléaire, www.sfen-index2020.org<br />

This is not a full list and may be subject to change.<br />

Calendar


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Feature | Environment and Safety<br />

Development of Per<strong>for</strong>mance Shaping<br />

Factors <strong>for</strong> Human Error Reduction during<br />

Reactor Decommissioning Activities<br />

through the Task Analysis Method<br />

Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam<br />

In this study, we present the process of deriving Per<strong>for</strong>mance Shaping Factors (PSFs) and developing a classification<br />

system <strong>for</strong> them to account <strong>for</strong> human errors that may occur during the decommissioning activities due to the shutdown<br />

of the Kori <strong>Nuclear</strong> <strong>Power</strong> Plant Unit 1. Furthermore, Human Reliability Analysis (HRA) <strong>for</strong> the reduction of human<br />

errors was per<strong>for</strong>med on the Reactor Pressure Vessel Internal (RPVI) cutting work. Task Analysis was conducted on the<br />

RPVI decommissioning activities and possible human errors were identified. The PSF selection criteria that affect the<br />

decommissioning activities were set up based on the human errors identified through the results of the Task Analysis<br />

and review of the PSFs applied to the field of nuclear energy power generation. Finally, the PSFs were derived and their<br />

classification system was developed.<br />

1 Introduction<br />

In recent years, industries have experienced a reduction in<br />

the rate of accidents due to technical problems because of<br />

the development of technologies <strong>for</strong> accident prevention.<br />

However, it is difficult to evaluate the reliability of such<br />

systems without considering the impact of human errors<br />

by their operators. In particular, according to real statistical<br />

accident data, about 60 to 90 % of all accidents are due to<br />

human errors, while the remainder is due to technical<br />

errors in the system [1-3]; thus, it is evident that even<br />

minor operator errors in real life can severely undermine<br />

the operational per<strong>for</strong>mance of the system. Furthermore,<br />

in large or crucial systems, such as large power plants<br />

( including nuclear power plants), air transportation, and<br />

railways, human error can lead to large and more harmful<br />

accidents; there<strong>for</strong>e, the reduction of human errors in the<br />

case of such systems requires significant attention [4].<br />

At present, Korea is preparing to safely and economically<br />

decommission Unit 1 of the Kori <strong>Nuclear</strong> <strong>Power</strong> Plant,<br />

whose operation was permanently suspended in June<br />

2017. Various technological developments are being<br />

conducted <strong>for</strong> the decommissioning of Kori Unit 1 with the<br />

primary objective of ensuring the safety of human life and<br />

property and improving economic efficiency. There<strong>for</strong>e, it<br />

is important to establish countermeasures to decrease<br />

human errors during this decommissioning, because these<br />

human errors not only hinder the safety of the operators,<br />

but also affect the economics of the project by causing<br />

delays in the decommissioning schedule. In order to do so,<br />

human error trends and influencing factors need to be<br />

identified through Human Reliability Analysis (HRA) [1];<br />

in particular, HRA is used to identify, model, and quantify<br />

the likelihood of human error [3]. Figure 1 shows the<br />

flowchart <strong>for</strong> HRA that needs to be applied <strong>for</strong> the cutting<br />

of the Reactor Pressure Vessel Internal (RPVI) components,<br />

which is the decommissioning activity targeted in this<br />

study.<br />

It is important to conduct an HRA <strong>for</strong> the cutting<br />

operation of RPVI components, because they have the<br />

highest levels of radioactivity among all the other parts of<br />

the unit, and this operation is highly susceptible to human<br />

errors because of the use of remote equipment, among<br />

others. There<strong>for</strong>e, in particular, the objective of this study<br />

is to obtain the Per<strong>for</strong>mance Shaping Factors (PSFs) that<br />

| Fig. 1.<br />

Procedure to per<strong>for</strong>m the HRA in the decommissioning activity [5].<br />

affect RPVI cutting and develop a classification system<br />

<strong>for</strong> these PSFs.<br />

Task characteristics, procedures, and in<strong>for</strong>mation on<br />

the decommissioning of RPVI components were identified<br />

by reviewing related domestic and <strong>for</strong>eign literature as<br />

well as by examining previous cases of decommissioning of<br />

overseas nuclear power plants. In addition, task analysis<br />

<strong>for</strong> the decommissioning of the RPVI was conducted<br />

to identify probable human errors. First, potential PSFs<br />

were identified <strong>for</strong> this study based on a combination of<br />

the expected human errors from the detailed task<br />

analysis and the general PSFs from the evaluation of<br />

PSFs in nuclear facilities. Then, the selection criteria <strong>for</strong><br />

the PSFs affecting the decommissioning activities were<br />

established. Finally, the PSFs were derived to construct<br />

the classification system.<br />

FEATURE | ENVIRONMENT AND SAFETY 515<br />

Feature<br />

Development of Per<strong>for</strong>mance Shaping Factors <strong>for</strong> Human Error Reduction during Reactor Decommissioning Activities through the Task Analysis Method ı Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

FEATURE | ENVIRONMENT AND SAFETY 516<br />

Task<br />

Task In<strong>for</strong>mation<br />

and Requirements<br />

Task to Consider<br />

Working Time<br />

Teamwork and<br />

Communication<br />

Workload<br />

Task Support<br />

Workplace Factors<br />

Hazard Identification<br />

Expected PSFs<br />

2 Task Analysis<br />

Task Analysis (TA) is the initial step in the case of human<br />

error assessment. It provides the characteristics of different<br />

tasks, their vulnerabilities, and properties by understanding<br />

the objectives, per<strong>for</strong>mance methods, scopes of<br />

the tasks, and procedures involved [6]. In addition, TA<br />

helps eliminate the conditions that can cause errors be<strong>for</strong>e<br />

they occur by providing detailed in<strong>for</strong>mation about the<br />

tasks as well as other in<strong>for</strong>mation <strong>for</strong> predicting and<br />

preventing errors.<br />

2.1 TA Method<br />

Although several TA methods exist, considering the<br />

physical characteristics (complexity) and Human Machine<br />

Interface (HMI) aspect (using remote devices) of the<br />

decommissioning activities, we utilized the Hierarchical<br />

TA (HTA) method.<br />

HTA is a systematic and detailed TA method that is used<br />

to achieve task objectives. It is appropriate <strong>for</strong> not only<br />

identifying detailed task configuration and conditions, but<br />

also expressing complex task steps as a simple, hierarchical<br />

structure. Furthermore, the HTA method was used because<br />

it can easily describe the characteristics of the work<br />

involved and identify significant in<strong>for</strong>mation about the<br />

Title<br />

| Tab. 1.<br />

Details of tasks to be per<strong>for</strong>med <strong>for</strong> TA.<br />

• Task influencing factors<br />

(cutting size, number of cutting operations, precision, etc.)<br />

• Task output requirements<br />

• Record feedback to indicate adequacy of action taken<br />

• Alarms and warnings<br />

• Actions to be taken<br />

• Equipment needed (type, size, per<strong>for</strong>mance,<br />

required utility, equipment usage constraint, etc.)<br />

• Task frequency and required accuracy<br />

• Physical position of the operator<br />

(standing, sitting, squatting, etc.)<br />

• Biomechanics<br />

• Movement<br />

(lifting, pushing, rotating, pulling, swaying, etc.)<br />

• Force required<br />

• Unit working time based on “work contents”<br />

• Additional work hours taking into account “work support”<br />

and “workplace environmental conditions”<br />

• Number of work shifts and workers per work shift<br />

• Cooperation required between the teams per<strong>for</strong>ming<br />

the work<br />

• Personal communication <strong>for</strong> monitoring or<br />

taking control actions<br />

• Cognitive workload<br />

• Physical workload<br />

• Overlap of task requirements (serial versus parallel task<br />

elements)<br />

• Special and protective clothing <strong>for</strong> work<br />

• Job aids, procedures, or reference materials needed<br />

• Required auxiliary tools and equipment<br />

• Ingress and egress paths to the work site<br />

• Workspace required to per<strong>for</strong>m the task<br />

• Typical workplace environmental conditions<br />

(e.g. lighting, temperature, noise, etc.)<br />

• Work breaks taking into account “work support” and<br />

“ workplace environmental conditions”<br />

• Identify work-related hazards,<br />

e.g. potential personal injury hazards<br />

Examples include:<br />

• Stress<br />

• Time pressure (critical path operations)<br />

• Extreme environmental conditions<br />

• Reduced staffing<br />

HMI, communication and decision making processes, as<br />

well as possible accidents. The HTA method involves<br />

describing the manner in which tasks need to be per<strong>for</strong>med<br />

after establishing their overall objectives and classifying<br />

them into their sub-tasks [6].<br />

Using a tabular <strong>for</strong>mat to per<strong>for</strong>m HTA allows one to<br />

express complex tasks that require significant skill in a<br />

suitable manner, because one can include detailed notes, if<br />

necessary. In this study, we comprehensively reviewed<br />

various items, such as HMI, Communication, Time, and<br />

Accident, <strong>for</strong> the decommissioning activity in a tabular<br />

<strong>for</strong>mat; this is shown in Table 1 [5].<br />

2.2 Target Decommissioning Activity <strong>for</strong> TA<br />

In general, one of the most challenging tasks during plant<br />

decommissioning is believed to be the removal of the<br />

highly radioactive internal components of the reactor pressure<br />

vessel (RPV); this is true <strong>for</strong> Kori Unit 1 as well. In<br />

addition, another reason that this is one of the most<br />

difficult activities is because these radioactive components<br />

must be dismantled and cut underwater owing to the<br />

severe radiological conditions of the RPVI components<br />

[7-8]. There<strong>for</strong>e, it is recommended that the reactor<br />

internals be removed as early as possible in the plant<br />

dismantling sequence, so that these water systems and<br />

their associated support systems can be released <strong>for</strong><br />

decommissioning, which minimizes the costs associated<br />

with maintaining these systems in operation after<br />

permanent plant shutdown [8].<br />

The cross-section of the RPV with the primary internal<br />

components at Kori Unit 1 is shown in Figure 2. As can be<br />

seen from the figure, the internal structures adjacent to the<br />

core barrel active region are the most highly activated, and<br />

in most cases, include intermediate level waste components<br />

that might require removal prior to the disposal of the<br />

remainder of the RPV and reactor internal components<br />

[7]. Thus, in this study, this RPV internal segmentation<br />

activity is selected from among the various dismantling<br />

activities in Kori Unit 1.<br />

Furthermore, in this study, the TA was per<strong>for</strong>med <strong>for</strong><br />

the most complex and highly radioactive RPVI cutting task<br />

among various disassembly activities by using the HTA<br />

method. The sequence of operations <strong>for</strong> each sub-activity<br />

in this target task is listed in Table 2.<br />

2.3 TA Results<br />

The TA <strong>for</strong> the RPVI Dismantling Activity was per<strong>for</strong>med as<br />

per the activities listed in Table 2 based on the items listed<br />

in Table 1. In our study, this analysis was per<strong>for</strong>med <strong>for</strong><br />

each of the 10 sub-activities. The summary of the TA <strong>for</strong><br />

the RPVI Dismantling Activity is given as follows.<br />

p In general, in the decommissioning of nuclear power<br />

plants, the cutting of parts in the RPVI is the most<br />

complicated and difficult task during the dismantling<br />

process. There<strong>for</strong>e, the influence on the internal factors<br />

of the workers was evaluated in order to have a relatively<br />

high value in terms of operator internal response.<br />

p This kind of dismantling operation, which is complex<br />

and requires accuracy and reliability, is significantly<br />

influenced by the internal and external characteristics<br />

of the worker. There<strong>for</strong>e, sufficient education and<br />

training is required. However, as the driving principle<br />

and operation method of these cutting equipment and<br />

accessories ( cutting equipment, remote control device,<br />

display, auxiliary equipment, etc.) are not complicated<br />

and operation is relatively simple, an operator is not<br />

required to have considerable experience in using them.<br />

Feature<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

No Sub activity Activity Description<br />

1 Control Rod Guide Tube Upper Area Cutting<br />

and Packaging<br />

2 Control Rod Guide Tube Lower Area Cutting<br />

and Packaging<br />

1) Lift up the Upper Internal Assembly and fix it to the turntable<br />

2) Cut the control rod guide tube upper area<br />

3) Transfer and package the cut section to a storage container<br />

1) Cut the control rod guide tube lower area<br />

2) Transfer and package the cut section to a storage container<br />

3 Upper Plate Cutting and Packaging 1) Move and release the Upper Support Plate in an empty space in the reactor tank<br />

2) Cut the Upper Core Plate fixed to the turntable<br />

3) Fix the Upper Core Plate to the turntable and cut the Upper Support Plate<br />

4) Transfer and package the cut pieces to a storage container<br />

4 Baffle Fixed Bolt Head Cutting 1) Install the mechanical drill <strong>for</strong> Baffle Separation<br />

2) Cut the Baffle Fixing Bolt Head by placing the mechanical drill inside the RPV<br />

5 Baffle Cutting and Packaging 1) Lift the Baffle and remove it from the Former<br />

2) After fixing it to the turntable, cut the Baffle<br />

3) Transfer and package the cut Baffle pieces to a storage container<br />

6 Core Barrel Lower Area Cutting 1) Lift up the Lower Internal Assembly and fix it to the turntable<br />

2) Rotate the turntable to cut the Lower Internal Assembly<br />

3) Lift the upper area of the Lower Internal Assembly into the Vessel<br />

7 Lower Internal Structure Assembly Cutting<br />

and Packaging<br />

8 Thermal Shield Separation, Cutting and<br />

Packaging<br />

1) Cut the Instrument Nozzle from the Core Support Structure Assembly<br />

2) Transfer and package the cut nozzle to a storage container<br />

3) Cut and package the tie plate fixed to the Turntable<br />

4) Fix the Lower Core Plate to the Turntable and cut it<br />

5) Transfer and package the cut pieces to a storage container<br />

1) Lift the upper area of the Lower Internal Assembly and fix it to the turntable<br />

2) Separate the Thermal Shield by cutting the Bolt Head<br />

3) Release the removed Core barrels from the Thermal Shield inside the vessel<br />

4) After fixing the Thermal Shield to the Turntable, remove the Irradiation Specimen Guide<br />

5) Cut the Turntable Thermal Shield Upper and Lower Panels<br />

9 Former Separation 1) Lift the Core Barrel to the turntable and fix it<br />

2) Cutting the Former fixing bolt head outside the Core Barrel<br />

3) Separate the Former from the Core Barrel<br />

4) Transfer and package the separated Former to a storage container<br />

10 Core Barrel Cutting and Packaging 1) Fix the Core Barrel to the turntable and cut it<br />

2) Temporarily release the cut Upper Core Barrel in the Vessel<br />

3) Segment the Lower part of the Core Barrel fixed to the turntable<br />

4) Transfer and package the cut pieces to a storage container<br />

5) Repeat the procedure <strong>for</strong> the cutting the remaining Core Barrel<br />

FEATURE | ENVIRONMENT AND SAFETY 517<br />

| Tab. 2.<br />

Task Description of RPVI Dismantling Activity.<br />

p Considering the characteristics of the work (equipment<br />

and facilities, the object to be cut, and clothing <strong>for</strong><br />

radiation protection in the work environment), a<br />

detailed work plan must be established in advance.<br />

Further, as this cutting work is time-consuming, the<br />

psychological and physical influences that the supervisor<br />

and the worker can receive are considerable.<br />

p If the cutting activity is dangerous, takes a long time,<br />

and has associated time constraints, it should be<br />

per<strong>for</strong>med during the day/night time. In this case,<br />

various difficulties (such as break time, clothing<br />

discom<strong>for</strong>t, and physio logical factors) are generated.<br />

Because these difficulties have a significant impact on<br />

the internal and external factors affecting the worker,<br />

much cooperation and communication is required<br />

between the worker and the supervisor in this working<br />

environment.<br />

p The radiation and the physical environment of the<br />

workplace are the major risk factors <strong>for</strong> the workers,<br />

and the influence of these working environments on the<br />

internal and external factors of the workers was<br />

con siderable.<br />

p <strong>Nuclear</strong> decommissioning work is not a frequently<br />

per<strong>for</strong>med task. There<strong>for</strong>e, workers may have insufficient<br />

experience and education/training. There<strong>for</strong>e,<br />

after the decommissioning activity has been carried<br />

out, it is necessary to feedback the results of the work to<br />

be reflected in the necessary work procedures and to be<br />

managed as experience data.<br />

| Fig. 2.<br />

Internal components <strong>for</strong> the RPV in Kori Unit 1 [7].<br />

3 PSFs in the Decommissioning Activities<br />

Because the workers’ activities are the fundamental factors<br />

that renders the system vulnerable, it is necessary to<br />

identify, model, and quantify the possibility of human<br />

error using HRA [3]. In particular, the nominal Human<br />

Error Probability (HEP) used in the HRA is estimated<br />

Feature<br />

Development of Per<strong>for</strong>mance Shaping Factors <strong>for</strong> Human Error Reduction during Reactor Decommissioning Activities through the Task Analysis Method ı Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

FEATURE | ENVIRONMENT AND SAFETY 518<br />

based on the workers’ activity, most often based on the<br />

PSFs. This is because PSFs characterize the important<br />

aspects of human error and provide numerical criteria <strong>for</strong><br />

adjusting the average HEP level [9]. There<strong>for</strong>e, the key<br />

step in the HRA <strong>for</strong> decommissioning activities of a nuclear<br />

power plant is to select appropriate PSFs, develop the<br />

classification system, and quantify their impacts on the<br />

decommissioning activity.<br />

3.1 Criteria <strong>for</strong> selecting PSFs<br />

When selecting PSFs that affect the decommissioning<br />

activities of nuclear plants, it is important to carefully<br />

select all the important situational factors that affect the<br />

hazard level so as to not miss any of the factors. In addition,<br />

it should be ensured that there is no overlap in the meaning<br />

and scope of the PSFs; furthermore, the factors should be<br />

selected based on their actual effect on human error<br />

analysis. There<strong>for</strong>e, the PSFs are selected <strong>for</strong> the HRA of<br />

the decommissioning activities of the nuclear power plant<br />

considering the following criteria.<br />

p The PSFs should be selected to appropriately reflect the<br />

work characteristics. There<strong>for</strong>e, the TA results of the<br />

work should be fully reflected by the PSFs.<br />

p The internal and external factors that might affect<br />

the workers’ per<strong>for</strong>mance should be comprehensively<br />

considered.<br />

p When a worker encounters an abnormal event, the<br />

internal factors should be included in the PSFs<br />

appro priately to account <strong>for</strong> the reactions that occur<br />

naturally in workers, including physical, cognitive, and<br />

emotional factors.<br />

p The external factors should comprehensively include<br />

those that directly or indirectly affect the workers’<br />

response to per<strong>for</strong>mance, i.e., the business organization<br />

and work environment factors, among others.<br />

p Based on previous research results, in our case, the<br />

PSFs will consist of three levels (namely, Level 1, Level<br />

2, and Level 3). Though there are direct and indirect<br />

dependencies between the PSFs in Level 1 and Level 2,<br />

the PSFs are selected in a manner that there is no<br />

dependency between the Level 3 PSFs.<br />

p In practice, the PSFs are selected based on factors<br />

that directly affect the trends of the ongoing events.<br />

There<strong>for</strong>e, factors such as workers’ training and work<br />

shifts, among others, indirectly affecting the work<br />

per<strong>for</strong>mance of the workers, are excluded from the PSF<br />

Classification Direct effect Indirect effect<br />

Personal Factors<br />

System factors<br />

Task factors<br />

| Tab. 3.<br />

Derived PSF factor.<br />

• Duration of mental stress<br />

• Mental tension<br />

• Pain or discom<strong>for</strong>t<br />

• Hunger or thirst<br />

• Emotional state<br />

• Duration of physical stress<br />

• Disruption of circadian rhythm<br />

• Lack of sleep<br />

• Work hours<br />

• Shift rotation<br />

• Suddenness of onset<br />

• State of current practice or skill<br />

• Motivation and attitudes<br />

• Personality and<br />

intelligence variables<br />

knowledge of required<br />

per<strong>for</strong>mance previous<br />

training/experience<br />

• Sensory deprivation<br />

• Distractions (noise, glare,<br />

movement, flicker. color)<br />

• Complexity<br />

• Movement constriction<br />

• Workplace layout<br />

• Threats of failure<br />

• Lack of physical exercise<br />

• High jeopardy risk<br />

• Conflicts of motives<br />

about job per<strong>for</strong>mance<br />

selection because they are considered to be inherent in<br />

the other PSFs themselves.<br />

p The PSFs should have clear definitions so that their<br />

meanings and roles do not overlap. There<strong>for</strong>e, as much<br />

as possible, the scale of the PSFs should be reduced by<br />

grouping all the PSFs and reducing the number of tasks<br />

involved.<br />

3.2 Selection and definition of PSFs<br />

Based on our review of existing literature related to the<br />

work per<strong>for</strong>mances of individual workers from various<br />

industrial fields, we observed that it is not easy to find<br />

a consensus on the factors influencing the workers’<br />

per<strong>for</strong>mance. However, considering the results of these<br />

previous studies, it is deemed that the possibility <strong>for</strong> human<br />

error can be determined based on the degradation of the<br />

workers’ human error factor, i.e., their task per<strong>for</strong>mance.<br />

The Institute of <strong>Nuclear</strong> <strong>Power</strong> Operations (INPO)<br />

presents 85 error precursors that can lead to possible<br />

human errors considering the business requirements, personal<br />

abilities, work environments, and human nature in<br />

terms of the operation of the nuclear power plants [10].<br />

These precursors are considered as the risk variables <strong>for</strong><br />

human errors made by workers.<br />

There<strong>for</strong>e, it is realistic to utilize an Error Precursor,<br />

i.e., PSF, that affects work per<strong>for</strong>mance. While estimating<br />

the probability of human error, it is desirable to add<br />

psychological factors that cannot be directly managed<br />

after the possibility of human error <strong>for</strong> manageable<br />

factors is reduced through safety management or accident<br />

prevention activities. The PSFs extracted <strong>for</strong> use in the<br />

human error assessment model in this study are listed<br />

in Table 3 below.<br />

3.3 Development of a classification system<br />

<strong>for</strong> PSFs<br />

TA was conducted to derive the PSFs <strong>for</strong> RPVI cutting,<br />

which is one of the primary tasks involved in the<br />

decommissioning of nuclear power plants. Because it is<br />

difficult to select PSFs such that their meanings and roles<br />

are clearly delineated, the PSFs are classified into three<br />

major categories, which are defined by their task analysis<br />

results. In other words, the human factors of the workers<br />

themselves, the operating system factors related to the<br />

work, and the ergonomic factors linking the worker and<br />

the dismantling activities were suitably classified.<br />

In particular, an important aspect of the TA is “Human,”<br />

i.e., a worker who per<strong>for</strong>ms the task of decommissioning<br />

the RPV and can cause human errors. It is noteworthy<br />

that human factors are important to consider not only<br />

in the field of nuclear power generation, but also, in<br />

other industrial fields, such as the railways and aviation<br />

[<strong>11</strong>-16].<br />

p Supervisors and workers might be affected psychologically<br />

while per<strong>for</strong>ming the tasks described above.<br />

Because the cutting process takes considerable time, it<br />

is important to consider the various stresses and<br />

emotional conditions that supervisors and workers<br />

might experience.<br />

p In particular, the work environment and conditions <strong>for</strong><br />

the cutting task in normal or abnormal situations affect<br />

the workers’ physical and physiological factors. Thus, to<br />

per<strong>for</strong>m the RPVI cutting tasks, wearing the appropriate,<br />

protective work clothes and protective cap<br />

and having good physical and health conditions to<br />

carry or handle the equipment are considered important.<br />

Furthermore, while per<strong>for</strong>ming the tasks, the<br />

Feature<br />

Development of Per<strong>for</strong>mance Shaping Factors <strong>for</strong> Human Error Reduction during Reactor Decommissioning Activities through the Task Analysis Method ı Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

physiological factors of the workers, such as fatigue,<br />

hunger, and excretion, is equally important.<br />

p When the nuclear power plant is dismantled <strong>for</strong> the<br />

first time, there might be a lack of personal skills and<br />

experiences, because the workers are not accustomed<br />

to the decommissioning work. There<strong>for</strong>e, in<strong>for</strong>mation<br />

about the complete decommissioning task and equipment<br />

used <strong>for</strong> each task should be taught through<br />

training and education.<br />

Next, there is a need <strong>for</strong> an operating system that can be<br />

managed, supervised, and overseen to efficiently per<strong>for</strong>m<br />

the decommissioning tasks. There<strong>for</strong>e, the “operation”<br />

elements, such as organization, including the supervisors<br />

and workers; task management <strong>for</strong> managing work<br />

schedules and workload efficiently; and documents<br />

include detailed in<strong>for</strong>mation on decommissioning, such as<br />

procedure and scenarios, are important.<br />

p In order to ensure that the cutting work is per<strong>for</strong>med<br />

smoothly, the supervisors and workers need to <strong>for</strong>m<br />

appropriate teams. Organizational factors, such as the<br />

objectives of these team, necessary team-building<br />

elements (cooperation, role, etc.), and precise and<br />

detailed communication of the supervisors’ judgement/decisions<br />

to the workers are important.<br />

p Next, the tasks should be well-designed including the<br />

preparation, methods and strategy, as well as processes<br />

involved in the tasks. Because the work is actively<br />

con ducted throughout the day, the work assignment<br />

Level 1 Level 2 Level 3<br />

Human<br />

Operation<br />

Ergonomic<br />

System<br />

Psychological<br />

State<br />

Physical<br />

State<br />

Per<strong>for</strong>mance<br />

Capability<br />

Organizational<br />

Factors<br />

Task Management<br />

Procedure and<br />

In<strong>for</strong>mation<br />

HMI<br />

(Human Machine<br />

Interface)<br />

Workplace Design<br />

Workplace Physical<br />

Environment<br />

Stress<br />

and shift work schedule should be coordinated and<br />

managed.<br />

p Finally, because supervisors and workers understand<br />

and per<strong>for</strong>m their work based on the procedures and<br />

scenarios related to the decommissioning of the RPVI<br />

components, in<strong>for</strong>mation on regulations, equipment<br />

and facilities, as well as accuracy and details of the<br />

procedures are important.<br />

Furthermore, because the cutting work is human- centric,<br />

the surrounding situations and conditions must be<br />

designed in a manner that is suitable <strong>for</strong> human beings.<br />

There<strong>for</strong>e, “ergonomic system” factors, such as worker<br />

interaction with the necessary equipment and facilities,<br />

which are based on the decommissioning task characteristics<br />

and requirements <strong>for</strong> RPVI components, workplace<br />

design based on the path of the supervisor and worker and<br />

task types, and optimal workplace environment <strong>for</strong><br />

workers, are important points to consider.<br />

p The systems and equipment the supervisors and<br />

workers operate must be interactive to ensure that they<br />

are not difficult to use to ensure the convenience and<br />

safety of the workers.<br />

p Because the RPV is located under water and is not<br />

directly visible and accessible to humans, the cutting<br />

processes under the RPVI decommissioning task<br />

are generally per<strong>for</strong>med using remotely controlled<br />

devices. There<strong>for</strong>e, the configuration, condition,<br />

and per <strong>for</strong>mance of these remote systems, such as<br />

Emotional State (Excitement, Boredom, Accomplishment, Frustration, Dissatisfaction)<br />

Safety Awareness<br />

Fatigue<br />

Physical Capability<br />

Discom<strong>for</strong>t<br />

Task Knowledge and Memory (Diagnoses, Action, Results, In<strong>for</strong>mation)<br />

Experience<br />

Personal Capability<br />

Overall Planning<br />

Supervision of Management including Decision Making<br />

Team Factors(Roles, Coordination, Risk Management, Atmosphere, Assisting Other)<br />

Work Process Design (considering Task Characteristic)<br />

Workload Management<br />

Problem Identification and Solution<br />

Communication Availability and Quality<br />

Procedure and In<strong>for</strong>mation Availability<br />

Procedure and In<strong>for</strong>mation Complexity<br />

Procedure and In<strong>for</strong>mation Accuracy and Completeness<br />

Procedure and In<strong>for</strong>mation Feedback and Recency<br />

Interaction Element (Menu, Push Button, Direct Manipulation, Special Symbol, Shape Type)<br />

Familiarity of Equipment and Facility<br />

Complexity of Equipment and Facility<br />

Maintenance<br />

Physical Access to Work Items<br />

Warning Sign (Alarm Location, Quantity, Intensity, Importance and Easy of Identification)<br />

Arrangement of Functional Areas<br />

Safety Device<br />

Noise<br />

Lighting<br />

Temperature<br />

Radiation Level<br />

FEATURE | ENVIRONMENT AND SAFETY 519<br />

| Tab. 4.<br />

PSF Classification System in the Reactor Decommissioning Activity.<br />

Feature<br />

Development of Per<strong>for</strong>mance Shaping Factors <strong>for</strong> Human Error Reduction during Reactor Decommissioning Activities through the Task Analysis Method ı Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

FEATURE | ENVIRONMENT AND SAFETY 520<br />

remote control, camera, and display should be well<br />

established, and designers should consider functional<br />

spatial arrangements and accessibility in order to use<br />

them.<br />

p An optimal working environment is required to ensure<br />

the safety of workers during the decommissioning<br />

work. In particular, it is important to establish an<br />

optimal physical environment <strong>for</strong> the workers by<br />

checking <strong>for</strong> factors, such as the noise of the equipment,<br />

temperature experienced by the workers wearing<br />

protective clothing, level of radiation exposure, cleanliness<br />

of the workplace, and turbidity of the water in<br />

which the cutting operation is being per<strong>for</strong>med.<br />

In particular, the PSFs are classified into three categories<br />

per the level of detail and importance required <strong>for</strong> the<br />

decommissioning activities. Level 1 includes “Human<br />

Factors,” “Operation Factors,” and “Ergonomic System<br />

Factors.” Next, Level 2 of Human Factors includes the<br />

basic elements responsible <strong>for</strong> human error, namely<br />

the psychological and physical state and per<strong>for</strong>mance<br />

capability of the supervisors and workers. Level 2 of<br />

Operation Factors is composed of Organizational Factors,<br />

Task Management, Procedures, and In<strong>for</strong>mation. Finally,<br />

Level 2 <strong>for</strong> human-oriented Ergonomic Factors is composed<br />

of HMI, Workplace Design, and Workplace Physical<br />

Environment. The PSFs under Level 3 are selected in a<br />

manner so as to not overlap with each other; in particular,<br />

these include a total of 44 factors as shown in Table 4.<br />

4 Concluding Remarks<br />

In this work, we study and present the methodology <strong>for</strong><br />

HRA <strong>for</strong> the cutting task of the <strong>Nuclear</strong> RPVI components<br />

in order to reduce human errors during the nuclear decommissioning<br />

activities. In order to do so, the HRA implementation<br />

procedures were reviewed to identify the components<br />

of the HRA and the most important TA was per<strong>for</strong>med<br />

in preparation <strong>for</strong> the HRA. The task characteristics,<br />

procedures, and in<strong>for</strong>mation on the decommissioning<br />

of the RPVI components were identified by reviewing relevant<br />

literature in the field and examining experiences in<br />

decommissioning at overseas nuclear power plants.<br />

In addition, PSFs required <strong>for</strong> the nuclear decommissioning<br />

activities were identified using TA results and<br />

PSFs selection criterion <strong>for</strong> RPVI cutting (10 tasks in total).<br />

The results of the PSF review applied to nuclear facilities<br />

and types of human errors identified through the detailed<br />

TA were synthesized. Furthermore, the selection criteria<br />

<strong>for</strong> the PSFs affecting the decommissioning activities were<br />

set. Finally, PSFs were selected and classified using a<br />

classification system.<br />

In a future study, we will examine the interrelationship<br />

among the PSFs and consider methods <strong>for</strong> assessing<br />

the PSFs. Moreover, we will develop a framework to model<br />

the mutual influences that exist among the PSFs with<br />

appropriate consideration of the relationships and<br />

dependencies among them. The collection of experience<br />

data <strong>for</strong> decommissioning nuclear power plants will also<br />

reduce the uncertainty in the in<strong>for</strong>mation used to per<strong>for</strong>m<br />

HRA.<br />

References<br />

[1] Madonna, M., et al.: Il Fattore Umano Nella Valutazione Dei Rischi: Confronto Metodologico Fra<br />

Le Tecniche Per L’analisi Dell’affidabilità Umana. Prevenzione Oggi. 5 (n. 1/2), 67–83 (2009).<br />

[2] Hollnagel, E.: Cognitive reliability and error analysis method (CREAM). Elsevier, (1998).<br />

[3] Griffith, C.D., Mahadevan, S.: Inclusion of fatigue effects in human reliability analysis. Reliability<br />

Engineering & System Safety, 96 (<strong>11</strong>), 1437–1447 (20<strong>11</strong>).<br />

[4] Lee, Y., et al.: Research Activities and Techniques <strong>for</strong> the Prevention of Human Errors during the<br />

Operation of <strong>Nuclear</strong> <strong>Power</strong> Plants. <strong>Journal</strong> of the Ergonomics Society of Korea, 30 (1), 75-86<br />

(20<strong>11</strong>).<br />

[5] O’Hara, J. M., et al.: Human Factors Engineering Program Review Model, US NRC, NUREG-07<strong>11</strong>,<br />

Rev. 3, 2012<br />

[6] Center <strong>for</strong> Chemical Process Safety : Guidelines <strong>for</strong> preventing human error in process safety<br />

(1994)<br />

[7] Byung-Sik Lee: Optimization of reactor pressure vessel internals segmentation in KoreaATW<br />

Vol.62 (2017), Issue <strong>11</strong>, 654~658, November, 2017<br />

[8] Boucau, J., et.al.: Best practices <strong>for</strong> preparing vessel internals segmentation projects. No.<br />

NEA-PREDEC--2016. February 16–18, Lyon, France (2016).<br />

[9] Boring, R.L.: Modelling human reliability analysis using MIDAS. In: <strong>International</strong> Workshop on<br />

Future Control Station Designs and Human Per<strong>for</strong>mance Issues in <strong>Nuclear</strong> <strong>Power</strong> Plants (2006).<br />

[10] Human per<strong>for</strong>mance reference manual, Institute of <strong>Nuclear</strong> <strong>Power</strong> Operations, INPO-06-003,<br />

2006<br />

[<strong>11</strong>] Chang, Y. H. J., Mosleh, A.: Cognitive modeling and dynamic probabilistic simulation of operating<br />

crew response to complex system accidents. Part 2: IDAC per<strong>for</strong>mance influencing factors model.<br />

Reliability Engineering & System Safety, 92(8), 1014-1040 (2006).<br />

[12] Blackman, H. S., Gertman, D. I., Boring, R. L.: Human error quantification using per<strong>for</strong>mance<br />

shaping factors in the SPAR-H method. In Proceedings of the human factors and ergonomics<br />

society annual meeting (Vol. 52, No. 21, pp. 1733-1737). Sage CA: Los Angeles, CA: SAGE<br />

Publications (2008).<br />

[13] Li, P. C., Chen, G. H., Dai, L. C., Zhang, L.: A fuzzy Bayesian network approach to improve the<br />

quantification of organizational influences in HRA frameworks. Safety science, 50(7), 1569-1583<br />

(2012).<br />

[14] Jung, K., Byun, S., Kim, J., Heo, E., Park, H.: An empirical study on evaluation of per<strong>for</strong>mance<br />

shaping factors on AHP. <strong>Journal</strong> of the Ergonomics Society of Korea, 30(1), 99-108 (20<strong>11</strong>).<br />

[15] Mindock, J.: Development and Application of Spaceflight PSFs <strong>for</strong> HRA. PhD Thesis , University of<br />

Colorado at Boulder (2012).<br />

[16] Baek, D., Koo, L., Lee, K., Kim, D., Yoon, W., Jung, M.: Taxonomy of PSFs <strong>for</strong> human error analysis<br />

of railway accidents. Business Administration, Hanyang University (2007).<br />

Authors<br />

Byung-Sik Lee<br />

Hyun-Jae Yoo<br />

Chang-Su Nam<br />

Dankook University<br />

<strong>11</strong>9, Dandae-ro, Dongnam-gu, Cheonan-si<br />

Chungnam, 3<strong>11</strong>16<br />

Republic of Korea<br />

Acknowledgments<br />

This work was supported by the National Research<br />

Foundation of Korea (NRF), granted financial resource<br />

from the Ministry of Science, ICT and Future Planning<br />

(No. 2017M2A8A5015148 and No.2016M2B2B1945086),<br />

Republic of Korea.<br />

Feature<br />

Development of Per<strong>for</strong>mance Shaping Factors <strong>for</strong> Human Error Reduction during Reactor Decommissioning Activities through the Task Analysis Method ı Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Root Causes of the<br />

Three Mile Island Accident<br />

Zoltan R. Rosztoczy<br />

The accident at Unit 2 of the Three Mile Island nuclear power plant, at that time operated and partly owned by<br />

Metropolitan Edison Company, occurred 40 years ago, on March 28, 1979. Following the accident, two major<br />

investigations were conducted, one by the President’s Commission on the Accident at Three Mile Island [1], appointed<br />

by President Carter, and the other by the <strong>Nuclear</strong> Regulatory Commission’s Special Inquiry Group. [2] The investigations<br />

documented the timeline of the accident and the availability and failure of equipment, and addressed operator actions<br />

during the accident, the training of operators, and NRC procedures that applied to the event. The design process <strong>for</strong> the<br />

plant and the designer’s responsibilities, including the plant’s safety analysis, were not addressed. Many additional<br />

studies and papers have been published over the past 40 years, none of which have addressed the design process or the<br />

safety analysis of the plant. The only ef<strong>for</strong>t specifically addressing the design of the plant and responsibility <strong>for</strong> the<br />

accident was Metropolitan Edison’s lawsuit against Babcock & Wilcox (B&W), the designer of the plant. A trial began<br />

but was terminated, and the case was settled out of court. The court records are sealed; in<strong>for</strong>mation is not available.<br />

More than 10 years prior to the TMI-2<br />

accident, B&W was designing its first<br />

nuclear power plant. In the designation<br />

of safety systems and in the safety<br />

analysis of the plant, there were two<br />

relatively minor but important omissions.<br />

These omissions turned out to<br />

be the root causes of the accident. If<br />

just one of them had been corrected<br />

during the intervening years, the<br />

accident would have been avoided.<br />

The TMI design was reviewed by<br />

utilities purchasing plants from B&W<br />

and by the NRC. The omissions<br />

remained undetected. The safety role<br />

of the pilot-operated relief valve<br />

(PORV) and the PORV block valve<br />

were not fully appreciated. The manufacturer<br />

of the PORV was not notified<br />

of the valve’s safety function, namely<br />

that it has to be able to close after<br />

being exposed to accident loads. [2]<br />

Also, the plant’s safety analysis report<br />

(SAR) did not address loss-of-coolant<br />

accidents (LOCA) initiated by very<br />

small breaks. Un<strong>for</strong>tunately, the plant<br />

responds very differently to an event<br />

initiated by a stuck PORV than to the<br />

small-break events presented in the<br />

SAR. At the time, this was unknown.<br />

Lessons learned from the omissions<br />

in the TMI design are timely today,<br />

when new types of reactors, such<br />

as small modular reactors, are on the<br />

drawing board. The designers of these<br />

new systems can learn from the TMI<br />

experience.<br />

The initiating event<br />

Operators attempting to clean a<br />

condensate polisher tripped the steam<br />

generator feedwater pumps. Then,<br />

the plant safety system tripped the<br />

turbine. The turbine was no longer<br />

removing heat from the reactor coolant<br />

system (RCS), the temperature<br />

and pressure of the RCS started rising<br />

rapidly, and the PORV opened, as<br />

designed.<br />

Upon shutdown of the feedwater<br />

pumps, the plant’s safety system<br />

turned on the emergency feedwater<br />

pumps. Due to a maintenance error,<br />

both emergency feedwater block<br />

valves, which are supposed to be open<br />

when the plant is operating, were<br />

closed, so no emergency feed-water<br />

reached the steam generators. The<br />

closed valves caused the RCS to heat<br />

up faster than in the case of a normal<br />

turbine trip, and the PORV was<br />

exposed to a larger load than normal,<br />

most likely a heavy two-phase flow<br />

(steam and water mixture) or water<br />

discharge. Thus, the closed valves<br />

could have played a role in causing<br />

the accident. This possibility is not<br />

addressed in the literature.<br />

As RCS pressure increased, the<br />

reactor protection system shut the<br />

reactor down, after which the RCS<br />

pressure dropped. The PORV should<br />

have closed, but instead it stuck open,<br />

and the plant faced a LOCA. The<br />

obvious question is, “Why did the<br />

PORV fail to close?”<br />

Designers of nuclear power plants<br />

have a dual responsibility. They must<br />

design the plant not only <strong>for</strong> normal<br />

operation of generating electricity,<br />

but also <strong>for</strong> safe per<strong>for</strong>mance in case<br />

of events that might occur during the<br />

lifetime of the plant and in case of<br />

postulated accidents.<br />

Components of systems that have<br />

both an operating function and a<br />

safety function have to be identified<br />

and designed to per<strong>for</strong>m both functions<br />

in a reliable manner. The PORV<br />

is a good example of such a component.<br />

During normal operation, the<br />

PORV maintains RCS pressure below<br />

specified limits by opening and closing<br />

and by discharging steam from the<br />

pressurizer. During abnormal events,<br />

as in the TMI-2 case, the PORV could<br />

be discharging two-phase flow or<br />

water. The valve must be designed to<br />

per<strong>for</strong>m its safety function – namely,<br />

to close following a two-phase flow or<br />

water discharge. Apparently, this was<br />

not the case at TMI. The PORV was<br />

not designed to per<strong>for</strong>m its safety<br />

function. The purchase order failed to<br />

specify this requirement, and the<br />

supplier of the valve did not know that<br />

the valve had a safety function and<br />

that it had to close following twophase<br />

flow or water discharge. [2]<br />

Designers are also responsible <strong>for</strong><br />

incorporating operating experience<br />

into their design. Prior to the TMI-2<br />

accident, PORVs failed to close seven<br />

times at B&W plants. [2] Despite<br />

this record, the PORV itself was not<br />

modified or replaced. Instead, an<br />

indicator light was installed to show<br />

whether the block valve upstream of<br />

the PORV had received a signal to<br />

close, but there was no indication in<br />

the control room that the valve had<br />

actually closed.<br />

Reprinted with<br />

permission from the<br />

March 2019 issue of<br />

<strong>Nuclear</strong> News<br />

Copyright © 2019 by<br />

the American <strong>Nuclear</strong><br />

Society<br />

| From right to left: President Jimmy Carter, Pennsylvania Gov. Richard<br />

Thornburgh, and the NRC’s Harold Denton tour the TMI-2 control room on<br />

April 1, 1979. Photo: The Jimmy Carter Presidential Library and Museum<br />

521<br />

ENVIRONMENT AND SAFETY<br />

Environment and Safety<br />

Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

ENVIRONMENT AND SAFETY 522<br />

PORVs opened relatively frequently<br />

on B&W-designed pressurized<br />

water reactors. The thermal hydraulic<br />

design of the reactor core was closer<br />

to acceptable limits than other PWR<br />

cores, and the amount of water<br />

contained in the secondary side of<br />

the steam generators was very small –<br />

only 25 percent of some other PWRs’<br />

water content. [2] These differences<br />

made the system react faster to<br />

changes. With quick plant response,<br />

the PORV came into action relatively<br />

frequently. More frequent use of the<br />

PORV led to more frequent failures.<br />

The PORV failure at TMI-2 was the<br />

eighth at a B&W plant, more than<br />

an order of magnitude higher than<br />

PORV failures with other reactor<br />

designs. [2]<br />

The NRC has specific requirements<br />

<strong>for</strong> equipment related to safety. Equipment<br />

essential to accident mitigation<br />

and equipment whose failure can<br />

cause or aggravate an accident are<br />

considered “safety related.” A stuckopen<br />

PORV causes a breach in the<br />

boundary of the RCS, creating a<br />

LOCA. Among postulated accidents,<br />

LOCAs are considered to be the<br />

most serious, and there<strong>for</strong>e they<br />

receive special attention. <strong>Nuclear</strong><br />

power plants are designed with three<br />

barriers to protect the public from<br />

radioactive material release: The<br />

fuel is enclosed in a sealed cladding,<br />

the reactor core is within the closed<br />

RCS, and the RCS is covered by a<br />

containment building. Among all<br />

postulated accidents, there is only<br />

one – the LOCA – where two of the<br />

barriers are predicted to be damaged.<br />

In the case of a LOCA, the event itself<br />

breaches the RCS, and the predicted<br />

consequences of the accident are<br />

expected to damage some of the<br />

fuel cladding. Protection of the public<br />

is reduced to a single barrier, the<br />

containment building. Furthermore,<br />

valve failures are more likely than<br />

pipe breaks. Thus, the most likely<br />

LOCA is the stuck-open PORV.<br />

Surprisingly, the PORV was not<br />

identified by the designer as safetyrelated<br />

equipment. The design was<br />

reviewed by Metropolitan Edison<br />

and evaluated by the NRC. Neither<br />

objected to the PORV not being<br />

designated as safety related, and<br />

the NRC approved the construction<br />

permit application. Had the PORV<br />

been designated as safety-related<br />

equipment, it would have had to<br />

meet reliability requirements and be<br />

tested under accident conditions. If<br />

the TMI-2 PORV had been tested, it<br />

most likely would not have passed.<br />

Following the accident, the manufacturer<br />

of the valve stated that the<br />

TMI-2 PORV was not qualified to close<br />

following a two-phase flow or water<br />

discharge. [2] If the PORV had been<br />

designated as safety related, it would<br />

have been replaced or modified.<br />

The reason given <strong>for</strong> not designating<br />

the PORV as safety related was<br />

the presence of a block valve upstream<br />

of the PORV. If the PORV is stuck<br />

open, the block valve can be closed,<br />

terminating the accident. Thus, the<br />

block valve is essential <strong>for</strong> the mitigation<br />

of a PORV failure accident, and<br />

it is also considered safety-related<br />

equipment. It must have automatic<br />

safety-grade actuation initiated from<br />

the stuck-open PORV or, if the initiation<br />

is manual, safety-grade position<br />

indication must be available in the<br />

control room with sufficient time<br />

<strong>for</strong> operator action. Neither of these<br />

conditions existed at TMI-2.<br />

Consequences of PORV failure<br />

Part of the designer’s responsibility is<br />

to conduct a complete and detailed<br />

safety analysis of the plant. The<br />

analysis must include transients that<br />

might occur in the plant. The analysis<br />

of transients must show that continued<br />

operation of the plant following<br />

these events is justified. The<br />

plant’s safety analysis also has to<br />

address all potential accidents, both<br />

system failures and operator errors<br />

that the plant could be subject to,<br />

unless they are considered to be<br />

extremely unlikely (severe accidents).<br />

It is the designer’s responsibility to<br />

identify all accident types specific<br />

to the design of the plant. In the case<br />

of water-cooled reactors, one of these<br />

accident types is a breach in the RCS<br />

– a LOCA.<br />

For PWRs such as TMI-2, it is an<br />

NRC requirement that a complete<br />

spectrum of breaches in the RCS be<br />

analyzed, starting from the doubleended<br />

break of the largest pipe in the<br />

RCS down to the break size that the<br />

makeup water system can keep up<br />

with. Un<strong>for</strong>tunately, it was not emphasized<br />

that a breach in the system<br />

includes stuck-open valves if the<br />

valve’s discharge area is within the<br />

size range of the postulated accident.<br />

The PORV falls within the size range.<br />

Complete spectrum also means all<br />

possible break locations. The consequences<br />

of a stuck-open valve on<br />

the top of the pressurizer could be<br />

different from a same-size break at a<br />

lower elevation.<br />

Typically, prior to the TMI accident,<br />

the large-break LOCA analysis<br />

| A six-page special report – <strong>Nuclear</strong> News’s<br />

initial coverage of the TMI accident –<br />

was mailed separately to subscribers and<br />

ANS members in April 1979.<br />

included break sizes in both the hot<br />

and cold legs of the RCS, starting<br />

from a double-ended break down<br />

to a 0.5 square-foot break. Usually,<br />

the consequences were most severe<br />

at one of the larger breaks. From<br />

there on, smaller sizes resulted in<br />

more favorable consequences. The<br />

small-break LOCA analysis ran from<br />

0.5 square foot down to about<br />

0.1 square foot. The trend was the<br />

same; smaller breaks had less severe<br />

consequences. Breaks even smaller<br />

were not analyzed <strong>for</strong> two reasons:<br />

(1) the calculations ran long on the<br />

computer and the analyses were<br />

expensive, and (2) the trend was<br />

already established. Instead, the<br />

assumption was made that the<br />

trend would continue down to the<br />

smallest required size. Also, smallbreak<br />

LOCA analysis was assumed<br />

to be independent of break location.<br />

Thus, breaks less than 0.5 square<br />

foot were not analyzed at different<br />

locations, and breaks less than<br />

0.1 square foot were not analyzed<br />

at all.<br />

The safety analysis of the plant<br />

serves many purposes. It provides<br />

both the designer and the operator of<br />

the plant with an understanding of<br />

how the plant responds to a specific<br />

event or accident, indicates potential<br />

damage if mitigating actions are<br />

not taken, guides the designer in the<br />

design of the needed safety systems,<br />

and provides in<strong>for</strong>mation <strong>for</strong> training<br />

the operating staff. The analysis<br />

shows how reactor operators can<br />

recognize a specific event and what<br />

actions they must take and provides<br />

the needed in<strong>for</strong>mation <strong>for</strong> the<br />

preparation of emergency procedures.<br />

The results of the analysis also show<br />

Environment and Safety<br />

Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

compliance with applicable regulations.<br />

It shows that potential damage<br />

has been mitigated and the safety<br />

of plant personnel and the public is<br />

ensured.<br />

The TMI-2 LOCA analysis was<br />

per<strong>for</strong>med by the designer. It was<br />

rather elaborate but was incomplete<br />

in the sense that it failed to show that<br />

very small breaks behave differently<br />

and have more serious consequences<br />

than small breaks. As part of the<br />

TMI-2 licensing process, no LOCA<br />

analysis was per<strong>for</strong>med by the<br />

designer, and no LOCA analysis was<br />

submitted to the NRC by the utility<br />

<strong>for</strong> a break size anywhere close to<br />

the size of the discharge opening of<br />

the PORV (about 2 square inches). No<br />

analyses were per<strong>for</strong>med <strong>for</strong> any size<br />

break at the top of the pressurizer or<br />

<strong>for</strong> a LOCA caused by a stuck-open<br />

PORV. Un<strong>for</strong>tunately, as was learned<br />

from the TMI accident and from<br />

analysis per<strong>for</strong>med after the accident,<br />

a stuck-open PORV.<br />

LOCA is very different from the<br />

small-break LOCA analyses presented<br />

<strong>for</strong> TMI-2. Analysis of smaller breaks<br />

showed that the trend reverses<br />

and the consequences increase with<br />

decreasing break size. The plant<br />

responds differently, reliance on<br />

safety systems and instrumentation<br />

changes, and different operator actions<br />

are required. The consequences<br />

of a PORV failure are even more<br />

different. RCS pressure can drop while<br />

the water level in the pressurizer is<br />

rising. Void can <strong>for</strong>m in the reactor<br />

core and accumulate at high locations<br />

of the RCS while the pressurizer water<br />

level is high. Furthermore, the water<br />

level can drop below the top of the<br />

core, resulting in core damage, while<br />

the pressurizer water level is still high.<br />

Obviously, pressurizer water level<br />

indication is not a useful tool <strong>for</strong> the<br />

handling of this accident.<br />

A special design feature of B&W<br />

plants further aggravates this effect.<br />

The pressurizer surge line is designed<br />

with a loop seal to prevent steam from<br />

entering the pressurizer. Eliminating<br />

steam flow to the pressurizer prevents<br />

water level drops in the pressurizer,<br />

keeping the water level high, while<br />

void is accumulating in the RCS.<br />

Due to the lack of analysis, the<br />

consequences of a PORV failure<br />

were unknown. As it turned out, the<br />

actual consequences without proper<br />

mitigation were a lot worse than<br />

one would expect. The assumption<br />

that consequences get better with<br />

decreasing break size was incorrect.<br />

The actual consequences of the<br />

accident equally surprised the designers,<br />

the owner/operator of the<br />

plant, and the regulators. The plant’s<br />

response to the PORV failure was<br />

totally unexpected.<br />

Accident management<br />

Early in the morning of March 28,<br />

1979, four young operators at<br />

TMI-2 realized that something had<br />

happened, but they had no idea what<br />

it was. The turbine shut down, the<br />

reactor scrammed, and a cascade of<br />

alarms sounded and flashed. The<br />

plant was acting strangely. RCS<br />

pressure was decreasing while the<br />

pressurizer water level was increasing.<br />

The operators had not faced this<br />

situation be<strong>for</strong>e. It was not covered in<br />

their training. They did not know<br />

what to do.<br />

The event facing the operating<br />

crew was a stuck-open PORV and a<br />

very small LOCA. They did not know<br />

that was the case. There was no direct<br />

indication of PORV position in the<br />

control room. They could not see that<br />

the PORV was stuck open.<br />

Not knowing what was going on<br />

and not having familiarity with the<br />

event, the operators were improvising,<br />

trying to maintain water level<br />

in the RCS within prescribed limits.<br />

They relied on the pressurizer water<br />

level reading, as they were trained to<br />

do. Un<strong>for</strong>tunately, they took a few<br />

inappropriate actions, which included<br />

turning off the high-pressure emergency<br />

core cooling system, opening<br />

the letdown line, ignoring signs of<br />

overheating of the reactor core, and<br />

pumping radioactive water to the<br />

auxiliary building. All of this occurred<br />

be<strong>for</strong>e they learned – two hours and<br />

20 minutes into the accident – that the<br />

PORV was stuck open. Then they<br />

took corrective action and closed<br />

the block valve.<br />

The obvious question is, “Why<br />

were the operators in the dark, and<br />

why did they lack familiarity with<br />

this event?” Their training covered<br />

mitigation of postulated accidents, including<br />

LOCAs. There was only one<br />

set of accidents missing, very small<br />

LOCAs, including PORV failure. Since<br />

the designer did not analyze this<br />

event, it was not included in operator<br />

training. Not knowing the plant’s<br />

response to a PORV failure, the<br />

designers and the training staff<br />

instructed the operators to always<br />

rely on the pressurizer water level<br />

indication <strong>for</strong> water level measurements<br />

in the RCS. The operators<br />

followed their training on that<br />

morning.<br />

Despite the total lack of training<br />

<strong>for</strong> a stuck-open PORV event, could<br />

the operators have realized what<br />

was going on and taken appropriate<br />

action? The answer is yes. [1] The<br />

temperature of the PORV drain pipe<br />

was monitored and showed high<br />

readings, an alarm signaled high<br />

water level in the containment building<br />

sump, high neutron level indications<br />

were observed in the reactor<br />

core, temperature and pressure were<br />

rising in the containment building,<br />

and the reactor coolant pumps were<br />

vibrating. Any of these observations,<br />

typical of a LOCA, could have brought<br />

attention to a stuck-open PORV. The<br />

remedy should have been obvious:<br />

Close the block valve.<br />

Once the block valve was closed,<br />

the LOCA was terminated. The next<br />

step was to cool the core by natural<br />

circulation of the water in the RCS.<br />

This was not possible, however, due to<br />

the large amount of void that had<br />

accumulated in the RCS. The operating<br />

staff had to improvise again to<br />

reduce the void and the bubble in<br />

the RCS, and then to establish neutral<br />

circulation. It took a couple of days’<br />

work <strong>for</strong> them to accomplish this.<br />

Both the plant’s designer and<br />

operator lacked the knowledge of how<br />

the plant would respond to a stuckopen<br />

PORV. What they did not know,<br />

they could not pass on to the operators.<br />

The operators’ training was<br />

misleading, and the emergency procedure<br />

was incorrect <strong>for</strong> the incident<br />

they were facing.<br />

Industry practice and<br />

oversight<br />

B&W’s two omissions – safety-related<br />

classification of the PORV and the<br />

PORV block valve, and the lack of<br />

PORV failure analysis – were not<br />

unique to B&W. The other U.S. PWR<br />

designers, Westinghouse and Combustion<br />

Engineering, made the same<br />

omissions. How could three independent<br />

sets of engineers make the<br />

same mistake? Licensing of the plants<br />

was a major consideration. The SAR<br />

was the centerpiece of the licensing<br />

review. Precedent provided guidance<br />

<strong>for</strong> the preparation of the report.<br />

Analyses presented in previous<br />

applications were included in the<br />

report; analysis that was not<br />

required was ignored. Dozens of<br />

utilities received SARs with the<br />

same omissions. The omissions<br />

had a direct and major effect on<br />

the training of reactor operators.<br />

The operators received training on<br />

a plant simulator, with postulated<br />

ENVIRONMENT AND SAFETY 523<br />

Environment and Safety<br />

Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

ENVIRONMENT AND SAFETY 524<br />

| TMI II – Late accident phase.<br />

Zoltan R. Rosztoczy<br />

was Manager of the<br />

Safety Analysis<br />

Department of<br />

Combustion Engineering’s<br />

<strong>Nuclear</strong> Division<br />

in its <strong>for</strong>mative years.<br />

He later joined the<br />

<strong>Nuclear</strong> Regulatory<br />

Commission and was<br />

a charter member of<br />

the commission’s<br />

Senior Executive<br />

Service.<br />

accidents programmed into the<br />

simulator. One accident, the PORV<br />

failure, was missing. Nobody noticed<br />

it or took corrective action. After<br />

PORVs failed seven times at B&W<br />

plants, this accident was still missing<br />

from the operator training program<br />

and from the simulator.<br />

In the case of PWR evaluations,<br />

the NRC had the distinct advantage<br />

of reviewing SARs from three<br />

independent designers. Comparisons<br />

among the three designs frequently<br />

helped in the reviews. The NRC,<br />

however, failed to recognize its own<br />

effect on plant design and analysis.<br />

A nuclear power plant is a complex<br />

system. A regulatory review and<br />

evaluation cannot address all aspects<br />

of the design, and priorities have to<br />

be set. There was a tendency not to<br />

require more from an applicant than<br />

was required from previous ones.<br />

Spending time reviewing areas of the<br />

design that weren’t reviewed in the<br />

past was discouraged. Consequently,<br />

regulators and designers addressed<br />

the same areas of the design and the<br />

safety analyses over and over again<br />

and ignored other areas.<br />

Conclusions<br />

Failure to incorporate the safety<br />

function of the PORV and the block<br />

valve in the design of the plant created<br />

the condition <strong>for</strong> the TMI accident.<br />

With no positive indication in the<br />

control room of an open PORV and no<br />

positive position indication of the<br />

block valve, the operators were left to<br />

guess what was going on and what<br />

needed to be done.<br />

Not having addressed PORV failure<br />

in the plant safety analysis, the<br />

designers, as well as the training and<br />

operating staff, were unfamiliar with<br />

the plant’s response to this type of<br />

accident. They did not know that<br />

| TMI II – Late accident phase.<br />

the plant conditions the operators<br />

were facing were possible, and as<br />

a result, training and instructions<br />

were inadequate. When similar plant<br />

designs are being reviewed or evaluated<br />

one after the other, there is a<br />

tendency to address the same issues<br />

in each case. Plants are very complex,<br />

and not everything can be evaluated<br />

as part of one review. It is appropriate<br />

to shift emphasis in subsequent<br />

reviews and to address issues previously<br />

not covered.<br />

Appropriate NRC regulations relative<br />

to LOCAs to control the design<br />

and operation of the plants’ safety<br />

systems and develop operator training<br />

programs and emergency procedures<br />

were evolving when B&W designed its<br />

first plants, but they were in place at<br />

the time of TMI-2’s licensing. The<br />

problem was that some of the<br />

regulations were not followed.<br />

The two omissions – not recognizing<br />

the safety function of the PORV<br />

and the block valve, and the failure to<br />

analyze the stuck-open PORV event –<br />

were the root causes of the TMI-2<br />

accident. Correcting the first omission<br />

would have prevented the accident.<br />

Correcting the second omission would<br />

have resulted in prompt and effective<br />

mitigation of the accident.<br />

Lessons learned<br />

Understanding the root causes of<br />

the TMI accident provides valuable<br />

guidance <strong>for</strong> nuclear power plant<br />

designers, especially <strong>for</strong> designers<br />

of new plant types, such as small<br />

modular reactors. The recognition<br />

of safety-related components and<br />

design-specific accidents is more<br />

complex and more difficult than it<br />

appears to be. It is the designer’s<br />

responsibility to identify all safetyrelated<br />

systems and components<br />

and to analyze all accident types.<br />

Many systems and components<br />

of a plant have both an operational<br />

function and a safety function. In the<br />

design of every system, the question<br />

must be raised as to whether a system<br />

or component has a safety function.<br />

Then, if applicable, it must be<br />

designed <strong>for</strong> both the operational<br />

function and the safety function.<br />

Plant response during accidents<br />

can be abnormal and never seen<br />

during normal operation. The plant’s<br />

safety analysis must be complete, and<br />

it must describe all potential plant<br />

responses.<br />

Designers cannot depend on<br />

utilities’ reviews and regulatory<br />

evaluations to correct shortcomings.<br />

The design must be done right in the<br />

first place, and the quality assurance<br />

process should guarantee perfection<br />

of the design.<br />

Acknowledgment<br />

I am grateful to Sheldon Trubatch <strong>for</strong><br />

his valuable suggestions, review of<br />

this article, and thoughtful comments<br />

and insight into the era from the legal<br />

perspective surrounding the accident.<br />

I have derived great benefit from our<br />

stimulating discussions.<br />

References<br />

1. John G. Kemeny, et al.: Report of the President’s Commission on<br />

the Accident at Three Mile Island (October 30, 1979).<br />

2. Mitchell Rogovin, George T. Frampton Jr.: Three Mile Island: A<br />

Report to the Commissioners and to the Public (January 1980).<br />

Author<br />

Zoltan R. Rosztoczy<br />

Environment and Safety<br />

Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Das neue Strahlenschutzrecht (IV) – Schutz vor Radon<br />

Christian Raetzke<br />

Den vorläufigen Abschluss der kleinen Reihe zum neuen Strahlenschutzrecht sollen heute die neuen Regelungen zum<br />

Schutz vor Radon bilden. Das ist zwar ein Thema, das in der Kernenergie und Kerntechnik nicht unbedingt im<br />

Vordergrund steht; dafür dürfte es künftig in Teilen der Wirtschaft und der Bevölkerung um so mehr für Aufmerksamkeit<br />

sorgen und ist sicherlich von allgemeinem Interesse.<br />

Der Schutz vor Radon wird im neuen Strahlenschutzrecht<br />

(im Strahlenschutzgesetz – StrlSchG – vom 27.06.2017<br />

und in der neuen Strahlenschutzverordnung – StrlSchV –<br />

vom 29.<strong>11</strong>.2018), das an vielen Stellen und besonders<br />

auch hier die Euratom-Grundnorm zum Strahlenschutz<br />

(Richtlinie 2013/59/Euratom) umsetzt, einschneidend<br />

neu und viel umfassender als bisher geregelt. In der alten<br />

StrlSchV gab es bei „Arbeiten“ (so die damalige Terminologie),<br />

die im Zusammenhang mit natürlich vorkommender<br />

Radioaktivität standen, bestimmte Regelungen für<br />

Arbeitsplätze bis hin zu einer Anzeigepflicht (§ 95 StrlSchV<br />

a.F.); dazu gehörte in Anlage XI Teil A eine kurze Liste<br />

„Arbeitsfelder mit erhöhten Radon-222-Expositionen“.<br />

Die Regelungen im neuen Strahlenschutzrecht sind viel<br />

umfassender aufgestellt. Sie betreffen nicht nur Arbeitsplätze,<br />

also die berufliche Strahlenexposition, sondern<br />

auch Aufenthaltsräume, also die Exposition der Bevölkerung.<br />

Einige Regelungen gelten speziell für eine der beiden<br />

Expositionskategorien, andere sind allgemeiner, übergreifender<br />

Art. Am besten findet man sich wohl zurecht,<br />

wenn man die Regelungen entsprechend in drei Gruppen<br />

betrachtet: allgemeine, auf Arbeitsplätze bezogene und<br />

auf Aufenthaltsräume bezogene Bestimmungen. So sind<br />

sie auch im Gesetz und in der Verordnung eingeteilt.<br />

Bei den allgemeinen Regelungen ist besonders die<br />

Pflicht der zuständigen Länderbehörden zu erwähnen, bis<br />

Ende 2020 sog. Radonvorsorgegebiete auszuweisen. Dieser<br />

Begriff (der so im Gesetz nicht auftaucht) be zeichnet<br />

Gebiete gem. § 121 StrlSchG, in denen die Radon-222-<br />

Aktivitätskonzentration in der Raumluft „in einer beträchtlichen<br />

Zahl von Gebäuden“ die Referenzwerte für Aufenthaltsräume<br />

oder Arbeitsplätze überschreitet. Die beiden<br />

Referenzwerte für Aufenthaltsräume und Arbeitsplätze sind<br />

im Gesetz <strong>for</strong>mal getrennt festgelegt (§§ 124 bzw. 126<br />

StrlSchG), lauten aber übereinstimmend 300 Becquerel<br />

(Bq) je Kubikmeter (m 3 ). Die Bestimmung der Vorsorgegebiete<br />

erfolgt „auf Grundlage einer wissenschaftlich<br />

basierten Methode“, die „Vorhersagen“ ermöglicht (so § 153<br />

StrlSchV); geeig nete Daten hierfür können auf tatsächlichen<br />

Messwerten, aber auch auf geologischen Daten, also<br />

letztlich auf Berechnungen beruhen. Eine „beträcht liche<br />

Zahl von Gebäuden“ soll erreicht sein, wenn auf mindestens<br />

75 % des auszuweisenden Gebiets der Referenzwert in<br />

min destens zehn Prozent der Gebäude überschritten wird.<br />

Sind erst einmal diese Radonvorsorgegebiete ausgewiesen,<br />

so ergibt sich aus § 123 StrlSchG die Pflicht, bei<br />

Neubauten in diesen Gebieten „geeignete Maßnahmen zu<br />

treffen, um den Zutritt von Radon aus dem Baugrund zu<br />

verhindern oder erheblich zu erschweren“. Das gilt einheitlich<br />

für Gebäude mit Arbeitsplätzen und/oder mit Aufenthaltsräumen,<br />

also auch für Wohngebäude oder Bauten<br />

für Schulen und Kindergärten. § 154 StrlSchV zählt fünf<br />

Arten von Maßnahmen auf und lässt „mindestens eine“<br />

davon genügen; damit soll offenkundig die An<strong>for</strong>derung<br />

handhabbar gemacht und eine Übersteigerung vermieden<br />

werden. Zu den Maßnahmen zählen etwa die Verringerung<br />

der Radon-222-Aktivitätskonzentration unter dem<br />

Gebäude, die Begrenzung der Rissbildung in Wänden und<br />

Böden mit Erdkontakt und die Absaugung von Radon an<br />

Randfugen oder unter Abdichtungen.<br />

Der Bund ist nach § 122 StrlSchG verpflichtet, einen<br />

Radonmaßnahmenplan zu erstellen; das ist mittlerweile<br />

geschehen (siehe www.bmu.de/publikation/radonmassnahmenplan).<br />

Das Dokument nennt, neben den gesetzlich<br />

ge<strong>for</strong>derten Maßnahmen, auch und vor allem In<strong>for</strong>mationskampagnen,<br />

Förderung der Allgemeinbildung<br />

zu Radon und der beruflichen Qualifikation und Weiterbildung,<br />

Entwicklung einheitlicher Messstrategien, Vergabe<br />

weiterer Studien und Ähnliches.<br />

Bei Arbeitsplätzen besteht dann eine Handlungspflicht,<br />

wenn sie entweder einem der bereits in der alten StrlSchV<br />

genannten „Arbeitsfelder“ angehören (vor allem Berg werke<br />

und Anlagen der Wassergewinnung und -ver teilung), und<br />

zwar unabhängig vom geografischen Gebiet, oder wenn sie<br />

sich im Erd- oder Kellergeschoss eines Gebäudes in einem<br />

Radonvorsorgegebiet befinden. §§ 127 bis 131 StrlSchG<br />

ordnen dann ein abgestuftes Verfahren an. Zeigt eine erste<br />

Messung, die innerhalb von 18 Monaten nach Einrichten<br />

des Arbeitsplatzes oder nach Ausweisung des Gebietes vorzunehmen<br />

ist, eine Überschreitung des Referenzwerts,<br />

muss der für den Arbeitsplatz Verantwortliche Maßnahmen<br />

zur Reduzierung der Radonkonzentration ergreifen und<br />

erneut messen. Liegt der Wert dann immer noch über dem<br />

Referenzwert, muss er den Arbeitsplatz bei der zuständigen<br />

Behörde anmelden und die Exposition abschätzen; vom<br />

Ergebnis dieser Abschätzung hängen dann bestimmte Maßnahmen<br />

des beruflichen Strahlenschutzes ab. Der Arbeitgeber<br />

wird damit nicht zum Strahlenschutzverantwortlichen,<br />

da es sich – in der neuen Terminologie des StrlSchG<br />

– nicht um eine geplante Expositionssituation, sondern um<br />

eine bestehende Expositionssituation handelt; er hat aber<br />

teils vergleich bare Pflichten.<br />

Für Aufenthaltsräume (also Wohnungen, Schulen,<br />

Kinderg ärten etc.) enthalten §§ 124 und 125 StrlSchG<br />

Regelungen, die keinen verpflichtenden Charakter haben,<br />

sondern auf Unterrichtung der Bevölkerung und auf das<br />

„Anregen“ von Messungen und ggf. von Schutzmaßnahmen<br />

hinauslaufen. Wie oben erwähnt, gibt es aber für<br />

Neubauten in Radonvorsorgegebieten, die Aufenthaltsräume<br />

enthalten, eine Pflicht, Maßnahmen zum Schutz<br />

vor Radon zu treffen; insofern sind nur Bestandsbauten<br />

vorerst von verbindlichen Maßnahmen verschont.<br />

Wie geht es weiter? Das Thema Radon steht schon seit<br />

einigen Jahren im Fokus der Strahlenschützer, im allgemeinen<br />

Bewusstsein dürfte es aber erst dann wirklich ankommen,<br />

wenn ab 2021 die Radonvorsorgegebiete ausgewiesen<br />

sind, In<strong>for</strong>mationskampagnen richtig anlaufen und echte<br />

Pflichten für viele Arbeitgeber und Bauherren entstehen. Es<br />

steht zu hoffen, dass die neuen Regelungen dann die<br />

gewünschte Steuerungswirkung entfalten und die findige<br />

Bevölkerung nicht einfach nur gemäß einem Spruch handelt,<br />

den der Verfasser von einem geschätzten Kollegen aus<br />

dem (radonmäßig stark betroffenen) Vogtland gehört hat:<br />

„Die Zahnarztpraxis unters Dach, die Oma in den Keller“.<br />

Author<br />

Rechtsanwalt Dr. Christian Raetzke,<br />

Beethovenstr. 19, 04107 Leipzig<br />

525<br />

SPOTLIGHT ON NUCLEAR LAW<br />

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

 The New Radiation Protection Law – Protection Against Radon ı Christian Raetzke


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

526<br />

RESEARCH AND INNOVATION<br />

Evaluation of a Double-Ended Guillotine<br />

LBLOCA Transient in a Generic<br />

Three-Loops PWR-900 with TRACE Code<br />

Coupled with DAKOTA Uncertainty<br />

Analysis<br />

Andrea Bersano and Fulvio Mascari<br />

In the present study, the model of a generic three-loops PWR-900 western type reactor has been developed and a<br />

double-ended guillotine break on the cold leg has been simulated by TRACE code. Through the SNAP graphical<br />

interface, a DAKOTA uncertainty analysis, based on the probabilistic method to propagate input uncertainty, has been<br />

per<strong>for</strong>med by selecting uncertain parameters related to the safety injection system and to the initial plant status. In<br />

particular, six uncertain input parameters have been considered: the accumulators’ initial pressure and temperature,<br />

the safety injection system temperature and flow rate, the reactor initial power and the containment initial pressure.<br />

The main figure of merit selected <strong>for</strong> the application of regression correlation is the hot rod cladding temperature. Both<br />

Pearson and Spearman’s correlation coefficients have been computed <strong>for</strong> the cladding temperature of the hot rod to<br />

characterize its correlation with the input uncertain parameters in the different phases of the transient. In addition, the<br />

dispersion of the calculated data have been discussed <strong>for</strong> selected relevant thermal-hydraulic parameters, such as the<br />

primary pressure, the core mass flow rate and the water collapsed level in the vessel.<br />

1 Introduction<br />

<strong>Nuclear</strong> energy is part of the energy<br />

mix of several countries considering<br />

its important role in the reduction of<br />

air pollution and CO 2 emissions<br />

caused by energy pro duction [1]. Due<br />

to the complexity of nuclear plants,<br />

caused by their complex geometry<br />

where multicomponent and twophase<br />

thermal hydraulic phenomena<br />

take place, their thermal hydraulic behavior<br />

is characterized with computational<br />

tools to assess their operating<br />

conditions and to evaluate their safety<br />

(both <strong>for</strong> Design Basis Accidents and<br />

Beyond Design Basis Accident) [2, 3].<br />

The computational tools used in<br />

the nuclear sector, also called codes,<br />

undergo a process of Verification and<br />

Validation (V&V) [4]. It is part of this<br />

rigorous process the use of an experimental<br />

“assessment database” [3];<br />

results from Separate Effect Test<br />

Facilities (SETF) and Integral Test<br />

Facilities (ITF) are used to evaluate<br />

the qualitative and quantitative code<br />

accuracy in the prediction of the<br />

phenomena of interest. The codes<br />

that use more realistic in<strong>for</strong>mation<br />

concerning phenomena and plant<br />

behavior are often referred to as Best<br />

Estimate (BE) codes [5].<br />

Though the high level of maturity<br />

reached by BE thermal hydraulic<br />

system codes in the last decades, in<br />

their application there are still some<br />

sources of uncertainty ( uncertainty is<br />

used as a measure of the error made<br />

with the code in predicting the plant<br />

behavior) affecting the calculation<br />

results [3]. In general the sources of<br />

uncertainty can be grouped as a) code<br />

uncertainty (e.g. approximations in<br />

the onservation equation and in the<br />

closure models and correlations)<br />

b) representation uncertainty (nodalization<br />

effect), c) scaling issue (codes<br />

validated against scaled-down facilities),<br />

d) plant uncertainty (e.g. initial<br />

and boundary conditions), e) user<br />

effect. For this reason, providing the<br />

result of a BE calculation alone may be<br />

not sufficient and the evaluation of<br />

the uncertainty on the results is<br />

required. Several methodologies have<br />

been developed in the past to per<strong>for</strong>m<br />

Uncertainty Analysis. In particular<br />

these uncertainty methodologies can<br />

be grouped in a) methods to propagate<br />

input uncertainty, divided in<br />

probabilistic (e.g. CSAU, GRS, IPSN,<br />

etc.) and deterministic methods (e.g.<br />

AEAW, EDF-Framatome, etc.); and b)<br />

method to extrapolate output uncertainty<br />

(e.g. UMAE) [6].<br />

In this framework, the target of the<br />

present paper is to use the probabi listic<br />

method to propagate the input<br />

uncertainty in the calculation of a cold<br />

leg double- ended guillotine break<br />

transient of a generic three-loops PWR-<br />

900 western type reactor with the<br />

avail ability of active and passive Emergency<br />

Core Cooling Systems (ECCS).<br />

The Uncertainty Quanti fication (UQ)<br />

application of this methodology is<br />

based on a set of statistical techniques<br />

<strong>for</strong> the evaluation of the number of<br />

needed code runs and <strong>for</strong> the correlation<br />

of the output results with the<br />

uncertain inputs. The method has been<br />

applied using DAKOTA toolkit [7],<br />

developed by Sandia National Laboratories,<br />

and has been used in the SNAP<br />

(Symbolic <strong>Nuclear</strong> Analysis Package)<br />

environment/architecture [8] with the<br />

related DAKOTA uncertainty plug-in<br />

[9]. The calculations have been per<strong>for</strong>med<br />

with TRACE (TRAC/RELAP<br />

Advanced Computational Engine) v5<br />

patch 4 BE thermal- hydraulic system<br />

code developed by USNRC [10]. The<br />

hot rod cladding temperature has been<br />

selected as a figure of merit and a<br />

limited number of input un certainty<br />

parameters, mostly related to plant<br />

initial conditions, have been chosen <strong>for</strong><br />

the UQ application.<br />

2 Description of probabilistic<br />

method to propagate<br />

input uncertainty,<br />

Dakota Toolkit, and UQ<br />

hyphothesis<br />

2.1 Probabilistic method<br />

to propagate input<br />

uncertainty<br />

The probabilistic method to propagate<br />

the input uncertainty [<strong>11</strong>] is, in brief,<br />

based on a random sampling of the<br />

input uncertain parameters selected<br />

by the user; a set of N code runs having<br />

in the input a combination of the uncertain<br />

parameters is created and<br />

solved with the selected code. Then,<br />

by using regression analysis, the effect<br />

of the input parameters uncertainty<br />

on the results is computed, in terms of<br />

selected Figure of Merits (FOMs). The<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

main advantages of this method are<br />

that the number of code runs does not<br />

depend on the number of input<br />

uncertain parameters and it is not<br />

necessary a prior development of<br />

a Phenomena Identification and<br />

Ranking Table (PIRT), which is a long<br />

process. The number of code runs N<br />

depends on the probability content<br />

and on the confidence level set by the<br />

user and on the number of FOMs<br />

selected <strong>for</strong> the analysis and it is computed<br />

using the Wilks <strong>for</strong>mula [12,<br />

13]. Each uncertain input parameter<br />

should be defined by its range of<br />

variation and its Probability Density<br />

Function (PDF); this means that even<br />

if the exact value of a parameter is<br />

uncertain, some values are more likely<br />

to be close to the real one than others.<br />

This is one of the main task to be<br />

tackled because the evaluation of the<br />

correct PDF <strong>for</strong> each parameter is not<br />

a trivial task and a careful study is<br />

required. The selection of the most<br />

suitable PDF <strong>for</strong> each parameter depends<br />

mainly on the physical quantity<br />

and on the availability of reducedscale<br />

or full-scale measured data. If<br />

data are available, it is possible to<br />

build a histogram and to derive the<br />

PDF. A comprehensive overview of<br />

PDF type and derivation techniques is<br />

provided in [14].<br />

2.2 DAKOTA toolkit in the<br />

SNAP environment/<br />

architecture<br />

DAKOTA [7] is an open source software<br />

written in C++ and developed by<br />

Sandia National Laboratories to per<strong>for</strong>m<br />

parametric and uncertainty<br />

analysis in a fast and automatic way.<br />

The aim of this toolkit is to bridge<br />

simulation codes and analysis methods<br />

<strong>for</strong> parametric evaluation, uncertainty<br />

quantification and system optimization<br />

[15]. The DAKOTA toolkit is also<br />

provided as a plug-in [9] <strong>for</strong> the<br />

Symbolic <strong>Nuclear</strong> Analysis Package<br />

(SNAP), which is a graphical user<br />

interface designed to support the use<br />

of USNRC nuclear codes (e.g. TRACE,<br />

RELAP, MELCOR, etc.). Using SNAP, it<br />

is possible to build the input deck in a<br />

graphical environment and to have a<br />

direct visualization of the code calculated<br />

data by using its animation<br />

capability. Through SNAP it is possible<br />

to set up the DAKOTA uncertainty<br />

analysis [16, 17] and to per<strong>for</strong>m automatically<br />

all the steps qualitatively<br />

described in the previous section.<br />

Figure 1 shows a schematic representation<br />

of DAKOTA uncertainty analysis<br />

workflow in a SNAP environment/<br />

architecture.<br />

In particular, DAKOTA plugin allows<br />

to:<br />

1) Enter the uncertain input parameters<br />

with their range and PDF;<br />

2) Select the sampling method<br />

( Monte Carlo or Latin Hypercube);<br />

3) Enter the desired FOMs <strong>for</strong> the<br />

analysis;<br />

4) Set the final report that contains<br />

the results of the uncertainty<br />

analysis application; the report is<br />

auto matically generated at the end<br />

of the uncertainty quantification<br />

analysis.<br />

DAKOTA is used at the beginning of<br />

the analysis to sample the uncertain<br />

input parameter values and to<br />

generate the set of code inputs. Then,<br />

after the solution of the set of code<br />

inputs and the extraction of the<br />

desired data, DAKOTA per<strong>for</strong>ms the<br />

uncertainty analysis and apply<br />

regression techniques to evaluate the<br />

correlation between input and output<br />

parameters selected as a FOM.<br />

The required number of code runs<br />

can be found solving the Wilks<br />

<strong>for</strong>mula with respect to N <strong>for</strong> a probability<br />

α and a confidence level β [<strong>11</strong>]:<br />

(1)<br />

With α=0.95 and β=0.95, <strong>for</strong> one<br />

FOM the required number of code<br />

runs is 59.<br />

2.3 UQ application hypothesis<br />

The target of this analyses, is not to be<br />

a detail uncertainty study in term of<br />

input uncertainty parameters as<br />

presented in [18], but a) to develop a<br />

full UQ application with TRACE and<br />

DAKOTA toolkit in a SNAP environment/architecture<br />

and b) to have<br />

some insights about the degree of<br />

corre lation between the input parameters<br />

selected and the FOM chosen<br />

<strong>for</strong> this analysis. Six uncertain parameters<br />

have been selected <strong>for</strong> this<br />

uncertainty application based also on<br />

BEMUSE program results [18] and<br />

through SNAP have been implemented<br />

in the DAKOTA and TRACE<br />

input: the Safety Injection System<br />

(SIS) temperature, the SIS characteristic,<br />

the accumulator initial temperature<br />

and pressure, the initial core<br />

power and the initial con tainment<br />

pressure. The SIS characteristic is a<br />

value that multiply the default injected<br />

flow rate curve as function of the<br />

primary pressure. Table 1 summarizes<br />

the uncertain input parameters,<br />

their mean value used <strong>for</strong> the reference<br />

calculation, the range of variation<br />

and the adopted PDF.<br />

One FOM was selected <strong>for</strong> the<br />

analysis, the cladding temperature of<br />

the hot rod; there<strong>for</strong>e, with a probability<br />

of 95% and a confidence level of<br />

95%, a total of 59 calculations were<br />

required based on Wilks <strong>for</strong>mula as<br />

previously described. Latin Hypercube<br />

sampling [19,20] has been used<br />

<strong>for</strong> this analysis. It is a stratified<br />

sampling method that, with respect to<br />

a pure Monte Carlo sampling, allows<br />

to achieve the target statistical<br />

| Fig. 1.<br />

DAKOTA uncertainty analysis workflow <strong>for</strong> TRACE code in a SNAP environment/architecture.<br />

RESEARCH AND INNOVATION 527<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Parameter Average value Range of variation PDF type<br />

RESEARCH AND INNOVATION 528<br />

Safety injection system (SIS) temperature [K] 285 [-10,+10] K Normal<br />

Safety injection system (SIS) characteristic [-] 1 [0.95,1.05] Normal<br />

Accumulator initial temperature [K] 325 [-10,+10] K Normal<br />

Accumulator initial pressure [bar] 40.8 [-2,+2] bar Normal<br />

Initial core power [MW] 2785 [0.98,1.02] Normal<br />

Initial containment pressure [bar] 1.013 [0.85,1.15] Uni<strong>for</strong>m<br />

| Tab. 1.<br />

Input uncertain parameters selected <strong>for</strong> the analysis.<br />

| Fig. 2.<br />

TRACE nodalization of the primary system and of the containment of the generic three-loops PWR developed with SNAP.<br />

accuracy with a minor number of<br />

samples. All 59 runs were correctly<br />

executed and they reached the end of<br />

calculation without any failure. To<br />

analyze the transient progression, the<br />

primary system pressure, the core<br />

flow rate, the cladding hot rod temperature<br />

and the vessel collapsed level<br />

have been plotted <strong>for</strong> the complete set<br />

of 59 runs, while the application of<br />

regression analysis has been carried<br />

out only <strong>for</strong> the cladding temperature.<br />

3 Generic PWR-900 TRACE<br />

model and steady state<br />

operation<br />

3.1 TRACE nodalization<br />

description<br />

TRACE, developed by USNRC, is a<br />

component-oriented code designed<br />

to per<strong>for</strong>m best-estimate thermalhydraulic<br />

analysis <strong>for</strong> LWR. It is a<br />

finite volume, two fluid, code with 3D<br />

capability and it is based on two fluid,<br />

two-phase field equations. This set of<br />

equations consists of the conservation<br />

laws of mass, momentum and energy<br />

<strong>for</strong> the liquid and gas phases [10, 21].<br />

The code version adopted in this<br />

analysis is TRACE code v5 patch 4 and<br />

the input deck has been developed<br />

with SNAP.<br />

The nodalization of a generic threeloops<br />

PWR-900, shown in Figure 2,<br />

has been developed to per<strong>for</strong>m the<br />

large break LOCA analysis. In order to<br />

minimize the computational time (in<br />

view of input uncertainty propagation<br />

with probabilistic method) and maintain<br />

an accurate prediction of target<br />

phenomena, the nodalization strategy<br />

used follows the general nodalization<br />

approach of the TRACE W4loops<br />

samples input-deck distributed with<br />

SNAP. Starting from that nodalization<br />

approach sample and considering the<br />

level of detail target of this analysis<br />

(e.g. modeling the three loops separately,<br />

modeling the interaction containment/primary<br />

coolant system,<br />

etc…), more details have been considered<br />

in the input-deck development.<br />

In the nodalization of the generic<br />

three-loop PWR-900 no lumped loops<br />

have been considered and the three<br />

loops have been modeled separately,<br />

one simulates the broken one (loop A)<br />

and two simulate the intact loops<br />

(loop B and C). The model is composed<br />

of 69 Hydraulic Components<br />

(HC) and 45 Heat Structures (HS).<br />

<strong>Power</strong> is provided to the three heat<br />

structures simulating the core through<br />

one power component, the total<br />

thermal power is around 2800 MW; in<br />

one of the HS that compose the core is<br />

inserted a supplemental rod that simulates<br />

the hot rod in the reactor with a<br />

total peaking factor of 2.278 [22]. The<br />

pressurizer is connected to the hot leg<br />

of an intact loop (loop B). The break<br />

has been modeled with a set of three<br />

valves; at the break opening one valve<br />

interrupts the connection between the<br />

two sections of the guillotine break on<br />

the cold leg of loop A; simultaneously,<br />

the other two valves connect the two<br />

closed sections of the leg to the components<br />

simulating the containment.<br />

Particular attention has been paid<br />

to the containment building modeling<br />

in order to simulate the interaction<br />

containment/primary system during<br />

the LOCA. In particular, the containment<br />

has been modeled with two connected<br />

hydraulic regions thermally<br />

coupled to the heat structure simulating<br />

the containment solid structure<br />

and the thermal interaction with<br />

the environment. Figure 2 shows<br />

the TRACE nodalization of the<br />

generic PWR primary system and the<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

| Fig. 3.<br />

TRACE nodalization of<br />

the RPV of the generic<br />

three-loops PWR developed<br />

with SNAP.<br />

| Fig. 4.<br />

SNAP animation showing the fluid condition of the primary system during the steady state (at 900 s).<br />

RESEARCH AND INNOVATION 529<br />

containment (the containment drawing<br />

is scaled down with a factor 0.2 in<br />

comparison with the other reactor<br />

components <strong>for</strong> a better visualization<br />

of the complete plant layout). The<br />

ECCS are connected on the cold leg of<br />

each loop; <strong>for</strong> each loop, the ECCS<br />

consist of an accumulator and the<br />

high pressure and low pressure safety<br />

injection systems (HPIS, LPIS). The<br />

HPIS and LPIS are modeled by a single<br />

“fill” component with a table that<br />

controls the injected flow rate as<br />

function of the pressure in the primary<br />

system. The Reactor Pressure Vessel<br />

(RPV) has been modeled by using the<br />

3D Vessel component available in<br />

TRACE. The 3D vessel nodalization<br />

has been divided into: 2 radial sectors,<br />

the inner one <strong>for</strong> the core and the<br />

outer one <strong>for</strong> the downcomer, 3 angular<br />

sectors, one <strong>for</strong> each primary loop<br />

and 7 axial sectors, 2 <strong>for</strong> the vessel<br />

lower plenum, 3 <strong>for</strong> the core and 2<br />

<strong>for</strong> the vessel upper plenum. The 3D<br />

vessel nodalization representation<br />

made with SNAP is shown in Figure 3,<br />

with the core region highlighted.<br />

3.2 Steady state<br />

characterization<br />

Be<strong>for</strong>e the beginning of the LOCA,<br />

1000 s of steady state operation have<br />

been simulated to test the nodalization<br />

in steady state condition. The<br />

steady state parameters are reported<br />

in Table 2 and they have been checked<br />

<strong>for</strong> consistency with public available<br />

in<strong>for</strong>mation [23, 24].<br />

Figure 4 shows the fluid con ditions<br />

in the primary system during the<br />

steady state; the color legend refers to<br />

the fluid status, ranging from blue<br />

(subcooled liquid) to red (superheated<br />

gas/steam). Figure 5 shows<br />

the axial pressure profile in the<br />

primary system in steady state conditions,<br />

to support the steady state<br />

qualification of the nodalization. The<br />

normalized pressure profile is consistent<br />

with other similar data available<br />

in the public scientific literature<br />

[25].<br />

4 Cold leg lbloca transient<br />

and uncertainty analysis<br />

4.1 LBLOCA transient analysis<br />

The analyzed transient is a doubleended<br />

guillotine break (200 %) in the<br />

cold leg of loop A. After 1000 s of<br />

steady state simulation the break is<br />

opened (start of the transient: t = 0 s)<br />

and 500 s of transient are simulated.<br />

The reactor SCRAM is supposed after<br />

0.5 s from the break. From ­Figure 6 to<br />

Figure 9 the results <strong>for</strong> the 59 runs are<br />

shown <strong>for</strong> the primary system pressure,<br />

the core flow rate, the cladding<br />

hot rod temperature and the vessel<br />

collapsed level. Figure 6 shows the<br />

behavior of the primary system pressure;<br />

after the LOCA initiation, water<br />

flows from the primary system to the<br />

containment and the primary system<br />

pressure drops significantly in few<br />

tens of milliseconds in agreement<br />

with the publically available technical<br />

scientific literature [26]; after this<br />

first drop of pressure and the phase<br />

of subcooled depressurization, the<br />

saturation con dition is reached and<br />

saturated depressurization starts at a<br />

reduced rate. The maximum flow rate<br />

through the break is limited by the<br />

critical velocity at the break. After the<br />

transient initiation, the flow rate in<br />

the core (Figure 7) drops from the<br />

nominal value and it is reversed since<br />

the flow is directed to the break<br />

location; there<strong>for</strong>e, water flows downwards<br />

in the core region and then<br />

upward in the downcomer to reach<br />

the break in the cold leg.<br />

The high voiding in the core and<br />

the subsequent SCRAM stop the chain<br />

reaction and, due to a sensible<br />

reduction of heat removal in the<br />

core, the heat stored in the fuel is<br />

Thermal power [MW] 2785<br />

Primary system pressure [bar] 155<br />

Total core flow rate [kg/s] 13,947<br />

Core inlet temperature [K] 558.4<br />

Core outlet temperature [K] 594.1<br />

Secondary system pressure [bar] 58<br />

Steam generator feedwater temperature [K] 440<br />

Steam generator feedwater flow rate [kg/s] 512<br />

| Tab. 2.<br />

Steady state parameters of the reference calculation.<br />

| Tab. 5.<br />

Normalized axial pressure profile along the primary system in steady state<br />

condition (at 900 s).<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

RESEARCH AND INNOVATION 530<br />

redistributed, leading to the first<br />

cladding temperature peak, as shown<br />

in Figure 8. The cladding temperature<br />

passes from the steady state value<br />

(around 617 K) to 955 K. The result<br />

of the core voiding is a sudden drop<br />

of the water collapsed level in the<br />

reactor vessel, as shown in Figure 9,<br />

under the Bottom of Active Fuel<br />

(BAF). When the pressure in the<br />

primary system is lower than the<br />

accumulator initial pressure, the<br />

accumulator check valves open and<br />

water is discharged in the primary<br />

system; this happens around <strong>11</strong> s after<br />

the transient initiation. The water<br />

initially injected in the cold legs by the<br />

accumulators bypasses the vessel<br />

lower plenum through the upper<br />

down comer region and it is directed to<br />

the break without penetrating the<br />

core. After the first fast depressurization<br />

phase, the primary pressure<br />

continues to reduce at a lower rate,<br />

equalizing the containment pressure<br />

(around 0.45 MPa after around 40 s)<br />

and ending the blowdown phase.<br />

Water is injected in the primary<br />

system by the accumulators and the<br />

low pressure safety injection system<br />

start to inject water after around 30 s<br />

from the break initiation. The refill<br />

phase starts around 40 s, when the<br />

emergency core coolant water reaches<br />

the vessel lower plenum and the<br />

collapsed level in the vessel starts to<br />

rise. During this phase the core is<br />

mainly uncovered and heat is not<br />

removed from the fuel rods, with the<br />

exception of a small amount of heat<br />

removed by thermal radiation and<br />

natural convection of the steam<br />

present in the core. For this reason<br />

during the refill period the cladding<br />

temperature increases (Figure 8) due<br />

to the quasi adiabatic heating of<br />

fission product decay. When, around<br />

50 s from the LOCA initiation, the<br />

water level reaches the core bottom<br />

(Figure 9) the refill period ends<br />

and the reflood phase starts. Water<br />

collapsed level rises quickly up to<br />

around 65 s (time of end of accumulators<br />

injection), and it continues at a<br />

lower rate due to the LPIS. In this<br />

phase the net core flow rate is positive,<br />

even if very small and with many<br />

oscillations. The water entering the<br />

core is heated up, starts to boil and<br />

entrains water droplets that help the<br />

cooling of the hottest parts of the core.<br />

With the rising of the water level, the<br />

cooling is increased and the cladding<br />

temperature starts to decrease. The<br />

complete rewetting of the cladding<br />

surface caused by the rising water<br />

level produces a strong temperature<br />

drop (core quenching). This happens<br />

around 125 s after the beginning of<br />

the LOCA.<br />

Analyzing the dispersion of the<br />

results, the primary pressure (Figure<br />

6) presents an almost negligible<br />

dispersion during the blowdown<br />

phase and the predictions of the 59<br />

runs are very similar; after the blowdown<br />

the dispersion band width is<br />

always lower than 0.1 MPa with a final<br />

average value of 0.475 MPa. As<br />

regards the core mass flow rate<br />

(Figure 7), cladding temperature<br />

(Figure 8) and vessel water collapsed<br />

level (Figure 9), the results dispersion<br />

is very limited during the blowdown<br />

phase, while it is more noticeable in<br />

the refill phase, especially <strong>for</strong> the<br />

vessel water collapsed level, and it is<br />

higher during the reflood phase.<br />

In particular, the refill initial time<br />

shows a dispersion band width of 12 s<br />

(33 – 45 s); during the reflood the<br />

collapsed level dispersion band width<br />

is around 1 m and the Top of Active<br />

Fuel (TAF) is reached in a time band<br />

of 33 s (126 – 159 s). The cladding<br />

temperature has, at the first peak, a<br />

low dispersion band width of 14 K<br />

and the peak has almost the same<br />

timing <strong>for</strong> all runs; instead, the dispersion<br />

band width is higher <strong>for</strong> the<br />

second peak (50 K) and with a time<br />

dispersion band width of 10 s. The hot<br />

rod cladding quenching time is also<br />

affected by a dispersion band width<br />

of 20 s (<strong>11</strong>5 – 135 s).<br />

| Tab. 6.<br />

Primary system pressure predicted by TRACE code.<br />

| Tab. 7.<br />

Core mass flow rate predicted by TRACE code.<br />

| Tab. 8.<br />

Cladding temperature of the hot rod predicted by TRACE code.<br />

| Tab. 9.<br />

Reactor vessel water collapsed level predicted by TRACE code.<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

| Tab. 10.<br />

Pearson’s simple regression correlation coefficients.<br />

4.2 Uncertainty analysis<br />

response correlations<br />

4.2.1 Description of regression<br />

correlations in DAKOTA<br />

As a result of the uncertainty analysis,<br />

DAKOTA [7, 8, 20] computes four<br />

response correlation coefficients:<br />

simple, partial, simple rank and<br />

partial rank. The simple coefficient is<br />

related to the actual input and output<br />

data. The simple coefficient r between<br />

an input variable x and an output<br />

variable y, in n samples, is computed<br />

using the Pearson’s correlation. It is a<br />

measure of the degree of linear correlation<br />

between the two variables and<br />

its value is comprised between -1 and<br />

1. If r0 the<br />

correlation is positive (an increment<br />

of x leads to an increment of y). The<br />

simple correlation coefficient between<br />

x and y is obtained by dividing the<br />

covariance of the two variables by the<br />

product of their standard deviations<br />

[27]:<br />

(2)<br />

The partial correlation coefficient is<br />

computed similarly to simple one but<br />

taking into accounts the effects of<br />

the other variables. This is useful, <strong>for</strong><br />

example, if there is a strong correlation<br />

between two inputs; in this way<br />

the correlation of the second input on<br />

the output may be adjusted after<br />

having considered the correlation between<br />

the first input and the output<br />

[7]. Rank correlation coefficients use<br />

the ranked data instead of the actual<br />

ones. Ranks are obtained by ordering<br />

the data in ascending order, and are<br />

more convenient to be used when<br />

inputs and outputs are characterized<br />

by sensible difference in magnitude; it<br />

is possible to understand if the input<br />

sample with the lower rank is associated<br />

to the output with the lower<br />

rank and so on [7, 20]. To compute the<br />

rank correlation, DAKOTA uses the<br />

Spearman’s rank correlation that is<br />

similar to Pearson’s one but with the<br />

ranked data instead of the actual<br />

values. If two variables are monotonically<br />

related, without repetitions,<br />

the Spearman coefficient is -1 or +1<br />

(depending if the function is monotonically<br />

decreasing or increasing),<br />

since the ranked values are used.<br />

Moreover, Spearman’s correlation is<br />

less sensitive to possible outlier values<br />

of the variables than Pearson’s one.<br />

4.2.2 Results of response correlation<br />

coefficients <strong>for</strong> the<br />

cold leg LBLOCA transient<br />

The time dependent computation of<br />

response correlation coefficients has<br />

been per<strong>for</strong>med extracting the FOM<br />

value at different selected instant of<br />

the transient evolution. Figure 10<br />

shows the Pearson’s simple correlation<br />

coefficient <strong>for</strong> the six input<br />

uncertain parameters from the LOCA<br />

beginning to the complete core<br />

quenching (after 130 s). Figure <strong>11</strong><br />

shows the Spearman’s rank correlation<br />

coefficient <strong>for</strong> the same parameters.<br />

On both graphs the values 0.2<br />

and 0.5 (and -0.2 and -0.5) have been<br />

highlighted as measure of the correlation<br />

between the input parameter<br />

and the FOM. As indicated in the [20],<br />

<strong>for</strong> the Spearman coefficient, if the<br />

coefficient is higher than 0.5 (or lower<br />

than -0.5) the correlation is significant,<br />

if it is between 0.2 and 0.5 (or<br />

-0.2 and -0.5) the correlation is<br />

moderate, otherwise it is low [20]. In<br />

| Tab. <strong>11</strong>.<br />

Spearman’s simple rank regression correlation coefficients.<br />

the following analysis the same<br />

threshold values have been adopted<br />

also <strong>for</strong> the Pearson coefficient. In this<br />

application, the two response correlations<br />

show similar trends; the main<br />

advantage of this time dependent<br />

analysis is the possibility to have a<br />

measure and characterize the correlation<br />

of the different input parameters<br />

on the uncertainty of the<br />

selected FOM in all the phases of the<br />

transient.<br />

In the blowdown phase (0-40 s) the<br />

effect of the initial core power on the<br />

hot rod cladding temperature is<br />

positive due to the heat stored in the<br />

solid structures. Both Pearson’s and<br />

Spearman’s coefficients are higher<br />

than 0.8 <strong>for</strong> this parameter so the<br />

correlation is significant and almost<br />

monotonically linear. In the remaining<br />

part of the transient both coefficients<br />

are close to 0.2 so the corre lation is<br />

much weaker than in the initial phase.<br />

After around 20 s from the LOCA<br />

initiation the accumulators initial<br />

pressure has a negative Pearson’s and<br />

Spearman’s coefficients around -0.3;<br />

there<strong>for</strong>e the lower is the accumulator<br />

pressure the higher is the cladding<br />

temperature since the accumulator<br />

starts to inject water later. In the refill<br />

phase (40-50 s), also the SIS characteristic<br />

has both response correlation<br />

coefficients around -0.3; in fact, with a<br />

lower injection flow rate the hot rod<br />

cladding temperature is higher. In the<br />

reflood phase (after 50 s) also the<br />

initial containment pressure has negative<br />

Pearson’s and Spearman’s coefficients<br />

around -0.35. This is due to the<br />

fact that with a lower pressure in the<br />

containment a greater amount of<br />

coolant water is expelled through<br />

the break and the cladding temperature<br />

is higher. The uncertainty in the<br />

accumulators and SIS temperature<br />

have a low correlation with the FOM.<br />

RESEARCH AND INNOVATION 531<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

RESEARCH AND INNOVATION 532<br />

5 Conclusions<br />

Uncertainty analysis is used in the<br />

nuclear sector to evaluate uncertainties<br />

still present in the results of code’s<br />

calculations. Among the several<br />

methodologies developed in the past<br />

to per<strong>for</strong>m Uncertainty Analysis, the<br />

probabi listic method to propagate the<br />

input uncertainties has been selected<br />

<strong>for</strong> this analysis considering its<br />

suitability to be coupled with simulation<br />

codes; moreover, several<br />

toolkits, some integrated in computational<br />

plat<strong>for</strong>ms, have been developed<br />

<strong>for</strong> this purpose. In the present<br />

activity, a double-ended cold leg<br />

LBLOCA transient has been simulated<br />

with the BE thermal hydraulic system<br />

code TRACE <strong>for</strong> a generic three-loops<br />

PWR-900 reactor. Using DAKOTA<br />

toolkit, in a SNAP environment/architecture,<br />

an uncertainty analysis has<br />

been carried out by selecting six<br />

uncertain input parameters and the<br />

hot rod cladding temperature as the<br />

main figure of merit. In addition, the<br />

primary pressure, the core flow rate<br />

and the pressure vessel collapsed level<br />

have been analyzed to evaluate the<br />

transient progression and the results<br />

dispersion. The aim of this analysis is<br />

not to be a detail and exhaustive<br />

uncertainty study in term of input<br />

uncertainty parameters but to develop<br />

a complete uncertainty quantification<br />

application with DAKOTA in a SNAP<br />

environment/architecture and to have<br />

some insights characterizing the<br />

correlation between the input uncertainty<br />

parameters and the selected<br />

FOM. Pearson’s and Spearman’s<br />

response correlation coefficients have<br />

been computed between the LOCA<br />

initiation and the complete core<br />

quenching. In the blowdown phase,<br />

the hot rod cladding temperature has<br />

a significant correlation with the<br />

initial core power; the accumulators’<br />

initial pressure has a moderate correlation<br />

with the FOM only in the period<br />

of water injection from the accumulators.<br />

The time dependent response<br />

analysis, adopted in this application,<br />

is very useful since it could be used to<br />

characterize the effect of the uncertain<br />

input parameters on the output<br />

global uncertainty in the different<br />

phases of a transient. In a future<br />

follow-up, additional uncertain input<br />

parameters and FOMs can be introduced<br />

in the analysis in order to have<br />

a more complete evaluation of the<br />

results uncertainty.<br />

Acknowledgement<br />

The authors are grateful to Ms Cristina<br />

Bertani <strong>for</strong> the review of the manuscript.<br />

Nomenclature<br />

N<br />

n<br />

p<br />

r<br />

x<br />

y<br />

α<br />

β<br />

Subscripts<br />

m<br />

Number of code runs<br />

Number of samples of a certain variable<br />

Number of Figure of Merit<br />

Pearson’s correlation coefficient<br />

Input variable in the computation of response correlations<br />

Output variable in the computation of response correlations<br />

Probability<br />

Confidence level<br />

mean value of the related parameter<br />

References<br />

1. F. Mascari, G. Vella, B.G. Woods, K. Welter, F. D’Auria, “Analysis<br />

of Primary/Containment Coupling Phenomena Characterizing<br />

the MASLWR Design During a SBLOCA Scenario”, <strong>Nuclear</strong><br />

<strong>Power</strong> Plant, Intech (2012)<br />

2. F. Mascari, H. Nakamura, F. De Rosa, F. D’ Auria, “Scaling<br />

Rationale Design and Extrapolation Problem <strong>for</strong> ITF and SETF”,<br />

Book of Abstract of <strong>International</strong> Workshop on <strong>Nuclear</strong> Safety<br />

and Severe Accident (NUSSA-2014), Kashiwa, Chiba, Japan,<br />

September 3-5 (2014)<br />

3. F. Mascari, H. Nakamura, K. Umminger, F. De Rosa, F. D’Auria,<br />

“Scaling Issues For The Experimental Characterization Of<br />

Reactor Coolant System In Integral Test Facilities And Role Of<br />

System Code As Extrapolation Tool”, Proceedings Of<br />

<strong>International</strong> Topical Meeting on <strong>Nuclear</strong> Reactor Thermal<br />

Hydraulics 2015, NURETH 2015, Volume 6, pp. 4921-4934<br />

(2015)<br />

4. W.L. Oberkampf, T.G. Trucano, “Verification and validation<br />

benchmarks”, <strong>Nuclear</strong> Engineering and Design 238,<br />

pp. 716-743 (2008)<br />

5. NEA/CSNI “A state-of-the-art report on scaling in system<br />

thermal hydraulics applications to nuclear reactor safety and<br />

design”, NEA/CSNI/R(2016)14 (2017)<br />

6. IAEA <strong>International</strong> Atomic Energy Agency “Best Estimate<br />

Safety Analysis <strong>for</strong> <strong>Nuclear</strong> <strong>Power</strong> Plants: Uncertainty<br />

Evaluation”, Safety Reports Series (2008)<br />

7. B.M. Adams, M.S. Ebeida, M.S. Eldred, G. Geraci, J.D. Jakeman,<br />

K.A. Maupin, J.A. Monschke, J.A. Stephens, L.P. Swiler,<br />

D.M. Vigil, T.M. Wildey, W.J. Bohnhoff, K.R. Dalbey, J.P. Eddy,<br />

J.R. Frye, R.W. Hooper, K.T. Hu, P.D. Hough, M. Khalil,<br />

E.M. Ridgway, A. Rushdi, Dakota, A Multilevel Parallel<br />

Object-Oriented Framework <strong>for</strong> Design Optimization,<br />

Parameter Estimation, Uncertainty Quanti fication, and<br />

Sensitivity Analysis: Version 6.7 User’s Manual,<br />

SAND2014-4633 (2017)<br />

8. Applied Programming Technology, Inc. Symbolic <strong>Nuclear</strong><br />

Analysis Package (SNAP) User’s Manual (2012)<br />

9. Applied Programming Technology, Inc. Uncertainty analyses<br />

User manual, Symbolic <strong>Nuclear</strong> Analyses Package (SNAP),<br />

(2012)<br />

10. U.S. <strong>Nuclear</strong> Regulatory Commission TRACE v5.840 Theory<br />

Manual (2013)<br />

<strong>11</strong>. H. Glaeser, “GRS Method <strong>for</strong> Uncertainty and Sensitivity<br />

Evaluation of Code Results and Applications”, Science and<br />

Technology of <strong>Nuclear</strong> Installations (2008)<br />

12. S.S. Wilks, “Determination of sample sizes <strong>for</strong> setting tolerance<br />

limits”, The Annals of Mathematical Statistics 12(1), pp. 91-96<br />

(1941)<br />

13. S.S. Wilks, “Statistical prediction with special reference to the<br />

problem of tolerance limits”, The Annals of Mathematical<br />

Statistics 13(4), pp. 400-409 (1942)<br />

14. M.E. Stephens, B.W. Goodwin, T.H. Andres, “Deriving<br />

parameter probability density functions”, Reliability<br />

Engineering & System Safety 42, pp. 271-291 (1993)<br />

15. https://dakota.sandia.gov/<br />

16. J.R. Wang, C.W. Tsai, H.T. Lin, C. Shih, “Per<strong>for</strong>ming Uncertainty<br />

Analysis of IIST facility SBLOCA by TRACE and DAKOTA”,<br />

NUREG/IA-0428 (2013)<br />

17. J.R. Wang, J.H. Yang, H.T. Lin, C. Shih, “Uncertainty Analysis <strong>for</strong><br />

Maanshan LBLOCA by TRACE and DAKOTA”, NUREG/IA-0448<br />

(2014)<br />

18. M. Perez, F. Reventos, L. Batet, A. Guba, I. Tóth, T. Mieusset,<br />

P. Bazin, A. de Crécy, S.Borisov, T. Skorek, H. Glaeser, J. Joucla,<br />

P. Probst, A. Ui, B.D. Chung, D.Y. Oh, R. Pernica, M. Kyncl,<br />

J. Macek, A. Manera, J. Freixa, A. Petruzzi, F. D’Auria,<br />

A. Del Nevo, “Uncertainty and sensitivity analysis of a LBLOCA<br />

in a PWR <strong>Nuclear</strong> <strong>Power</strong> Plant: Results of the Phase V of the<br />

BEMUSE programme”, <strong>Nuclear</strong> Engineering and Design 241,<br />

pp. 4206-4222 (20<strong>11</strong>)<br />

19. J.C. Helton, F.J. Davis “Latin hypercube sampling and the<br />

propagation of uncertainty in analyses of complex systems”,<br />

Reliability Engineering & System Safety 81, pp. 23-69 (2003)<br />

20. K.A. Gamble, L.P. Swiler “Uncertainty Quantification and<br />

Sensitivity Analysis Applications to Fuel Per<strong>for</strong>mance<br />

Modeling”, SAND2016-4597C<br />

21. F. Mascari, F. De Rosa, B.G. Woods, K. Welter, G. Vella,<br />

F. D’Auria, “Analysis of the OSU-MASLWR 001 and 002 Tests by<br />

Using the TRACE Code”, NUREG/IA-0466 (2015)<br />

22. T. Iguchi, T. Okubo, Y. Murao, “Effect of Loop Seal on Reflood<br />

Phenomena in PWR”, <strong>Journal</strong> of <strong>Nuclear</strong> Science and<br />

Technology 25(6), pp. 520-527 (1988)<br />

23. P. Coppolani, N. Hassenboehler, J. Joseph, J.F. Petetrot, J.P. Py,<br />

J.S. Zampa, La chaudière des réacteurs à eau sous pression,<br />

EDP Sciences, Les Ulis, France (2004)<br />

24. F. Mascari, J.C. De La Rosa Blul, M. Sangiorgi, G. Bandini,<br />

“Analyses of an unmitigated station blackout transient with<br />

ASTEC, MAAP and MELCOR code”, 9th Meeting of the<br />

“ European MELCOR User Group” (2017)<br />

25. F. D’Auria, M. Frogheri, W. Giannotti “RELAP/MOD3.2 Post Test<br />

Analysis and Accuracy Quantification of SPES Test SP-SB-04”,<br />

NUREG/IA-0155 (1999)<br />

26. O.C. Jones, <strong>Nuclear</strong> Reactor Safety Heat Transfer, Hemisphere<br />

Publishing Corporation, Washington, USA (1981)<br />

27. J.L. Rodgers, W.A. Nicewander, “Thirteen Ways to Look at the<br />

Correlation Coefficient”, The American Statistician 42(1),<br />

pp. 59-66<br />

Authors<br />

Andrea Bersano<br />

Energy Department (DENERG)<br />

Politecnico di Torino<br />

Corso Duca degli Abruzzi 24<br />

10129, Turin<br />

Italy<br />

Fulvio Mascari<br />

<strong>Nuclear</strong> Safety<br />

Security and Sustainability Division<br />

(FSN-SICNUC), ENEA<br />

Via Martiri di Monte Sole 4<br />

40129, Bologna<br />

Italy<br />

Research and Innovation<br />

Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Experiment Research on the Insurge<br />

Transient Behavior of Gas-steam<br />

Pressurizer under Various Pressure<br />

Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun<br />

For small and medium sized reactors with gas-steam pressurizer, the transient behavior of gas-steam pressurizer plays<br />

a vital role on the safety of the nuclear reactor operation. This paper focuses on the transient behavior of gas-steam<br />

pressurizer. The pressure response was investigated under various insurge experimental conditions, and the influence<br />

factors of pressure changes were analyzed.<br />

Research shows initial pressure of the pressurizer and the presence of a non-condensable gas with varying mass<br />

fraction will all have some kind of effect on the transient behavior of gas-steam pressurizer. Initial pressure decides initial<br />

magnitude of liquid temperature of the pressurizer, and the presence of a non-condensable gas with varying mass<br />

fraction can greatly affect the heat and mass transfer process both wall and interface, and further affect the system<br />

pressure variations. This paper focus on the effects on pressurizer insurge transients under various initial pressure of the<br />

pressurizer and the presence of a non-condensable gas with varying mass fraction.<br />

Introduction<br />

A transient can be caused from a<br />

simple loss of secondary steam flow to<br />

more complicated accidents. After<br />

the Three Mile Island accident [1],<br />

Chernobyl accident [2] and Fukushima<br />

nuclear accident [3]. Many people<br />

realized the importance of small break<br />

LOCA’s and the necessity <strong>for</strong> having<br />

reliable physical models <strong>for</strong> all of the<br />

components in the loop.<br />

In view of increasing pressure on<br />

energy restructuring and the serious<br />

environmental pollution concerns,<br />

the integrated natural circulation<br />

reactor is receiving a great attention<br />

<strong>for</strong> its ability to provide energy that is<br />

clean, safe and economic. And the integrated<br />

reactor is going to be one of<br />

the best option <strong>for</strong> small and medium<br />

sized reactors (SMRs) in lots of countries<br />

with the high level of great safety<br />

and reliability. SMRs have been developed<br />

in many countries <strong>for</strong> small-scale<br />

power generation, district heating,<br />

and seawater desalination. Argentine<br />

CAREM reactor, Russian VVER-300/<br />

VK-300 reactor, Korean REX-10/<br />

SMART reactor, Japanese IMR reactor,<br />

American NuScale reactor, Chinese<br />

NHR reactor and so on, these are the<br />

typical representatives of the SMRs<br />

[4]. In order to simplify the structure<br />

and design and enhance safety, the<br />

steam-gas pressurizer is generally<br />

utilized in the integrated reactor. The<br />

non-condensable gas is used to keep<br />

the pressure stable in the steam-gas<br />

pressurizer and the transient behavior<br />

of gas-steam pressurizer plays a vital<br />

role on the safety of the nuclear reactor<br />

operation. As so far, early pressurizer<br />

transient models were generally<br />

developed under certain conditions,<br />

slow insurge velocity or low pressure<br />

<strong>for</strong> example. And these models may<br />

not be applicable to various velocity or<br />

create a greater risk of inaccurate<br />

results under the high pressure.<br />

Westinghouse [5] developed the<br />

TOPS pressurizer model included<br />

the effects of wall condensation by applying<br />

Nusselt Laminar film theory<br />

to estimate a wall heat transfer<br />

coefficient. Saedi [6] investigated the<br />

relative magnitudes of the physical associated<br />

with insurge transients, and<br />

initiated a data base (at low pressure)<br />

<strong>for</strong> a model, developed<br />

by Kim [7], at MIT. In some other<br />

pressurizer analysis, Mark [8] came<br />

up with the effects of the presence of a<br />

non-condensable gas on insurge<br />

transient. Leonard [9] per<strong>for</strong>med the<br />

experiment on the pressure behavior<br />

of steam-gas pressurizer during the<br />

insurge and he focused on the various<br />

non-condensable gas, various gas<br />

content and stratification in his study.<br />

Kim [10] found the condensation heat<br />

transfer at wall is an important physical<br />

phenomenon during the insurge<br />

transient. Paulsen [<strong>11</strong>] improved and<br />

developed the theoretical modeling<br />

<strong>for</strong> RELAP5. Wu lei [12, 13] and Ma<br />

Xizhen [14, 15] established the nonequilibrium<br />

gas-steam pressurizer<br />

model and improved the steam condensation<br />

heat transfer model in<br />

presence of non-condensable gas.<br />

In order to understand and model<br />

an accident, one should recognize the<br />

processes that take place during a<br />

transient. In general, these processes<br />

include insurges, outsurges, and combined<br />

insurges and outsurges. Early<br />

pressurizer transient models were<br />

generally developed under certain<br />

conditions, slow insurge flow rate or<br />

low pressure <strong>for</strong> example, and these<br />

models may not be applicable under<br />

various pressure. At present, the<br />

transient behavior of steam-gas pressurizer<br />

under various pressure need to<br />

be further studied in the integrated<br />

natural circulation reactor with<br />

steam-gas pressurizer. Through the<br />

design and establish of the experiment<br />

system, the combination of<br />

theoretical and experimental which<br />

can provide the data support <strong>for</strong> the<br />

design and operation of the integrated<br />

natural circulation reactor.<br />

1 Description of<br />

Experimental System<br />

To find out the experimental physical<br />

processes occurs in a pressurizer, an<br />

experimental apparatus has been<br />

built (Figure 1).<br />

The primary tank (2420 mm high<br />

and 450 mm ID) which models the<br />

pressurizer volume equipped with a<br />

magnetic level gauge to accurately<br />

measure the initial liquid level and<br />

the change of liquid level during the<br />

insurge. The storage tank (2490 mm<br />

high and 400 mm ID) is filled with<br />

cold water to a level also measured<br />

with a magnetic level gauge and<br />

pressurized with nitrogen. This tank<br />

| Fig. 1.<br />

Schematic diagram of apparatus.<br />

Nomenclature<br />

p t system pressure<br />

T w temperature at each<br />

measured point of gasphase<br />

space<br />

p N2<br />

nitrogen partial pressure<br />

p s steam partial pressure<br />

n N2<br />

nitrogen amount of<br />

substance<br />

n s steam amount of substanc<br />

V g gas-phase space volume<br />

Z N2<br />

nitrogen compressibility<br />

factor<br />

Z s<br />

steam compressibility<br />

factor<br />

x N2<br />

nitrogen mole fraction<br />

T g average temperature in<br />

gas-phase space<br />

RESEARCH AND INNOVATION 533<br />

Research and Innovation<br />

Experiment Research on the Insurge Transient Behavior of Gas-steam Pressurizer under Various Pressure ı Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

RESEARCH AND INNOVATION 534<br />

| Fig. 2.<br />

Heating rod installation method and voltage regulator.<br />

| Fig. 3.<br />

Data acquisition system and computer data network.<br />

serves as a reservoir <strong>for</strong> “cold” injection<br />

water.<br />

Two tanks are both equipped with<br />

nine heating rods totaling 36 kW and<br />

the electric heating system utilizes the<br />

TSGC2J-40 type voltage regulator to<br />

adjust the load voltage of the heating<br />

rod, and then control the heating rod<br />

power to meet the requirements<br />

of each experimental condition<br />

(Figure 2).<br />

Besides, the experimental system<br />

also consists of electric heating system<br />

and data measurement and acquisition<br />

system (Figure 3).<br />

2 Experimental conditions<br />

and parameter control<br />

Several precautionary procedures were<br />

listed as followed prior to initiating<br />

each of the transient experiments. The<br />

primary and storage tanks were filled<br />

to the required levels with deionized<br />

water. Steam was bubbled through the<br />

primary tank to assist the heating rods<br />

in bringing the tank up to saturation<br />

conditions. The relief valve was opened<br />

and the water allowed to boil <strong>for</strong> a<br />

while when the primary tank reached<br />

saturation. This process was intended<br />

to rid the system of dissolved gases in<br />

the water. For the experiments with an<br />

initial non-condensable gas fraction in<br />

the vapor, the gas was injected into the<br />

vapor space after the tank had been<br />

purged of dissolved gases, and be<strong>for</strong>e<br />

the system had reached the desired<br />

operating pressure.<br />

The partial pressure of the steam<br />

and the temperature of gas-phase<br />

space, which are directly determined<br />

by nitrogen mole fraction in the<br />

pressurizer. There<strong>for</strong>e, the experiment<br />

should put strict controls on its<br />

initial value. Based on the real gas<br />

equations, the nitrogen mole fraction<br />

can be obtained by measuring the<br />

temperature of gas-phase space and<br />

system pressure. It should be noted<br />

that the calculation leaves out of<br />

account stratification of hot gas and<br />

steam superheating phenomenon.<br />

The concrete calculating methods are<br />

as follows: measurement parameters:<br />

system pressure p t ,the temperature at<br />

each measured point of gas-phase<br />

space in the pressurizer T w1 , T w2 … T wn .<br />

Based on the real gas equations,<br />

the nitrogen mole fraction can be<br />

obtained by measuring the temperature<br />

of gas-phase space and system<br />

pressure. It should be noted that the<br />

calculation leaves out of account<br />

stratification of hot gas and steam<br />

superheating phenomenon. The<br />

concrete calculating methods are as<br />

follows:<br />

Measurement parameters: system<br />

pressure, the temperature at each<br />

measured point of gas-phase space in<br />

the pressurizer T w1 , T w2 … T wn .<br />

Middle parameters: nitrogen<br />

partial pressure p N2<br />

, steam partial<br />

pressure p s , nitrogen amount of<br />

substance n N2<br />

, steam amount of substance<br />

n s , gas-phase space volume V g ,<br />

nitrogen compressibility factor Z N2<br />

,<br />

steam compressibility factor Z s .<br />

The average temperature T g in gasphase<br />

space can be obtained by<br />

measuring the temperature at each<br />

measured point of gas-phase space<br />

in the pressurizer.<br />

The real gas equations of steam and<br />

nitrogen are as follows where R indicates<br />

gas constant:<br />

Then, it can come to the nitrogen mole<br />

fraction X n2<br />

.<br />

=<br />

| Fig. 4.<br />

Computing framework.<br />

The computing framework is demonstrated<br />

by Figure 4.<br />

Research and Innovation<br />

Experiment Research on the Insurge Transient Behavior of Gas-steam Pressurizer under Various Pressure ı Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Parameter Value Unit<br />

Non-condensable gas species Nitrogen /<br />

Non-condensable gas mass fraction 0, 10%, 20% /<br />

Regulator space initial pressure 0.5, 1.0, 1.5 MPa<br />

Initial height of liquid level 800 mm<br />

Final height of liquid level 1200 mm<br />

Insurge time 20 s<br />

| Tab. 1.<br />

Specific experimental conditions.<br />

The data acquisition system was<br />

initiated prior to beginning the<br />

insurge to allow some steady data to<br />

be taken, the specific experimental<br />

conditions are shown in Table 1.<br />

3 Experimental results and<br />

analysis<br />

There are two main reasons affected<br />

the pressure in a transient experiment,<br />

the first part is the increase of<br />

liquid level plays a role that compresses<br />

gas-phase space and the<br />

pressure is going up, the second part is<br />

a higher initial pressure leading to a<br />

higher initial liquid temperature of<br />

the pressurizer and affect the heat and<br />

mass transfer process both wall and<br />

interface, and further affect the system<br />

pressure variations. The pressure<br />

variation is a combination of the two<br />

aspects.<br />

The pressure and temperature<br />

histories <strong>for</strong> the experiments with<br />

20 % mass fraction nitrogen are<br />

shown in Figure 5 and Figure 6. To<br />

make conclusions of trials more<br />

comparable, this analysis applies<br />

dimensionless method to the system<br />

pressure, and pressure variation<br />

was represented by the ratio of<br />

system pressure to initial pressure.<br />

Figure 7 shows the dynamic change<br />

process of dimensionless pressure<br />

corresponding to different initial<br />

pressures.<br />

The system pressure growth rate<br />

and the initial pressure of the pressurizer<br />

have positive correlation. The<br />

higher the initial pressure, the faster<br />

the system pressure growth rate in the<br />

transient.<br />

The presence of a non-condensable<br />

gas with varying mass fraction will all<br />

have some kind of effect on the<br />

transient behavior of gas-steam<br />

pressurizer. At the same system pressure,<br />

the main effects of the presence<br />

of a non-condensable gas are given as<br />

follows:<br />

I<br />

Affect the nitrogen partial pressure,<br />

the steam partial pressure<br />

and the vapor temperature of the<br />

pressurizer.<br />

II Affect the compressibility of the<br />

vapor space in the pressurizer.<br />

III Affect the heat and mass transfer<br />

process of the vapor space in the<br />

pressurizer.<br />

The pressure histories <strong>for</strong> the experiments<br />

with various mass fraction<br />

nitrogen are shown in Figure 8,<br />

Figure 9 and Figure 10. The initial<br />

primary tank non-condensable gas<br />

mass fraction <strong>for</strong> the transient is<br />

labeled in the figure. The most obvious<br />

difference in the results is the large<br />

variation in peak pressure, the higher<br />

initial non-condensable gas mass<br />

fraction in the primary tank, the<br />

higher the peak pressure in the<br />

transient. And the specific values are<br />

shown in Table 2.<br />

RESEARCH AND INNOVATION 535<br />

| Fig. 5.<br />

The system pressure histories with 20 % mass fraction nitrogen.<br />

| Fig. 6.<br />

The temperature histories with 20 % mass fraction nitrogen.<br />

| Fig. 7.<br />

The dimensionless pressure histories with 20 % mass fraction nitrogen.<br />

| Fig. 8.<br />

The pressure histories with various mass fraction nitrogen under 0.5MPa pressure.<br />

Research and Innovation<br />

Experiment Research on the Insurge Transient Behavior of Gas-steam Pressurizer under Various Pressure ı Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

RESEARCH AND INNOVATION 536<br />

Initial<br />

system pressure<br />

| Tab. 2.<br />

Specific experimental values.<br />

| Fig. 9.<br />

The pressure histories with various mass fraction nitrogen under 1.0 MPa<br />

pressure.<br />

Non-condensable gas<br />

mass fraction<br />

Peak<br />

pressure<br />

0.5 0 0.55<br />

0.5 10 % 0.59<br />

0.5 20 % 0.61<br />

1.0 0 1.08<br />

1.0 10 % 1.21<br />

1.0 20 % 1.25<br />

1.5 0 1.69<br />

1.5 10 % 1.87<br />

1.5 20 % 1.92<br />

In particular, the research demonstrated<br />

that when the pressurizer is<br />

free from non-condensable gas,<br />

may present the twice peak pressure<br />

phenomenon during the transient,<br />

and the first peak pressure will present<br />

about the same time. While only<br />

one peak pressure will present during<br />

the transient containing the noncondensable<br />

gas.<br />

There are two main reasons<br />

affected the pressure in a transient<br />

experiment, the first part is the increase<br />

of liquid level plays a role that<br />

compresses gas-phase space and the<br />

pressure is going up, the second part is<br />

steam condensation leading to a decrease<br />

pressure at some degrees. The<br />

peak pressure is combination of the<br />

two aspects. In the early stage of the<br />

transient, the compression of the<br />

vapor space plays leading roles and<br />

the pressure is going up. At this stage,<br />

there was no distinct effect on the temperature<br />

of the interface by incoming<br />

the cold water during the transient. In<br />

the second stage, steam condensation<br />

leads to a decrease pressure at some<br />

degrees due to the temperature reduction<br />

of the interface. And in next stage,<br />

the compression of the vapor space<br />

also plays leading roles comparing<br />

with the steam condensation and the<br />

pressure is going up.<br />

Conclusion<br />

Based on the proceeding analysis and<br />

experiments, an experimental study<br />

of the pressure response during an<br />

insurge transient under various pressure<br />

has been per<strong>for</strong>med. The following<br />

conclusions are drawn based upon<br />

the experimental and analytical<br />

results:<br />

I<br />

The system pressure growth rate<br />

and the initial pressure of the pressurizer<br />

have positive correlation.<br />

The higher the initial pressure, the<br />

faster the system pressure growth<br />

rate in the transient.<br />

II The changes in system temperature<br />

coincided well with those of<br />

system pressure.<br />

III The higher initial non-condensable<br />

gas mass fraction in the primary<br />

tank, the higher the peak pressure<br />

in the transient.<br />

IV In particular, when the pressurizer<br />

is free from non-condensable gas,<br />

may present the twice peak pressure<br />

phenomenon during the transient,<br />

and the first peak pressure<br />

will present about the same time.<br />

While only one peak pressure will<br />

present during the transient containing<br />

the non-condensable gas.<br />

Acknowledgments<br />

Special thanks should go to Mr Ma<br />

Xizhen who have put considerable<br />

time and ef<strong>for</strong>t into the designing and<br />

constructing work of the experiment<br />

plant.<br />

References<br />

[1]. Bot P L. Human reliability data, human error and accident<br />

models—illustration through the Three Mile Island accident<br />

analysis[J]. Reliability Engineering & System Safety, 2004,<br />

83(2):153-167.<br />

[2]. Jr H T P. Summary report on the post-accident review meeting<br />

on the chernobyl accident : <strong>International</strong> <strong>Nuclear</strong> Safety Advisory<br />

Group. <strong>International</strong> Atomic Energy Agency (IAEA) Safety Series<br />

No. 75-INSAG-1 (STI/PUB/ 740), Vienna, IAEA, 1986. 260 Austrian<br />

Schillings[J]. <strong>Journal</strong> of Environmental Radioactivity, 1987,<br />

5(5):403-404.<br />

| Fig. 10.<br />

The pressure histories with various mass fraction nitrogen under 1.5 MPa<br />

pressure.<br />

[3]. Kinoshita N, Sueki K, Sasa K, et al. Assessment of individual<br />

radionuclide distributions from the Fukushima nuclear accident<br />

covering central-east Japan[J]. Proceedings of the National<br />

Academy of Sciences of the United States of America, 20<strong>11</strong>,<br />

108(49):19526-19529.<br />

[4]. <strong>International</strong> Atomic Energy Agency: Advances in small modular<br />

reactor technology developments. Vienna: the IAEA in Austria,2014.<br />

[5]. Redfield, J. A., Prescop, V., & Margolis, S. G. (1968). Pressurizer<br />

per<strong>for</strong>mance during loss-of-load tests at shippingport: analysis<br />

and test. , 4(3), 173-181.<br />

[6]. Saedi, H. R. (1982). Insurge pressure response and heat<br />

transfer <strong>for</strong> PWR pressurizer. Massachusetts Institute of Technology.<br />

[7]. Kim, S. N. (1984). An experimental and analytical model of a<br />

pwr pressurizer during transients. British <strong>Journal</strong> of Surgery,<br />

87(12), 1615–1616.<br />

[8]. Leonard, M. T., & Griffith, P. (1983). The effects of a<br />

noncondensable gas on pressurizer insurge transients. Trans.<br />

Am. Nucl. Soc.; (United States), 46(6), 844-845. NOMURA Katsuya,<br />

K.S.O.Y., Numerical analysis of droplet breakup behavior using<br />

particle method. 1999. 38(12): p. 1057-1064.<br />

[9]. Leonard M T, Griffith P. The effects of a non-condensable gas<br />

on pressurizer insurge transients[J]. Trans. Am. Nucl. Soc.; (United<br />

States), 1984, 46(6):844-845.<br />

[10]. Kim, S.N., Griffith, P., 1987. PWR pressurizer modeling. Nucl.<br />

Eng. Des.102, 199–209.<br />

[<strong>11</strong>]. Paulsen, M.P., et al., 1996. RETRAN-3D—a program <strong>for</strong> transient<br />

thermal–hydraulic analysis of complex fluid flow systems.<br />

Electric <strong>Power</strong> Research Institute, NP-7450.<br />

[12]. WU Lei JIA Hai-jun LIU Yang MA Xi-zhen. Transient Characteristics<br />

of Integrated Non-condensable Gas-steam Pressurizer [J],<br />

Atomic Energy Science and Technology, 2014, 48(s1):200-207.<br />

[13]. WU Lei LIU Yang JIA Hai-jun YANG Xing-tuan. Research on<br />

Steam Condensation Heat Transfer Model in Presence of Noncondensable<br />

Gas at High Pressure [J]. Atomic Energy Science and<br />

Technology, 2016, 50(2):261-266.<br />

[14]. MA Xi-zhen JIA Hai-jun LIU Yang WU Lei. Numerical<br />

Simulation Study on Steam Condensation in Presence of Noncondensable<br />

Gas [J]. Atomic Energy Science and Technology,<br />

2015, 49(s1):265-269.<br />

[15]. MA Xi-zhen JIA Hai-jun LIU Yang WU Lei. Effect of Noncondensable<br />

Gas on Steam-gas Pressurizer [J]. Atomic Energy<br />

Science and Technology, 2016, 50(9):1586-1591.<br />

Authors<br />

Wang Bolong<br />

Li Weihua<br />

Jia Haijun<br />

Li Jun<br />

Zhang Yajun<br />

Institute of <strong>Nuclear</strong> and New Energy<br />

Technology<br />

Tsinghua University<br />

Beijing 100084, China<br />

Collaborative Innovation Center of<br />

Advanced <strong>Nuclear</strong> Energy Technology<br />

Tsinghua University<br />

Beijing 100084, China<br />

Key Laboratory of Advanced Reactor<br />

Engineering and Safety of Ministry of<br />

Education<br />

Tsinghua University<br />

Beijing 100084, China<br />

Research and Innovation<br />

Experiment Research on the Insurge Transient Behavior of Gas-steam Pressurizer under Various Pressure ı Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Stilllegung und Rückbau des<br />

Rossendorfer Forschungsreaktors RFR<br />

Teil 1: Objektbeschreibung, Genehmigungsverfahren, Ausgangssituation,<br />

Planungskonzept und Meilensteine<br />

Reinhard Knappik, Klaus Geyer, Sven Jansen und Cornelia Graetz<br />

Mit der Entlassung des Rossendorfer Forschungsreaktors im September 2019 aus dem Geltungsbereich des Atomgesetzes<br />

(AtG) sind die Stilllegung und der Rückbau der alten kerntechnischen Anlagen des ehemaligen Zentralinstituts für Kern<strong>for</strong>schung<br />

(ZfK) der Akademie der Wissenschaften der DDR bis auf ein ca. 50 m langes Reststück einer Rohrleitung der<br />

Speziellen Kanalisation abgeschlossen.<br />

In der zweiteiligen Veröffentlichung werden die Arbeiten zur Stilllegung und zum Rückbau des Rossendorfer<br />

Forschungsreaktors (RFR) übersichtsartig vorgestellt, wobei im Teil 1 (<strong>atw</strong> <strong>11</strong>-12 2019) anknüpfend an die Objektbeschreibung<br />

die Genehmigungsverfahren, die Ausgangssituation radiologisch und konventionell, das realisierte Planungskonzept<br />

sowie die Meilensteine vor gestellt werden. Im Teil 2 (<strong>atw</strong> 1/2020) wird auf ausgewählte Aspekte der Stilllegung- und<br />

Rückbaudurchführung, zum Strahlenschutz, zur Freigabe sowie zum Reststoff- und Abfallmanagement eingegangen.<br />

1 Einleitung Die Geschichte des Rossendorfer Forschungsreaktors und damit des Forschungsstandortes<br />

Rossendorf bei Dresden (jetzt Dresden-Rossendorf) begann mit dem „Abkommen über die Hilfeleistung der Union der<br />

Sozialistischen Sowjetrepubliken an die Deutsche Demokratische Republik bei der Entwicklung der Forschung auf dem<br />

Gebiet der Physik des Atomkerns und der Nutzung der Kernenergie für die Bedürfnisse der Volkswirtschaft“, das am<br />

28. April 1955 abgeschlossen wurde. Auf dieser Grundlage erfolgte die Lieferung zweier Großgeräte, ein 2 MW<br />

Forschungsreaktor vom Typ WWR-S (Abb. 1) und ein 25 MeV-Zyklotron, an das am 1. Januar 1956 gegründete<br />

Zentral institut für Kernphysik (später Zentralinstitut für Kern<strong>for</strong>schung).<br />

Am 14. Dezember 1957 wurde der<br />

erste Forschungsreaktor der DDR<br />

nach einer Bauzeit von nur 21<br />

Monaten erstmals kritisch und am<br />

16. Dezember 1957 offiziell im Rahmen<br />

eines Staatsaktes in Betrieb<br />

genommen. Der RFR war ein leichtwassermoderierter<br />

und -gekühlter<br />

Tankreaktor. Über einen Zwischenschritt<br />

(von 2 MW auf 4 MW und<br />

5 MW im Jahre 1965) erfolgte zuletzt<br />

eine Leistungs erhöhung auf 10 MW<br />

(1967), die ab 1981 im Dauerbetrieb<br />

realisiert wurde. Die Leistungserhöhung<br />

auf 10 MW konnte nur durch<br />

Einsatz neuer Brennstäbe mit einer<br />

Anreicherung von 36 % U-235 gegenüber<br />

vorher von 10 % U-235 erreicht<br />

werden. Die Energieabgabe des RFR<br />

betrug insgesamt rund 28.000 MWd<br />

bei einem Leistungseinsatz von rund<br />

105.000 Stunden. Es gab während der<br />

Betriebszeit des RFR kein Ereignis,<br />

welches strahlenschutzrelevant war.<br />

Interessante Details über die RFR-<br />

Betriebszeit enthält das Buch „Beiträge<br />

zur Geschichte des Rossendorfer<br />

Forschungsreaktors RFR“ [1].<br />

Im Zuge der weiteren Entwicklung<br />

des Forschungsstandortes entstanden<br />

eine Vielzahl von weiteren Anlagen<br />

und Einrichtungen in denen eine kerntechnische<br />

Nutzung bis 1991 erfolgte,<br />

wobei hier der RFR nur bis zum<br />

27. Juni 1991 betrieben wurde. Mit<br />

der Neuordnung des Forschungsstandortes<br />

nach der Wiedervereinigung und<br />

der Auflösung der Akademie der<br />

Wissenschaften zum 31. Dezember<br />

1991 wurden dann auf der Grundlage<br />

mehrerer in den Jahren 1993 und<br />

1996 vollzogener Kabinettsbeschlüsse<br />

mit der Stilllegung und dem Rückbau<br />

der kerntechnischen Anlagen am<br />

Forschungsstandort begonnen. Der<br />

Freistaat Sachsen beauftragte den<br />

Verein für Kern verfahrenstechnik und<br />

Analytik Rossendorf e. V. (VKTA, Umbenennung<br />

in „VKTA – Strahlenschutz,<br />

Analytik & Entsorgung Rossendorf<br />

e. V.” im Dezember 2014) mit der vollständigen<br />

Beseitigung der nuklearen<br />

Altlasten des Forschungsstandortes<br />

Rossendorf. Im Folgenden werden die<br />

wichtigsten Rückbauobjekte mit<br />

Termin der Ent lassung (in Klammern)<br />

aus dem Geltungsbereich des Atomgesetzes<br />

aufgeführt:<br />

p die Rossendorfer Anordnung für<br />

kritische Experimente (RAKE,<br />

1998)<br />

p das Urantechnikum (2000)<br />

p der Rossendorfer Ringzonenreaktor<br />

(RRR, 2000)<br />

p die Spezielle Kanalisation am<br />

Forschungsstandort (AtG-Teilentlassungen<br />

2010, 2013 und 2018)<br />

p die Objekte und Anlagen der Isotopenproduktion<br />

(von 2009 bis<br />

2014, [2])<br />

p die Anlagen und das Gelände der<br />

Entsorgungs wirtschaft des ehemaligen<br />

ZfK (2005, 20<strong>11</strong> und 2018)<br />

p der RFR (2009, 2010, 2019)<br />

| Abb. 1.<br />

Ansicht des Forschungsreaktors während der Fertigstellungsphase.<br />

Die Kosten für Stilllegung, Rückbau<br />

und Entsorgung werden vollständig<br />

vom Freistaat Sachsen getragen,<br />

sodass die Mittelbereitstellung auch<br />

Einfluss auf die terminlichen Arbeitsabläufe<br />

hatte. So mussten in den<br />

Jahren 2005 und 2006 die Rückbauarbeiten<br />

aufgrund der zu knapp bemessenen<br />

Finanzmittel durch den<br />

Freistaat Sachsen eingestellt werden,<br />

was zu entsprechend längeren Rückbauzeiten<br />

führte, da z. B. Mitarbeiter<br />

nicht mehr zur Verfügung standen.<br />

2 Objektbeschreibung<br />

Zu den RFR-Anlagen (Abb. 2) gehörten<br />

der „Rückbau komplex RFR“<br />

mit einer Fläche von ca. 9.000 m 2 und<br />

den Objekten:<br />

p Labortrakt (1) mit Reaktorwarte<br />

(2)<br />

537<br />

DECOMMISSIONING AND WASTE MANAGEMENT<br />

Decommissioning and Waste Management<br />

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

DECOMMISSIONING AND WASTE MANAGEMENT 538<br />

| Abb. 2.<br />

Territoriale Lage der RFR-Anlagen.<br />

p Reaktorhalle mit Vorhalle und<br />

Anbau (3)<br />

p Pavillon und einen weiteren<br />

Anbau (4)<br />

p Ventilations- und Filtergebäude<br />

mit Fortluftschornstein (5)<br />

p Schauer (6)<br />

sowie die außerhalb des Komplexes<br />

liegenden Objekte:<br />

p Trafostation (7)<br />

p Notstromgebäude (8)<br />

p altes Armaturenhaus (9)<br />

p Armaturenhaus (10)<br />

p Trockenkühlturm 1 und 2 (<strong>11</strong>)<br />

p Rohrleitungen des 2. Kühlkreislaufes<br />

(12)<br />

Weiterhin gab es außerhalb noch die<br />

Sammelbehälter anlage mit der Pumpbedienstation,<br />

in der die radioaktiven<br />

Abwässer des RFR aufgefangen<br />

wurden. Diese Anlagen befanden sich<br />

im Kontrollbereich der Entsorgungswirtschaft<br />

und wurden im Rahmen<br />

dieses Rückbaukomplexes aus der<br />

atomrechtlichen Aufsicht entlassen.<br />

Die territoriale Lage der nummerierten<br />

RFR-Anlagen ist schematisch in der<br />

Abb. 2 dargestellt. Im Betriebshof befanden<br />

sich befestigte Verkehrsflächen,<br />

Rohre der Regen- und Schmutzwasserkanalisation<br />

sowie Leitungen für<br />

konta minationsverdächtige Abwässer<br />

bis zu einer Tiefe von 5 m, unterirdische<br />

Abluftleitungen, Heizungskanäle,<br />

Schächte sowie Medienleitungen.<br />

Im Folgenden werden wichtige<br />

Objekte kurz charakterisiert:<br />

Labortrakt mit Reaktorwarte<br />

(1 und 2)<br />

Der Labortrakt war ein viergeschossiger<br />

Ziegelbau mit Stahlbetondecken,<br />

dessen Grundfläche ca. 960 m 2<br />

betrug. In diesem Gebäude waren<br />

ursprünglich Laboratorien, die Reaktorwarte<br />

und Büroräume untergebracht.<br />

Im Zuge der Erneuerung<br />

des RFR wurde eine neue Warte in<br />

Stahlbeton- Skelettbauweise mit einer<br />

Fläche von ca. 260 m 2 an der NO-Seite<br />

des Labortraktes angebaut.<br />

Reaktorhalle (3)<br />

Die Reaktorhalle war ein unterkellerter<br />

Ziegel-Stahl- Skelettbau mit<br />

einer Gesamthöhe von 24,5 m und<br />

einer Grundfläche von ca. 700 m 2 .<br />

Darin befanden sich neben dem<br />

Reaktor unter anderem Brennelemente-Lagerbecken<br />

(AB 1 und AB 2)<br />

sowie im Keller geschoss vier Heiße<br />

Kammern.<br />

Pavillon (4)<br />

Der Gebäudebereich „Pavillon“ mit<br />

einer Grundfläche von ca. 170 m 2<br />

bildete einen stark gegliederten einbzw.<br />

zweigeschossigen Anbau, unter<br />

dem Abluftkanäle zum Filter- und<br />

Ventilationsgebäude sowie Rohrleitungen<br />

für kontaminierte Abwässer<br />

verliefen. Im Gebäudebereich waren<br />

ein radiochemisches Labor und eine<br />

Gamma- Bestrahlungsanlage untergebracht.<br />

Ventilations- und Filtergebäude (5)<br />

Das Filter- und Ventilationsgebäude<br />

mit einer Grundfläche von ca. 350 m 2<br />

war ein anderthalbgeschossiger<br />

Ziegel bau mit einem Flachdach, auf<br />

dem sich der ca. 33 m hohe Fortluftschornstein<br />

befand. Im Inneren<br />

waren die Venti lationskammern mit<br />

den dazugehörigen Filteranlagen.<br />

Altes Armaturenhaus (9)<br />

Das Gebäude mit einer Grundfläche<br />

von ca. 22,7 m x 7,2 m und einer Höhe<br />

von ca. 4,5 m hatte über zwei Drittel<br />

der Grundfläche eine ca. 3,5 m tiefe<br />

Grube mit dem Pumpenfundament.<br />

Das Gebäude bestand aus Mauerwerk;<br />

die Fundamente und Teile der<br />

aufgehenden Wände aus Stampfbeton.<br />

Armaturenhaus (10)<br />

Das Armaturenhaus mit einer<br />

Abmessung von ca. 18,4 m x 12,4 m x<br />

6,0 m wurde im Rahmen der Gesamtrekonstruktion<br />

RFR im Jahr 1985 neu<br />

errichtet und beinhaltete Pumpen,<br />

Armaturen und eine Reinigungs anlage<br />

für das Sekundärkühlsystem. Es<br />

handelte sich um eine Stahl betonskelettkonstruktion<br />

aus Fertigteilen,<br />

| Abb. 3.<br />

Schnittdarstellung RFR-Gebäudekomplex.<br />

Decommissioning and Waste Management<br />

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

wobei die Außenwände aus Gasbetonfertigteilen<br />

bestanden.<br />

Trockenkühltürme (<strong>11</strong>)<br />

Die Trockentürme hatten eine Abmessung<br />

von ca. 23,3 m x 13,2 m x 8,0 m.<br />

Sie bestanden aus einem Traggerüst<br />

aus Stahl, das mit Leichtmetallelementen<br />

verkleidet war.<br />

Die umbaute Fläche des RFR ohne<br />

Filter- und Venti lationsgebäude betrug<br />

ca. 2.100 m 2 , wobei der Kontrollbereich<br />

die Reaktorhalle mit Kellergeschoss,<br />

Teile des Kellergeschosses des<br />

Anbaus der Reaktorhalle und Teile des<br />

Kellergeschosses des Labortraktes<br />

umfasste. Die Schnittdarstellung in<br />

der Abb. 3 vermittelt einen Eindruck<br />

von der baulichen Anordnung.<br />

Neben dem Reaktor (Detaildarstellung<br />

nach Re konstruktion in Abb. 4)<br />

waren die vier Heißen Zellen, das<br />

Lagerbecken AB 2 und der Pumpenraum<br />

(vgl. Abb. 5) hinsichtlich des<br />

Rückbaus schwierige Teilobjekte.<br />

Aus der Betriebshistorie waren für<br />

die ersten Schritte des Rückbaus und<br />

der Entsorgung vor allem zu berücksichtigen:<br />

p betriebliche und strahlenschutzrelevante<br />

Ereignisse<br />

p der Einsatz von Brennelementen<br />

mit einem hohen Anreicherungsgrad<br />

p die nahezu vollständige Erneuerung<br />

des RFR in den Jahren 1986<br />

bis 1989 mit der Wiederinbetriebnahme<br />

am 27. Januar 1990 sowie<br />

die endgültige Einstellung des<br />

nuklearen Betriebes des RFR am<br />

27. Juni 1991 bedingt durch eine<br />

befristete Genehmigung zum<br />

Versuchsbetrieb.<br />

3 Genehmigungsverfahren<br />

Im Jahre 1991 wurden Aufsichtliche<br />

Anordnungen gemäß § 19 Absatz<br />

3 AtG durch das Sächsische<br />

Staatsministerium für Umwelt und<br />

Landesentwicklung (SMU) erlassen,<br />

um bedingt durch die Auflösung der<br />

Institute der Akademie der Wissenschaften<br />

der DDR zum 31. Dezember<br />

1991 letztendlich einen ungeregelten<br />

Zustand für den RFR zu vermeiden<br />

sowie am 19. Dezember 1991 den<br />

Betreiberwechsel zum VKTA vorzunehmen.<br />

1993 fasste die Sächsische<br />

Staatsregierung den Kabinettsbeschluss,<br />

den Forschungsreaktor endgültig<br />

stillzulegen und zurück zubauen.<br />

Um das Genehmigungsverfahren<br />

nach § 7 Absatz 3 AtG einzuleiten,<br />

stellte der VKTA bereits im Dezember<br />

1994 beim Sächsischen Staatsministerium<br />

für Umwelt und Landesentwicklung<br />

einen Antrag auf Genehmigung<br />

zur Stilllegung und zum Abbau<br />

| Abb. 4.<br />

Schnittdarstellung RFR.<br />

des RFR. Von 1993 bis 1998 wurden<br />

technische, sicherheitstech nische und<br />

strahlenschutztechnische Maßnahmen<br />

zur Anpassung an den bundesdeutschen<br />

Standard durch geführt.<br />

Hervorzuheben ist die Erstellung<br />

eines brandschutztechnischen Gutachtens<br />

durch eine Fremdfirma, auf<br />

deren Grundlage bautechnische Maßnahmen,<br />

die Reduzierung der Brandlast<br />

sowie die Inbetriebnahme einer<br />

neuen Brandschutzanlage erfolgten.<br />

Weiterhin mussten am Forschungsstandort<br />

Rossendorf erst die Voraussetzungen<br />

geschaffen werden, um die<br />

Kernbrennstoffe bzw. Kernmaterialien<br />

verwahren, die radioaktiven Abfälle<br />

und Reststoffe zwischenlagern,<br />

behandeln, analysieren und freigeben<br />

zu können. Dies er<strong>for</strong>derte zahlreiche<br />

atom- und strahlenschutzrechtliche<br />

Genehmigungen sowie deren bauliche<br />

Umsetzung, die in den Jahren<br />

1992 bis 1999 ebenfalls stattfanden.<br />

Hervorzuheben sind hierbei die Einrichtung<br />

zur Entsorgung von Kernmaterial,<br />

die Transportbereitstellungshalle<br />

für die CASTOREN, das<br />

Zwischenlager, die Reststoffbehandlungsanlage,<br />

das Freimesszentrum,<br />

das Analytiklabor sowie die notwendigen<br />

Einrichtungen für den<br />

Strahlenschutz, wie die Inkorporationsmessstellen<br />

sowie für die<br />

| Abb. 5.<br />

Schnittdarstellung RFR.<br />

Umgebungsüberwachung. Nach Erhalt<br />

der ersten beiden Genehmigungen<br />

zur RFR-Stilllegung [3, 4] wurden<br />

von 1998 bis 2001 die Anlagen kernbrennstoff-<br />

und kernmaterialfrei [5,<br />

6, 7] gefahren und für den VKTA mit<br />

dem Rückbau des 2. Kühlwasserkreislaufes<br />

der Abschluss der Stilllegungsphase<br />

erreicht. Von 2001 bis 2018<br />

erfolgten dann der Rückbau sowie die<br />

Geländesanierung mit abschließender<br />

Flächenprofilierung im Rahmen<br />

DECOMMISSIONING AND WASTE MANAGEMENT 539<br />

Decommissioning and Waste Management<br />

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

DECOMMISSIONING AND WASTE MANAGEMENT 540<br />

der Dritten und Vierten RFR-<br />

Stilllegungsgenehmigungen.<br />

Die Grundlage für die Stilllegung<br />

und den Rückbau des RFR waren insgesamt<br />

vier Genehmigungen mit ihren<br />

Änderungen. Die Erste Genehmigung<br />

45-4653.18 VKTA 01 zur Stilllegung<br />

des Rossendorfer Forschungsreaktors<br />

RFR [3] wurde am 30. Januar 1998<br />

durch das SMU erteilt. Deren 1. Änderungsgenehmigung<br />

45-1653.18 VKTA<br />

01/1, erteilt vom Sächsischen Staatsministerium<br />

für Umwelt und Landwirtschaft<br />

(SMUL) am 06. November<br />

2000, bein hal tete vor allem das Innehaben<br />

der endgültig abgeschalteten<br />

Anlage, die sichere Betriebsführung<br />

der abgeschalteten Anlage zum Zwecke<br />

der Stilllegung, die Überführung<br />

von Brennelementen aus der Spaltzone<br />

in das Brennelement lagerbecken<br />

AB 2 und den innerbetrieblichen<br />

Transport.<br />

Mit der Zweiten Genehmigung<br />

45-4653.18 VKTA 02 des SMUL vom<br />

30. Oktober 1998 zur Stilllegung des<br />

Rossendorfer Forschungsreaktors<br />

RFR [4] (eingereicht im Oktober<br />

1997) konnte der Rückbau des 2.<br />

Kühlkreislaufes realisiert werden,<br />

deren 1. Änderung vom <strong>11</strong>. Februar<br />

beinhaltete die Erweiterung des<br />

Genehmigungsumfanges bzgl. eines<br />

Raumes im RFR-Labortrakt.<br />

| Abb. 6.<br />

Überblick zum Aktivitätsinventar.<br />

Die Dritte Genehmigung 4653.18<br />

VKTA 03 zur Still legung und zum<br />

Abbau des Rossendorfer Forschungsreaktors<br />

RFR des SMUL vom 3. April<br />

2001 [8] ermög lichte die Entsorgung<br />

der Betriebsmedien sowie die Außerbetriebnahme<br />

und den Rückbau der<br />

nicht mehr benötigten Systeme und<br />

Komponenten des RFR. Dazu zählte<br />

z. B. der Rückbau des 1. Kühlkreislaufes,<br />

des Speisewassersystems und<br />

des Reaktorbehälters. Insgesamt gab<br />

es 14 Vorhaben, wobei mit Erteilung<br />

der Genehmigung bereits für vier<br />

Vorhaben die Zustimmungen des<br />

SMUL vorlagen. Die restlichen Vorhaben<br />

wurden abbaubegleitend mit<br />

dem SMUL abgestimmt.<br />

Mit der Vierten Genehmigung<br />

4653.18 VKTA 04 des SMUL vom<br />

1. Februar 2005 [9] konnte schließlich<br />

der Abbau der Restanlage des RFR vorgenommen<br />

werden. Der Änderungsbescheid<br />

4653.18 VKTA 04/1 vom<br />

9. November 2010 beinhaltete die Freigabe<br />

und Ent lassung des Raumes 01 im<br />

Gebäude 103 (Notstrom gebäude) und<br />

mit der 2. Änderungsgenehmigung<br />

4653.18 VKTA 04/2 des SMUL vom<br />

9. Januar 2014 wurden die Änderungen<br />

des räumlichen Geltungsbereiches<br />

sowie des Genehmigungsumfanges<br />

beschieden. Hintergrund war die Entscheidung<br />

des VKTA aufgrund von<br />

radiologischen Voruntersuchungen sowie<br />

durch technologisch bedingte<br />

Änderungen der Planungen des Abbaus,<br />

einen Total abbruch der RFR-<br />

Restanlage unter Strahlenschutzbedingungen<br />

vorzunehmen, wobei<br />

hierfür die notwen digen Erläuterungsberichte<br />

benötigt wurden wie ebenfalls<br />

für die Baufreiheit die Erweiterung des<br />

räumlichen Geltungsbereiches.<br />

Zur jeweiligen Genehmigungsplanung<br />

erfolgte die grundlegende<br />

Beschreibung des Gesamtvorhabens<br />

durch Erläuterungsberichte. Im<br />

Rahmen der Vierten Genehmigung<br />

gab es beispielsweise 18 Vorhaben,<br />

wobei das Vor haben <strong>11</strong> und das Vorhaben<br />

15 jeweils in drei Teile<br />

gegliedert war. Es wurden folglich<br />

22 Erläuterungs berichte erstellt, von<br />

denen sieben bereits bei Erteilung der<br />

Genehmigung von der zuständigen<br />

Behörde bestätigt wurden. Die weiteren<br />

Erläuterungsberichte erhielten im<br />

Zuge des Aufsichtsverfahrens die<br />

Zustimmung. Nach Abschluss eines<br />

Vorhabens wurde ein Abschlussbericht<br />

erstellt und an die zuständige<br />

Behörde übergeben.<br />

4 Ausgangssituation<br />

(radiologisch,<br />

konventionell)<br />

Die radiologische Ausgangssituation<br />

war geprägt durch das Vorhandensein<br />

von be- und unbestrahltem Kernbrennstoff,<br />

Kernmaterial sowie von<br />

Aktivierung und Kontamination<br />

unterschiedlichster Stoffe, die lokal<br />

und aktivitäts mäßig sehr differenziert<br />

waren. Einen Grobüberblick zum<br />

Aktivitätsinventar gibt die Abb. 6.<br />

Die Herstellung der Kernbrennstoff-<br />

Freiheit der RFR- Anlage war ein<br />

wichtiger Schritt für den Start des<br />

Rückbaus. Dazu mussten sowohl die<br />

951 bestrahlten Brenn elemente mit<br />

einer Gesamtaktivität von 8,91 E+15<br />

Bq als auch einige Posten kernbrennstoffhaltige<br />

Abfälle aus der RFR-<br />

Anlage entfernt werden. Diese Vorhaben<br />

wurden im Februar 2001 abgeschlossen.<br />

Weitere Einzelheiten sind<br />

im Abschnitt 7.2 zu entnehmen (s. <strong>atw</strong><br />

1/2020).<br />

Hinsichtlich der Aktivierung waren<br />

alle reaktornahen Baustrukturen<br />

und Anlagenteile, wie z. B. die Inneneinbauten<br />

des Reaktors, die Bestrahlungskanäle,<br />

das Biologische Schild,<br />

die Thermische Säule zu beachten.<br />

Wichtige Aktivierungsnuklide waren<br />

beispielsweise Eu-152, Ba-133, H-3,<br />

Co-60 (Beton/Bauschutt), Fe-55,<br />

Co-60, Ni-63, H-3 (Eisenteile), Fe-55,<br />

Co-60, Ni-63, H-3 (Stahlguss-Reaktordeckel),<br />

H-3, Ni-63, Co-60, Be-10<br />

Decommissioning and Waste Management<br />

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

(Be-Reflektorelemente), Ni-63, Ni-59,<br />

H-3, C-14, Co-60 (Borcarbid-<br />

Absorberstäbe), C-14, H-3, Eu-152,<br />

Co-60 (Graphit). Unerwartet wurde<br />

bei der radiologischen Voruntersuchung<br />

festgestellt, dass anstatt<br />

des in den auunterlagen ausgewiesenen<br />

Barytbetons das Biologische<br />

Schild aus Beton mit Eisenteilen<br />

bestand.<br />

Kontaminationen resultierten vor<br />

allem aus flüssigkeitsgetragenem<br />

Transport nach Freisetzung sowie aus<br />

einem Am-241-Ereignis (Freisetzung<br />

aus einem Target, 1969). Die Hauptnuklide<br />

in der RFR-Anlage waren<br />

H-3, Co-60, Sr-90+, Cs-137, Eu-152,<br />

U- und Pu-Nuklide und Am-241. Zur<br />

Erfassung der jeweiligen Ausgangssituation<br />

wurde nach der historischen<br />

Erkundung die radiologische Voruntersuchung<br />

nach Möglichkeit mit<br />

konventionellem Schadstoffuntersuchungen<br />

gekoppelt. Dies reduzierte<br />

später den zeitlichen und finanziellen<br />

Aufwand und so konnten z. T. potentielle<br />

Verdachtsflächen besser beurteilt<br />

und die Entsorgung schneller<br />

voran gebracht werden.<br />

Das Auftreten von chemotoxischen<br />

(konventionellen) Schadstoffen resultiert<br />

sowohl aus Betriebsabläufen<br />

als auch im Wesentlichen aus schädlichen<br />

Bestandteilen der Baustoffe.<br />

Zu den Gebäudeschadstoffen zählen<br />

beispielsweise Polyzyklische Aromatische<br />

Kohlenwasserstoffe (PAK,<br />

insbesondere in Teeranstrichen),<br />

Polychlorierte Biphenyle (PCB, Oberflächenbeschichtungen),<br />

künstliche<br />

Mineralfasern (KMF) und Asbest (in<br />

Dämmstoffen). In Ausrüstungen<br />

und Anlagen fand man z. B. PCB<br />

(Trans<strong>for</strong>matoren, Kondensatoren),<br />

Mineral kohlenwasserstoffe (MKW,<br />

Pumpen) oder Asbest (Dichtungen).<br />

Aus Betriebsabläufen stammen u. a.<br />

Kontaminationen von Schwermetallen<br />

und MKW. Das frühzeitige<br />

Erkennen der Schadstoffsituation<br />

ermöglichte die planerische Einarbeitung<br />

in die Rückbauprozesse<br />

sowie vor allem einen zeitlichen Vorlauf<br />

für eine sachgerechte Entsorgung<br />

des betreffenden Materials zu erhalten.<br />

Zu der Schadstoffproblematik<br />

veröffentlichte der VKTA einige<br />

Beiträge [10–12]. Bezüglich des RFR-<br />

Rückbaus sind insbesondere festgestellte<br />

PAK-Kontaminationen im Beton<br />

und Erdreich, das Vorfinden von<br />

Asbest hinter den Stahlbecken im<br />

Lager becken AB 2 (nicht in Bauzeichnung<br />

erwähnt) sowie lokale Kontaminationen<br />

mit MKW, Quecksilber<br />

und Schwermetallen im Beton hervorzuheben.<br />

5 Planungskonzept<br />

Grundsatz für die Planung war zum<br />

einen das Ziel, bei Stilllegung und<br />

Rückbau die Entsorgung von freigegebenen<br />

Stoffen parallel vorzunehmen<br />

und die radio aktiven Abfälle im<br />

1999 errichteten Zwischenlager<br />

Rossendorf, welches 2000 erweitert<br />

wurde, ordnungs gemäß für die spätere<br />

Endlagerung zu lagern. Zum<br />

anderen sollte das Betriebspersonal<br />

des RFR so weit wie möglich eingebunden<br />

werden und der VKTA nicht<br />

nur die Planungshoheit innehaben,<br />

sondern auch möglichst viele kostengünstige<br />

Beiträge zur Aufgabenerfüllung<br />

leisten. Da der VKTA planungstechnisch<br />

nicht alles abdecken konnte,<br />

wurde ab 2004 mit der heutigen<br />

Siempelkamp NIS Ingenieurgesellschaft<br />

mbH (NIS) eine Fachfirma eingebunden.<br />

NIS war, unterstützt durch<br />

Fachplaner in den Gebieten Lüftung,<br />

Tragwerksplanung und Schadstofferkundung,<br />

mit der Planung, der<br />

Erstellung und Bewertung von Ausschreibungsunterlagen<br />

sowie der<br />

Bau überwachung beauftragt. Ebenso<br />

wurden für spezielle Aufgaben weitere<br />

Fachfirmen gebunden. Der VKTA<br />

selbst, zeitweilig unterstützt durch<br />

einen Projektsteuerer, stellte die<br />

Rückbauleitung, bestehend aus:<br />

p Rückbauleiter<br />

p Gebäudeverantwortlicher,<br />

Strahlen schutzbeauftragter<br />

p Strahlenschutzingenieur/in<br />

p Strahlenschutzfachkraft<br />

p technischem Personal für strahlenschutztechnische<br />

Messungen,<br />

Transporte und kleineren technischen<br />

Aufgaben<br />

Die personelle Absicherung der Rückbauleitung<br />

in der Rückbau-Etappe bis<br />

2007 wurde dabei ausschließlich<br />

durch das ehemalige Betriebspersonal<br />

des RFR sicher gestellt. Die ab 2006<br />

auszuführenden Arbeiten wurden<br />

strukturiert, geplant, ausgeschrieben<br />

und an externe Dienstleister im<br />

Rahmen von Verträgen nach VOL<br />

oder VOB vergeben. Hierzu wurde<br />

eine Struktur verwendet, die sich<br />

einheitlich in Arbeits-, Termin- und<br />

Finanzpläne gliederte mit den<br />

Baulosgruppen:<br />

p Vorbereitende Maßnahmen<br />

p Bereitstellung von Ausrüstungen<br />

p Dienstleistungen Abbau<br />

p Abbaubegleitende lufttechnische<br />

Maßnahmen<br />

p Dienstleistungen Dekontamination<br />

p Arbeitsbegleitender Strahlenschutz<br />

p Mess- und Freimessprogramme in<br />

Planung und Durchführung<br />

p Abbaubegleitende sonstige<br />

Maßnahmen<br />

p Abbrucharbeiten bis zur Geländeprofilierung<br />

Die Leistungen des arbeitsbegleitenden<br />

Strahlenschutzes, der Planung<br />

und Durchführung von radiolo gischen<br />

und konventionellen Messungen/Analysen<br />

und von (Frei)Mess programmen<br />

übernahm der VKTA weitestgehend<br />

selbst. Dazu nutzte der VKTA u. a. sein<br />

nach DIN EN ISO/IEC 17025 akkreditiertes<br />

Labor für Umwelt- und Radionuklidanalytik.<br />

Andere auszuführende<br />

Arbeiten wurden in rund 200 Losen<br />

i. d. R. ausgeschrieben und den Losgruppen<br />

zugeordnet. Die bewusst<br />

kleinteilige Vergabe von Losen erlaubte<br />

es dem VKTA, mit seinen Planern<br />

auf besondere und erst im Verlauf des<br />

Abbaus erkennbare neue Situationen<br />

zügig zu reagieren, ortsnahe Firmen<br />

einzubinden und letztlich Kosten zu<br />

sparen.<br />

Der Arbeitsumfang des Gesamtprojektes<br />

entsprechend der jeweiligen<br />

Genehmigung wurde, wie bereits<br />

erwähnt, in Einzelvorhaben gegliedert,<br />

die in Erläuterungsberichten<br />

der Genehmigungsbehörde dargelegt<br />

wurden. Die zeit liche Abarbeitung der<br />

einzelnen Vorhaben erfolgte entsprechend<br />

eines Rahmenplanes, der<br />

mehrmals den objektiven Umständen<br />

angepasst werden musste. Für jedes<br />

Einzelvorhaben lag ein Detailplan vor.<br />

Die Durch führung der Arbeiten erfolgte<br />

auf der Grundlage von Arbeitsanweisungen<br />

und Ablaufplänen,<br />

wobei u. a. vor jedem Vorhaben ein<br />

Rückbauerlaubnis- sowie ein Arbeitserlaubnisverfahren<br />

gemäß der entsprechenden<br />

VKTA- Regelung durchzuführen<br />

war.<br />

6 Meilensteine Stilllegung<br />

und Rückbau<br />

Nach dem Erhalt der jeweiligen<br />

Genehmigung konnten die geplanten<br />

Arbeiten durchgeführt werden, wobei<br />

nach folgend wichtige Meilensteine<br />

des Rückbau<strong>for</strong>tschrittes beim RFR<br />

chronologisch (Durchführungszeitraum<br />

in Klammern) aufgeführt<br />

werden:<br />

p Kabinettsbeschluss des Freistaates<br />

Sachsen zur „Endgültige Stilllegung“<br />

(13. Juli 1993)<br />

p Antrag auf Stilllegung<br />

(21. Dezember 1994), Erhalt der<br />

Ersten Genehmigung (30. Januar<br />

1998, vgl. Abschnitt 3) und somit<br />

ab 2. Februar 1998 RFR-Betriebsführung<br />

nicht mehr auf der<br />

Grund lage der Aufsichtlichen<br />

Anord nungen des SMU, u. a.<br />

vom 28. Juni 1991<br />

p Herstellen der Kernbrennstofffreiheit<br />

des Reaktor behälters<br />

DECOMMISSIONING AND WASTE MANAGEMENT 541<br />

Decommissioning and Waste Management<br />

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

DECOMMISSIONING AND WASTE MANAGEMENT 542<br />

durch Umlagerung der Brennelemente<br />

aus dem Reaktorkern ins<br />

Lagerbecken AB 2 und dem Ausbau<br />

der kernbrennstoffhaltigen<br />

Neutro nen detektoren (6. April<br />

1998)<br />

p Antrag auf Zweite Stilllegungsgenehmigung<br />

(10. Oktober 1997);<br />

Erhalt der Zweiten Genehmigung<br />

(30. Oktober 1998) mit<br />

1. Änderung (<strong>11</strong>. Februar 1999)<br />

p Rückbau des 2. Kühlkreislaufes<br />

mit Entkernung und Abriss des<br />

Armaturenhauses, der Trockenkühltürme,<br />

des Pumpenhauses<br />

sowie der Verbindungsleitungen<br />

zwischen den Gebäuden und dem<br />

Reaktorgebäude; mit SMUL-<br />

Schreiben vom 16. August 1999<br />

waren diese Anlagen nicht mehr<br />

Bestandteil der RFR-Anlage<br />

p Überführung der Brennelemente<br />

aus dem Lagerbecken AB 2 in<br />

CASTOR® MTR 2-Behälter, Endabfertigung<br />

mit anschließendem<br />

Transport in die Transportbereitstellunghalle<br />

des VKTA<br />

(24. November 2000)<br />

p RFR-Anlage kernbrennstofffrei<br />

nach Abgabe/Konditionierung der<br />

restlichen Posten Kernbrennstoffabfälle<br />

(26. Februar 2001)<br />

p Antrag auf Dritte Stilllegungsgenehmigung<br />

(29. Dezember<br />

1998); Erhalt der Dritten<br />

Genehmigung (3. April 2001)<br />

p Ausbau des Reaktorbehälters,<br />

Arbeiten zur Transportbereitstellung<br />

und Transport zum<br />

Konditionierer (2001/2002)<br />

p Antrag auf Vierte Stilllegungsgenehmigung<br />

(31. Juli 2003);<br />

Erhalt der Vierten Genehmigung<br />

(1. Februar 2005) mit 2. Änderungsgenehmigung<br />

(9. Januar<br />

2014)<br />

p Transport der RFR-Brennelemente<br />

mittels CASTOREN ins Zwischenlager<br />

Ahaus (3 Konvoitransporte<br />

vom 30. Mai 2005 bis 13. Juni<br />

2005)<br />

p Abbau des RFR-Baukörpers einschließlich<br />

der Aus kleidung des<br />

Lagerbeckens AB 1 (2008/2009)<br />

p Ausräumen, Dekontamination und<br />

Abbruch der Heißen Zellen<br />

(2009/2010)<br />

p Ausbau des Deaerators<br />

(2010/20<strong>11</strong>)<br />

p Ausräumen, Dekontamination und<br />

Entkernung des Kellergeschosses<br />

(20<strong>11</strong>/2012)<br />

p Abbau, Dekontamination und<br />

Entsorgung der Teile des<br />

frei gegebenen Fortluftschornsteines<br />

(2013/2014)<br />

p Dekontamination, Entkernung<br />

und Abriss des Filter- und<br />

Ven tilationshauses (2014/2015)<br />

p Dekontamination, Entkernung<br />

und Abriss des Labortraktes mit<br />

Reaktorhalle und -warte (2016)<br />

p Abschluss des Ausbaus aller<br />

tiefliegender Bau strukturen und<br />

Bodensanierung (2016)<br />

p Abschluss der Baugrubenverfüllung<br />

sowie Beendigung der<br />

restlichen bautechnischen<br />

Maß nahmen (Profi lierung,<br />

Oberflächengestaltung<br />

(2017/2018)<br />

Autoren<br />

Reinhard Knappik<br />

Klaus Geyer<br />

Sven Jansen<br />

Cornelia Graetz<br />

VKTA Rossendorf<br />

Bautzner Landstraße 400<br />

01328 Dresden<br />

Deutschland<br />

Imprint<br />

| Editorial Advisory Board<br />

Frank Apel<br />

Erik Baumann<br />

Dr. Erwin Fischer<br />

Carsten George<br />

Eckehard Göring<br />

Dr. Florian Gremme<br />

Dr. Ralf Güldner<br />

Carsten Haferkamp<br />

Christian Jurianz<br />

Dr. Anton Kastenmüller<br />

Prof. Dr. Marco K. Koch<br />

Ulf Kutscher<br />

Herbert Lenz<br />

Jan-Christan Lewitz<br />

Andreas Loeb<br />

Dr. Thomas Mull<br />

Dr. Joachim Ohnemus<br />

Olaf Oldiges<br />

Dr. Tatiana Salnikova<br />

Dr. Andreas Schaffrath<br />

Dr. Jens Schröder<br />

Norbert Schröder<br />

Prof. Dr. Jörg Starflinger<br />

Dr. Brigitte Trolldenier<br />

Dr. Walter Tromm<br />

Dr. Hans-Georg Willschütz<br />

Dr. Hannes Wimmer<br />

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ISSN 1431-5254<br />

Decommissioning and Waste Management<br />

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

First On-site Demonstration of Laser- based<br />

Decontamination Technology in Germany<br />

Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann<br />

and Antonio Hurtado<br />

1 Introduction Several ongoing dismantling and decommissioning (D&D) projects of nuclear facilities in<br />

Germany produce wide experiences in terms of the regulation framework and the practical implementation. The D&D<br />

of commercial and research reactors require a comprehensive adaption of the available technologies to meet the<br />

demands of the associated health and safety requirements. This causes a rising monetary and human ef<strong>for</strong>t.<br />

To face this challenge the German<br />

federal government focuses the<br />

research and development activities<br />

to:<br />

a- protect human beings and the<br />

environment during nuclear<br />

decommissioning,<br />

b- educate staff and young academics<br />

and<br />

c- develop and optimize decommission<br />

methods.<br />

The dismantling and decommissioning<br />

of the various nuclear facilities<br />

located at Karlsruhe Institute of<br />

Technology (KIT) Campus North, the<br />

Kerntechnische Entsorgung Karlsruhe<br />

GmbH (KTE) faces several technical<br />

challenges.<br />

On the one hand, metallic surfaces,<br />

e. g. in the Hot Cells Facility (HZ),<br />

must be decontaminated radiologically,<br />

on the other hand, paint layers<br />

as well as radiological contamination<br />

must be removed from metallic components<br />

during post treatment, e. g.<br />

in the Waste Treatment Department<br />

(EB).<br />

Furthermore, large surfaces of<br />

concrete structures, e. g. in the<br />

Reprocessing Building (PG) of the<br />

Karlsruhe Reprocessing Plant (WAK),<br />

in the Multi-Purpose Research Reactor<br />

(MZFR) and in the Research Reactor<br />

(FR-2) must be cleared from paint<br />

layers, containing sometimes high rate<br />

of polychlorinated biphenyls (PCB).<br />

The overriding task in these<br />

processes is to enclose the radioactive<br />

substances and limit the radiation<br />

exposure of personnel and environment.<br />

Furthermore, the volume of<br />

waste should be minimized. The work<br />

situation <strong>for</strong> the operating personnel<br />

must be constantly monitored and<br />

the workload should be reduced<br />

wherever possible.<br />

Sometimes, the currently available<br />

chemical and mechanical processes,<br />

reach their limit to comply with the<br />

objects mentioned be<strong>for</strong>e.<br />

Abrasive blasting of metallic structures<br />

results in the production of<br />

secondary waste, which has to be<br />

further processed in case of PCB. In<br />

chemical decontamination processes,<br />

pickling acids also must be specially<br />

treated or stored.<br />

Furthermore, all mechanical processes,<br />

such as shaving or sandblasting,<br />

also cause high workloads<br />

<strong>for</strong> the operating personnel due to<br />

vibrations, restoring <strong>for</strong>ces and noise.<br />

Chemical processes, however, can<br />

cause damage through skin contact<br />

and inhalation.<br />

Decontamination by laser beam<br />

represents an interesting alternative.<br />

This technology can expand the<br />

repertory of currently available decontamination<br />

tech nologies. Being a<br />

contact-free procedure it comes along<br />

with the advantage of waste minimization<br />

and reduces restoring <strong>for</strong>ce <strong>for</strong><br />

hand-held as well as semi-automatic<br />

application, following NEA recommendations<br />

[1].<br />

At TU Dresden laser-based decontamination<br />

tech nology has been developed<br />

during the last years. The<br />

research project LaPLUS has been<br />

aimed at the optimi zation of the<br />

chemical-toxic decontamination of<br />

concrete surfaces and the technology<br />

transfer <strong>for</strong> the radiologic decontamination<br />

of metal surfaces. For that<br />

purpose, special hand-held laser tools<br />

<strong>for</strong> the use in nuclear sites were<br />

designed and tested on a laboratory<br />

scale. An essential part of the project<br />

was the technology transfer from<br />

laboratory scale to prototype status.<br />

For that purpose, the laser-based zi.<br />

decontamination was demonstrated<br />

in a realistic environment at the<br />

Multi-Purpose Research Reactor<br />

( MZFR).<br />

Under project leadership of TU<br />

Dresden, TU Berg akademie Freiberg<br />

(Development of process analysis<br />

tools) and IABG mbH (Design of laser<br />

tools) participated in the project. The<br />

KTE accompanied the project as an<br />

associated partner and supported the<br />

technology transfer to the MZFR.<br />

This paper presents the major<br />

results achieved inside the project. It<br />

covers the laser-based decontamination<br />

of PCB containing coatings on<br />

concrete and that of radiologic contaminated<br />

metal surfaces, the design<br />

and the test of special laser tools<br />

developed <strong>for</strong> the application in<br />

nuclear facilities.<br />

2 Application of laser<br />

systems <strong>for</strong> decontamination<br />

When a laser beam hits a surface,<br />

absorption of the energy leads to rapid<br />

heating of the substrate. The resulting<br />

spatial and temporal temperature<br />

distribution depends on optical and<br />

thermal characteristics of the substrate<br />

as well as on the parameters of<br />

the laser process [2]. Heating of the<br />

surfaces can result in melting, evaporation<br />

or sublimation. All of those<br />

mechanisms can be used <strong>for</strong> the<br />

removal of unwanted species, as<br />

necessary <strong>for</strong> surface decontamination.<br />

Additionally different<br />

practical requirements defined by the<br />

decontamination task require a<br />

careful selection of a suitable laser<br />

system.<br />

Rapid heating of the treated<br />

substrate surface above evaporation<br />

temperature can be achieved using<br />

pulsed lasers. The pulse duration is directly<br />

proportional to the volume of<br />

the heat affected zone [3], which<br />

means that shorter pulses lead to<br />

smaller heat affected volume. The<br />

possibilities and requirements of<br />

radiologic decontami nation using<br />

short pulsed lasers is treated in<br />

section 3.<br />

A different result can be achieved<br />

applying a con tinuous working (cw)<br />

laser beam, which continuously heats<br />

the surface. In case of interaction with<br />

paint, e.g. epoxy based coatings, the<br />

volatile components of that coating<br />

will evaporate and incinerate after<br />

reaching the ignition temperature. A<br />

subsequent combustion process of the<br />

DECOMMISSIONING AND WASTE MANAGEMENT 543<br />

Decommissioning and Waste Management<br />

First On-site Demonstration of Laser- based Decontamination Technology in Germany ı Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann and Antonio Hurtado


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

DECOMMISSIONING AND WASTE MANAGEMENT 544<br />

coating is established. Chemical-toxic<br />

decontamination of PCB-containing<br />

paint using a continuous wave laser is<br />

treated in section 4. Application of<br />

high power diode lasers allows rapid<br />

treatment of surfaces with economically<br />

desirable laser equipment.<br />

Apart from beam parameters,<br />

further practical requirements have to<br />

be met <strong>for</strong> successful application of<br />

lasers in nuclear facilities.<br />

High mobility and toughness is<br />

required from the beam shaping optics,<br />

as opposed to regular industrial<br />

application. This demand is coped<br />

with using simplified 1D scanning<br />

optics.<br />

For the cost-efficient use of lasers,<br />

the application of optical fibres <strong>for</strong><br />

beam transport is mandatory. This<br />

allows to arrange the laser-system<br />

outside the contaminated area, thus<br />

eliminating the risk of contamination<br />

<strong>for</strong> the laser- system. Only part of the<br />

fibre and the beam-shaping laser<br />

optics will be placed inside the control<br />

area and can still be reused on several<br />

decommissioning sites, as explained<br />

in section 6.<br />

3 Decontamination<br />

of metal surfaces<br />

Laser-based cleaning of metal surfaces<br />

has been established within the last<br />

decade in applications like pre-treatment<br />

<strong>for</strong> welding, restauration of art<br />

and decoating of paint. In these cases,<br />

a laser beam is scanned and moved<br />

over the soiled surfaces and removes<br />

adhering unwanted species. High<br />

process selectivity can be achieved<br />

on metal surfaces, because a large<br />

fraction of the beam is reflected on<br />

blank metals as opposed to higher<br />

absorption on organics and oxides.<br />

Laser-based cleaning substitutes<br />

chemical as well as mechanic abrasive<br />

processes and results in a reduction of<br />

waste of factor 2.6 J/cm² <strong>for</strong><br />

austenitic steel, >3.5 J/cm² <strong>for</strong> ferritic<br />

steel and >5.3 J/cm² <strong>for</strong> zinc plated<br />

Wavelength<br />

Average <strong>Power</strong><br />

Pulse-Energy<br />

Peak <strong>Power</strong><br />

Pulse length<br />

Pulse frequency<br />

Scanner<br />

1064 nm<br />

150 W<br />

<strong>11</strong>.5 mJ @ 12 kHz<br />

<strong>11</strong>2 kW @ 12 kHz<br />

102 ns@ 12kHz<br />

12 - 40 kHz<br />

Spot diameter 472 µm<br />

Scanning width<br />

Mass of optics<br />

60 mm<br />

1,7 kg<br />

| Tab. 1.<br />

Specifications of Nd:YAG Laser CL150.<br />

| Fig. 1.<br />

Idealized scenarios of contamination, blue color symbolizes contamination:<br />

contamination of uncovered surfaces (A),<br />

on top of covered surface (B) and<br />

on as well as under covering layer (C),<br />

contamination within oxide layer (D).<br />

| Fig. 2.<br />

Ablation depth as function of cumulative energy with a pulse duration<br />

of 105 ns.<br />

| Fig. 3.<br />

Concentration of surrogates on sample surface prior and after<br />

decontamination.<br />

Decommissioning and Waste Management<br />

First On-site Demonstration of Laser- based Decontamination Technology in Germany ı Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann and Antonio Hurtado


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

steel. Tested paint layers can be<br />

ablated at fluence above 1.8 J/cm²,<br />

while oxide layers can be removed at<br />

even lower fluence.<br />

The achieved ablation depth <strong>for</strong><br />

the above mentioned materials is<br />

plotted in Figure 2.<br />

It can be seen that the removal<br />

speed is about 100 times higher <strong>for</strong><br />

paint than <strong>for</strong> metals, what supports<br />

the selective removal of the paint on<br />

the metal ground. A variation of laser<br />

fluences between 1.8 – 3.5 J/cm² has<br />

been tested <strong>for</strong> its decontamination<br />

effect on blank austenitic steel. Single<br />

scans resulted in a contamination<br />

reduction of 50 – 90 % <strong>for</strong> all surrogates.<br />

Similar tests have been conducted<br />

on painted surface applying a<br />

laser fluence of 4.4 J/cm², to prove<br />

complete removal of the covering<br />

layer (Figure 3).<br />

Both tested cases result in a<br />

decreased contaminant concentration<br />

and decontamination factors up to<br />

98.9 have been achieved.<br />

4 Laser-based PCB<br />

degradation<br />

The TU Dresden has developed a PCB<br />

decontamination process by utilising<br />

a continuous wave diode laser.<br />

Laboratory experiments on concrete<br />

surfaces coated with epoxy paint<br />

demonstrated a reduction of 96.83 %<br />

of the PCB value. The PCB decomposition<br />

rate on the surface and in<br />

the exhaust gas be<strong>for</strong>e filtering is<br />

88.75 % [16].<br />

PCB decomposes at temperatures<br />

above 800 °C. At optimized process<br />

parameters the temperature in the<br />

laser spot is much higher [17], enable<br />

the full decomposition of the PCB.<br />

Rapid cooling of the exhaust gas is<br />

required to prevent the <strong>for</strong>mation of<br />

toxic polychlorinated dioxins and<br />

furans (PCDD/PCDF) which can be<br />

ensured by the laser processing. The<br />

application of a fabric filter in the<br />

extraction and filtration unit limits<br />

the temperature of filtered exhaust<br />

gas to


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

DECOMMISSIONING AND WASTE MANAGEMENT 546<br />

| Fig. 6.<br />

Overview of tested samples as supplied by KTE (A– lid of a 200 liter drum, B– steel sheet, C– coated u<br />

profile, D– locker door part).<br />

| Fig. 7.<br />

Components <strong>for</strong> decontamination of complex metal surfaces, (A) laser tool <strong>for</strong> the decoating of flat<br />

surfaces and (B) laser tool <strong>for</strong> the decontamination of edges.<br />

transported using optical fibres to<br />

reach the experimental area. This<br />

feature allows <strong>for</strong> the protection of<br />

costly laser equipment from contamination<br />

in any case, as described in<br />

section 2. The user can simply equip<br />

the required laser tool and start the<br />

decontamination process. Suction<br />

and filtration of mobilized contamination<br />

at high efficiency is facilitated<br />

applying a commercial filtration and<br />

suction unit by ULT GmbH. The<br />

localized suction of aerosols leads to<br />

prevention of recontamination and<br />

guarantees clear vision <strong>for</strong> the user at<br />

all times.<br />

Restricted access to the experimental<br />

area is beneficial to simplify the<br />

laser safety requirements, even in case<br />

of a technology demonstration.<br />

A full-fledged risk assessment<br />

according to §5 of ArbSchG [18] and<br />

§3 of ArbStättV [19] has been filed to<br />

scan <strong>for</strong> any arising risks <strong>for</strong> operators<br />

coming along with the application of<br />

the whole set up. All requirements of<br />

the occupational safety have been<br />

tested and proven. The risk assessment<br />

covers the following subject<br />

area:<br />

p General occupational safety<br />

p Laser safety<br />

p Safety from toxic substances (PCB)<br />

p Radiation safety<br />

p Respiratory protection<br />

The risk assessment complies with the<br />

laws, guidelines (ASR), workers compensation<br />

board rules and provisions.<br />

The risk assessment was finalized by<br />

TU Dresden and IABG and checked by<br />

KTE and has been the fundamental<br />

milestone <strong>for</strong> approval of the on-site<br />

demonstration at MZFR.<br />

5.3 Decontamination of<br />

complex metal geometries<br />

The decontamination of metal surfaces<br />

at MZFR was conducted on<br />

paints and metals prior unknown, to<br />

check the ability to transfer results<br />

from laboratory scale to on-site tests.<br />

To test limits of the laser tool, samples<br />

of varying geometries have been<br />

supplied by KTE, e.g. painted steel<br />

sheets, bracket steel or complex surfaces<br />

like ridges/spline/serration, as<br />

shown in Figure 6.<br />

Different laser tools were designed<br />

to ensure the decontamination at<br />

different surface geometries, e.g. in<br />

case of tighter angles the tool was<br />

adapted to maintain the complete<br />

suction of particles and aerosols<br />

during the laser process. A tool change<br />

can be completed within 1.5 min,<br />

Figure 7.<br />

Similar ablation characteristics<br />

were found <strong>for</strong> all tested samples as<br />

compared to the laboratory results<br />

(Figure 2). This implies that generalized<br />

material characteristics <strong>for</strong> paint<br />

layers and metals can be applied.<br />

From practical point of view, the<br />

flexibility of the laser ablation process<br />

was verified, as all tested geometries<br />

were completely cleaned without<br />

residues (Figure 8).<br />

5.4 Decontamination of<br />

concrete walls and floor<br />

at MZFR<br />

Decontamination of concrete demands<br />

solutions <strong>for</strong> a multitude of<br />

demands, e.g. uneven surfaces, varying<br />

thicknesses of paint, different<br />

shapes of surfaces and unknown PCB<br />

concentration in the decontamination<br />

paint. The practicability of the<br />

developed laser system was tested on<br />

walls and floors at the MZFR on<br />

sample surfaces sized 30 x 30 cm².<br />

The decontamination paint on the<br />

wall was thin and exhibited a high<br />

PCB concen tration (see Table 3). The<br />

floor featured lower PCB concentration<br />

and very thick paint (approx.<br />

1.5 mm).<br />

In Figure 9 the side view of the<br />

laser head is shown during the decontamination.<br />

During the process all<br />

arising by-products, like visible particles<br />

and gases, are extracted by the<br />

extraction and filtration unit and do<br />

not endanger the working staff.<br />

Figure 10 displays the result<br />

of decontamination on a floor<br />

| Fig. 8.<br />

Steel sheet partly decoated.<br />

| Fig. 9.<br />

Laser head during the decontamination of the floor.<br />

Decommissioning and Waste Management<br />

First On-site Demonstration of Laser- based Decontamination Technology in Germany ı Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann and Antonio Hurtado


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

| Fig. 10.<br />

Demonstration field floor (A) be<strong>for</strong>e and (B) after the decontamination<br />

with multiple scans.<br />

Sample number Position Be<strong>for</strong>e decontamination<br />

PCB in mg/kg<br />

demonstration field be<strong>for</strong>e (A) and<br />

after (B) the laser process.<br />

In Figure <strong>11</strong> the laser head is<br />

displayed with different adaptions<br />

<strong>for</strong> flat areas (picture A and B) and<br />

internal corners (picture C). The<br />

modular set-up enables a fast change<br />

of the adaptions omitting any changes<br />

to the leaser heads basic components.<br />

Using such adaptions the decontamination<br />

of a wall with adjoining<br />

corners was demonstrated, illustrating<br />

the clear advantage compared<br />

to shaving equipment, where usually<br />

boundary areas need an additional,<br />

time- consuming handling step.<br />

The removal of thick layers and/or<br />

remaining soot can be realized by<br />

multiple passes. Both have been<br />

demonstrated in Karlsruhe.<br />

During the demonstration an<br />

average laser power of 3.7 kW was<br />

applied, resulting in a practical<br />

determined ablation speed of 6 m²/h<br />

5.5 PCB sampling at the MZFR<br />

Within the demonstration PCB<br />

sampling was per<strong>for</strong>med to prove the<br />

decontamination rate of the laserbased<br />

paint removal. This study aimed<br />

at the evaluation of the PCB concentration<br />

of the primary source. To<br />

release material the PCB-concentration<br />

must amount less than 50 mg/<br />

kg. Samples with mass > 50 mg were<br />

extracted using an electric scraper.<br />

AGROLAB Labor GmbH per<strong>for</strong>med<br />

the analysis of the samples according<br />

to DIN EN 15308. The samples taken<br />

be<strong>for</strong>e the decontamination consisted<br />

mainly of decontamination paint.<br />

After the decontamination no paint<br />

remained on the surface anymore,<br />

there<strong>for</strong>e the now surfacing concrete<br />

was sampled.<br />

In Figure 12 the sampled surfaces<br />

are shown be<strong>for</strong>e PCB sampling (A),<br />

after PCB sampling (B) and after<br />

decontamination and the consecutive<br />

PCB sampling (C). The corresponding<br />

PCB concentration is shown in<br />

Table 3.<br />

The laser-based decontamination<br />

process resulted in an average reduction<br />

of PCB of 98.7 % within a practical<br />

application. This value is even<br />

higher than the value of the laboratory<br />

experiments per<strong>for</strong>med at the TU<br />

Dresden [16]. The remaining PCB<br />

concentration after the decontamination<br />

is mainly caused by the PCB<br />

inside the concrete matrix. After sealing<br />

the walls with PCB containing<br />

paint the PCB diffuses from the decontamination<br />

paint (primary source)<br />

into the concrete. Alternative laserbased<br />

technologies are available to<br />

remove and to vitrify concrete surfaces<br />

in a single process step [20; 21].<br />

| Fig. <strong>11</strong>.<br />

Laser head <strong>for</strong> concrete decontamination;<br />

A- laser head (round version), B- laser head (angled), C- laser head internal corner.<br />

After decontamination<br />

PCB in mg/kg<br />

Difference<br />

mg/kg<br />

| Fig. 12.<br />

Examples of a demonstration field:<br />

(A)- be<strong>for</strong>e sampling, (B)- after decontamination and (C)- after decontamination and sampling.<br />

6 Conclusion and future<br />

prospects<br />

The TU Dresden successful demonstrated<br />

the application of laser-based<br />

decontamination on radiologic and<br />

chemical-toxic contaminated metal<br />

and concrete surfaces at the MZFR in<br />

Karlsruhe. All necessary documents to<br />

obtain acceptance of the KTE were<br />

prepared by TU Dresden and approved<br />

by KTE. An independent work<br />

permit was available to TU Dresden.<br />

The selective decontamination on<br />

metal surfaces is characterized by a<br />

process stop after removing any coating<br />

(decontamination paint, oxide<br />

and/or contamination). The practical<br />

removal rate on metal is 0.42 m²/h<br />

using a mean laser power of 150 W<br />

and an average paint thickness of<br />

165 µm. The application of laser-based<br />

decontamination to PCB-contaminated<br />

concrete surfaces resulted in<br />

a mean reduction of 98.7 % of the<br />

PCB-concentration. During the onsite-demonstration<br />

a removal rate of<br />

6 m²/h of paint from concrete walls<br />

with 3.7 kW laser power were accomplished.<br />

The handheld laser tools<br />

Reduction<br />

%<br />

1 Floor 9.3 0.23 9.07 97.5<br />

2 Floor 30.1 0.04 30.06 99.9<br />

3 Wall 3,810 67.3 3,742.7 98.2<br />

4 Wall 2,360 19 2,341 99.2<br />

5 Wall 1,976 18.9 1,957.1 99.0<br />

6 Wall 2,510 14.2 2,495.8 99.4<br />

7 Wall 2,272 49.7 2,222.3 97.8<br />

| Tab. 3.<br />

PCB concentration of the sampling be<strong>for</strong>e and after the decontamination.<br />

Ø 98.7 %<br />

DECOMMISSIONING AND WASTE MANAGEMENT 547<br />

Decommissioning and Waste Management<br />

First On-site Demonstration of Laser- based Decontamination Technology in Germany ı Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann and Antonio Hurtado


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

DECOMMISSIONING AND WASTE MANAGEMENT 548<br />

enable high flexibility and reduce<br />

restraining <strong>for</strong>ces <strong>for</strong> the working<br />

staff. The dry laser cleaning process<br />

runs completely without abrasive or<br />

chemical aids thus reducing the<br />

amount of secondary waste drastically.<br />

The determined removal rates on<br />

concrete and metal are highly affected<br />

by the thickness and optical pro perties<br />

of the decontamination paint. The<br />

great advantage of lasers is the scalability<br />

of their power. The use of<br />

commercial available lasers with<br />

higher power would result in higher<br />

ablation rates without any changes to<br />

the applied laser tools.<br />

The on-site demonstration at<br />

MZFR Karlsruhe verified the possibilities<br />

of that technology <strong>for</strong> decontamination<br />

of contaminated surfaces.<br />

Robust laser tools can offer an<br />

alternative <strong>for</strong> surface decontamination<br />

in the near future. Following the<br />

positive results of the research project<br />

LaPLUS more detailed insight into the<br />

behavior of mobilized aerosols as well<br />

as further tests within control areas<br />

will be provided from follow up projects.<br />

Industrial application of the<br />

laser-based decontamination process<br />

should become feasible in the next<br />

years<br />

Acknowledgement<br />

The on-site demonstration at the<br />

MZFR in Karlsruhe was carried out<br />

inside the Research and Development<br />

Project LaPLUS financed by The<br />

German Federal Ministry of Education<br />

and Research (BMBF) under the<br />

contract number 15S9215A.<br />

We would like to thank all colleagues<br />

of the MZFR that supported<br />

the on-site-demonstration with enthusiasm<br />

and useful evidences as well<br />

as practical drive. Our special thanks<br />

go to the members of our technical<br />

staff, who were a great support during<br />

the exhausting time of demonstration<br />

and the whole project as well.<br />

References<br />

[1] <strong>Nuclear</strong> Energy Agency: R&D and Innovation Needs <strong>for</strong><br />

Decommissioning of <strong>Nuclear</strong> Facilities. 2014 (7191)<br />

[2] BLIEDTNER, M.: Lasermaterialbearbeitung. München:<br />

CARL HANSER Verlag GMBH, 2013<br />

[3] Leitz, K.-H.; Redlingshöfer, B.; Reg, Y.; Otto, A.; Schmidt, M.:<br />

Metal Ablation with Short and Ultrashort Laser Pulses. In:<br />

Physics Procedia 12 (20<strong>11</strong>), S. 230–238<br />

[4] Büchter, E.: Entwicklung eines Hochleistungs-Laserstrahl-<br />

Reinigungsgerätes zur Ressourcen schonenden<br />

Entschichtung von Oberflächen. Herzogenrath, 2004<br />

[5] Carvalho, L.; Pacquentin, W.; Tabarant, M.; Maskrot, H.;<br />

Semerok, A.: Growth of micrometric oxide layers to explore<br />

laser decontamination of metallic surfaces. In: EPJ <strong>Nuclear</strong><br />

Sciences & Technologies 3 (2017), S. 30<br />

[6] Delaporte, Ph.; Gastaud, M.; Marine, W.; Sentis, M.;<br />

Uteza, O.; Thouvenot, P.; Alcaraz, J. L.; Le Samedy, J. M.;<br />

Blin, D.: Radioactive oxide removal by XeCl laser.<br />

In: Applied Surface Science 197-198 (2002), S. 826–830<br />

[7] Kim, D.; Lim, H.: Laser Decontamination of Carbon Steel<br />

Surfaces. In: ISIJ <strong>International</strong> (2003), Nr. 43, S. 1289–1291<br />

[8] Leontyev, A.; Semerok, A.; Farcage, D.; Thro, P.-Y.; Grisolia, C.;<br />

Widdowson, A.; Coad, P.; Rubel, M.: Theoretical and<br />

experimental studies on molybdenum and stainless steel<br />

mirrors cleaning by high repetition rate laser beam.<br />

In: Fusion Engineering and Design 86 (20<strong>11</strong>),<br />

9-<strong>11</strong>, S. 1728–1731<br />

[9] Edelson, M. C.; Pang, H.: A laser-based solution to industrial<br />

decontamination problems. 1995 (ICALEO 1995 768)<br />

[10] Potiens, A. J.; Dellamano, J. C.; Vicente, R.; Raele, M. P.;<br />

Wetter, N. U.; Landulfo, E.: Laser decontamination of the<br />

radioactive lightning rods. In: Radiation Physics and<br />

Chemistry 95 (2014), S. 188–190<br />

[<strong>11</strong>] Sadanori, S.; Seiji, A.; Inoue, T.: Applying laser technology to<br />

decommissioning <strong>for</strong> nuclear power plant (Advanced<br />

High-<strong>Power</strong> Lasers and Applications). Osaka, Japan, 1999<br />

[12] Takakuni, H.; Yutaka, K.; Masato, M.: Application of a laser to<br />

decontamination and decommissioning of nuclear facilities<br />

at JAERI (Advanced High-<strong>Power</strong> Lasers and Applications).<br />

Osaka, Japan, 1999<br />

[13] Vatry, A.; Grisolia, C.; Delaporte, Ph.; Sentis, M.: Removal<br />

of in vessel Tokamak dust by laser techniques. In: Fusion<br />

Engineering and Design 86 (20<strong>11</strong>), 9-<strong>11</strong>, S. 2717–2721<br />

[14] Nilaya, J. P.; Raote, P.; Kumar, A.; Biswas, D. J.: Laser-assisted<br />

decontamination – A wavelength dependent study. In:<br />

Applied Surface Science 254 (2008), Nr. 22, S. 7377–7380<br />

[15] Delaporte, Ph.; Gastaud, M.; Marine, W.; Sentis, M.; Uteza, O.;<br />

Thouvenot, P.; Alcaraz, J. L.; Le Samedy, J. M.; Blin, D.: Dry<br />

excimer laser cleaning applied to nuclear decontamination.<br />

In: Applied Surface Science 208-209 (2003), S. 298–305<br />

[16] Anthofer, A.; Kögler, P.; Friedrich, C.; Lippmann, W.;<br />

Hurtado, A.: Laser decontamination and decomposition<br />

of PCB-containing paint. In: Optics & Laser Technology 87<br />

(2017), S. 31–42<br />

[17] Anthofer, A.: Oberflächenentschichtung mittels<br />

Laserstrahlung. Untersuchungen zur Dekontamination<br />

radioaktiv und chemisch-toxisch belasteter Betonoberflächen<br />

mittels Lasertechnologie. Dissertation, Dresden, 2014<br />

[18] Bundesministerium der Justiz und für Verbraucherschutz:<br />

Gesetz über die Durchführung von Maßnahmen des<br />

Arbeitsschutzes zur Verbesserung der Sicherheit und des<br />

Gesundheitsschutzes der Beschäftigten bei der Arbeit<br />

(idF v. 31. 8. 2015) (1996-08-07)<br />

[19] Bundesministerium der Justiz und für Verbraucherschutz:<br />

Verordnung über Arbeitsstätten<br />

(idF v. 18. 10. 2017) (2004-08-12)<br />

[20] Lippmann, W.; Herrmann, M.; Pietsch, C.; Reinecke, A.;<br />

Hille, C.; Wolf, R.; Zeuner, A.: LASABA II : Dekontamination<br />

silikatischer Oberflächen in kerntechnischen Anlagen mittels<br />

Laserabtrag bei gleichzeitiger Abproduktkonditionierung.<br />

Abschlussbericht, Dresden, 2008<br />

[21] Hurtado, A.; Littwin, R.; Lippmann, W.: MANOLA: Manipulator<br />

gesteuerter Oberflächenabtrag durch Lasertechnologie.<br />

Abschlussbericht, Dresden, 20<strong>11</strong><br />

Authors<br />

Torsten Kahl,<br />

Georg Greifzu,<br />

Marion Herrmann,<br />

Wolfgang Lippmann,<br />

Antonio Hurtado<br />

Technische Universität Dresden<br />

Chair of Hydrogen and<br />

<strong>Nuclear</strong> Energy<br />

Institute of <strong>Power</strong> Engineering<br />

George-Bähr-Str. 3b<br />

01062 Dresden<br />

Germany<br />

Carsten Friedrich,<br />

Christian Held<br />

Kerntechnische<br />

Entsorgung Karlsruhe<br />

Hermann-von-Helmholtz-Platz 1<br />

76344 Eggenstein-Leopoldshafen<br />

Germany<br />

Decommissioning and Waste Management<br />

First On-site Demonstration of Laser- based Decontamination Technology in Germany ı Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann and Antonio Hurtado


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Scientists and Professionals from all around the World in Karlsruhe:<br />

The 25 th Edition of the Frédéric Joliot/Otto Hahn Summer School<br />

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

Victor Hugo Sanchez-Espinoza<br />

The Institute of Neutron Physics and Reactor Technology (INR) of the Karlsruhe Institute of Technology (KIT) together<br />

with the Commissariat à l’Énergie Atomique (CEA) hosted this year the “Frédéric Joliot/Otto Hahn (FJOH) Summer<br />

School“ at the Akademiehotel in Karlsruhe from August 21 st to August 30 th 2019. The topic of this year’s school was<br />

“Innovative Reactors: Matching the Design to Future Deployment and Energy Needs”.<br />

549<br />

REPORT<br />

It was organized in six technical blocks<br />

and one seminar devoted to “Public<br />

Acceptance of Energy Technologies”<br />

given by S. Hirschberg from PSI<br />

Switzerland. The first introduc tory<br />

block was devoted to “Various Innovative<br />

Reactor Concepts <strong>for</strong> Various<br />

Missions” and consisted in two<br />

lectures. The first one entitle “The<br />

Deciding Factors in Opting <strong>for</strong> a<br />

Particular Reactor Technology” was<br />

given by J. Guidez (CEA, France) and<br />

the second one entitled “Promising<br />

Reactor Concepts <strong>for</strong> Multiple Applications”<br />

given by S. Monti (IAEA).<br />

The second block was focused on the<br />

topic “Near-term-deployment <strong>Power</strong>to-grid<br />

Integrated LWR Technology”<br />

and it consisted of three lectures as<br />

follows. The first lecture deals with<br />

“From Specific Design Criteria to an<br />

Advanced Modular LWR Concept: Approach,<br />

Methods, Validation” given by<br />

F. Morin (CEA, France). The second<br />

lecture was entitled “The Challenges<br />

of Designing and Licensing a First-ofa-kind<br />

Reactor Prototype, even a Small<br />

One” given by D. Delmastro (CNEA,<br />

Argentina) and finally the third lecture<br />

was entitled “The Reliability and<br />

Safety Case of a Reactor Equipped<br />

with Passive Systems” presented by<br />

A. Schaffrath (GRS, Germany). The<br />

third block was dedicated to the topic<br />

“Multi-purpose Molten Salt Reactors”<br />

and it was organized as three<br />

lectures to the following issues. The<br />

first lecture was entitled “MSR Design<br />

Principles, Concepts, Modelling Approaches,<br />

and Methods” given by T.<br />

Abram (Univ. of Man chester, UK). The<br />

second lecture was devoted to “From<br />

the MSR Physics Principles to a Plant<br />

Layout” given by E. Merle (Grenoble<br />

INP & CNRS) and the third lecture<br />

entitled “Fuel Salts Chemistry and<br />

Materials Compati bility” given by<br />

V. Ignatev (KI, Russia). The fourth<br />

block was devoted to “­<strong>Nuclear</strong> Technology<br />

<strong>for</strong> Space ­Propulsion and<br />

Manned Space ­Exploration” and it<br />

consisted of two lectures. The first one<br />

entitled “ <strong>Nuclear</strong> Rocket Propulsion:<br />

Background, Physics and Methods,<br />

Design and Tests” given by W. Emrich<br />

Jr. ( NASA, USA) and the second one<br />

entitled “The Challenge of Fueling a<br />

<strong>Nuclear</strong> Engine <strong>for</strong> Space Exploration”<br />

given by J. Witter (BWXT, USA).<br />

The fifth block was devoted to<br />

“Minimum-intervention Long-life<br />

Breed-­and-burn Fast Reactors” and<br />

it consisted of two lectures. The first<br />

one entitled “Physics of Breed-andburn<br />

Reactors, Optimized Core and<br />

Fuel Design, Licensing Case” given by<br />

K. Weaver (INL, USA) and the second<br />

one entitled “Cladding and Structural<br />

Materials <strong>for</strong> Very Long In-core<br />

Residence Times“ given by Y. Decarlan<br />

(CEA, France). The sixth block was<br />

devoted to “Reactor Concepts <strong>for</strong><br />

Process Heat and <strong>Power</strong>-to-gas<br />

­Applications” and it consisted of two<br />

lectures. The first one entitled<br />

“ Designing a Small Reactor to Bring<br />

<strong>Power</strong> to Remote Areas or to Produce<br />

Process Heat” given by J. Kloosterman<br />

(TU Delft, Netherlands) and the<br />

second one entitled „Techno-economic<br />

Assessment of Hydrogen Production<br />

from <strong>Nuclear</strong> Energy“ given by<br />

J. Witter (BWXT, USA). In the frame of<br />

the technical visit, a guided tour was<br />

organized <strong>for</strong> the TrasnetBW GmbH<br />

in Wendlingen, which is one of the<br />

largest system Control Centre <strong>for</strong><br />

Baden Württemberg with headquarter<br />

in Stuttgart. It operates the electricity<br />

transmission grid in the German state<br />

of Baden-Württemberg, control and<br />

monitor the energy flows through the<br />

grid, and per<strong>for</strong>m the necessary<br />

maintenance and network planning<br />

and development activities. This year,<br />

44 participants of 18 countries (EU,<br />

Asia, Latin America, East Europe,<br />

Middle East, Africa, and USA) attended<br />

the FJOH Summer School.<br />

Recognized experts of 10 different<br />

countries from Asia, EU, South<br />

America, USA and East Europe from<br />

Academia, industry, research and<br />

TSOs gave high-level lectures on topics<br />

of their expertise. During the ten days,<br />

the participants had the opportunity<br />

to exhaustive discussions with the<br />

lecturers and other participants. Apart<br />

| The 25 th Edition of the Frédéric Joliot/Otto Hahn Summer School on <strong>Nuclear</strong><br />

Reactors “Physics, Fuels and Systems“, participants and lecturers in Karlsruhe.<br />

from the technical issues, another goal<br />

of the FJOH Summer School is to<br />

intensify the networking among the<br />

participants of different continents<br />

and nationalities with the common objective<br />

of enhanced safety worldwide.<br />

A well-recognized tradition during<br />

the FJOH Summer School is an extensive<br />

and diverse program with social<br />

events to get familiar with the German<br />

culture and way of life as well as to<br />

foster the exchange among the participants.<br />

This year, the Sunday trip<br />

consisted in the visit of the Technical<br />

Museum Speyer and afterwards, a<br />

Canoe Tour at the Old Rhine. After the<br />

Museum visit, the participants have<br />

free time to get familiar with the<br />

downtown of Speyer and to visit the<br />

Romanesque Cathedral that houses<br />

the grave f most important kings and<br />

emperors e.g. it is the burial place <strong>for</strong><br />

emperor Konrad II and his wife.<br />

The next Summer School will be<br />

hosted by CEA in Aix-en-Provence,<br />

from August 26 th to September 4 th ,<br />

2020 and it will be devoted to “High-­<br />

fidelity Modelling and Simulation<br />

of <strong>Nuclear</strong> Reactors: Turning a<br />

Promise into Reality”.<br />

Author<br />

Dr.-Ing. Victor Hugo Sanchez-Espinoza<br />

Head of Group<br />

“Reactor Physics and Dynamics”<br />

Project Leader<br />

“LWR Safety Methods and Codes ”<br />

Karlsruhe Institute of Technology (KIT)<br />

Institute <strong>for</strong> Neutron Physics and<br />

Reactor Technology (INR)<br />

Hermann-von-Helmholtz-Platz 1<br />

76344 Eggenstein-Leopoldshafen<br />

Report<br />

The 25th Edition of the Frédéric Joliot/Otto Hahn Summer School on <strong>Nuclear</strong> Reactors “Physics, Fuels and Systems“ ı Victor Hugo Sanchez-Espinoza


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Special Topic | A Journey Through 50 Years AMNT<br />

550<br />

SPECIAL TOPIC | A JOURNEY THROUGH 50 YEARS AMNT<br />

Am 7. und 8. Mai<br />

2019 begingen wir<br />

das 50. Jubiläum<br />

unserer Jahrestagung<br />

Kerntechnik. Zu<br />

diesem Anlass öffnen<br />

wir unser <strong>atw</strong>-Archiv<br />

für Sie und präsentieren<br />

Ihnen in jeder<br />

Ausgabe einen<br />

historischen Beitrag.<br />

Überarbeitete<br />

Fassung eines<br />

Vortrags gehalten<br />

am 14. Mai 2002<br />

auf der Jahrestagung<br />

Kerntechnik 2002,<br />

Stuttgart,<br />

14.-16.5.2002<br />

Schutz von Mensch und Umwelt –<br />

Nukleare Anwendungen außerhalb des<br />

Energiesektors<br />

P.P. De Regge, Wien<br />

1 Einleitung Nukleare Technologien liefern immer mehr bedeutende Beiträge zum Schutz des Menschen und<br />

seiner Umwelt sowie zur Verbesserung der Lebensstandards. Die <strong>International</strong>e Atomenergie Organisation (IAEO) ist<br />

durch die Vereinten Nationen (UN) per Statut beauftragt, solche nuklearen Technologien zu entwickeln, zu fördern<br />

und deren Einsatz zu unterstützen. Sie erreicht dies durch Veröffentlichung von In<strong>for</strong>mationen, Ausbildung, Zulieferung<br />

von Ausrüstung, Dienstleistungen und Erstellung von Sicherheitsnormen.<br />

In diesem Beitrag wird diese Rolle der IAEO erläutert<br />

und es wird weiter der heutige Stand der nuklearen<br />

Anwendungen in folgenden Bereichen dargestellt:<br />

p Nuklearmedizin und Gesundheitspflege<br />

p Veterinärmedizin und Viehzucht<br />

p Bodenkultivierung und Düngemittel<br />

p Wasserversorgung und Umweltschutz<br />

p Anbau von Kulturpflanzen<br />

p Schädlingsbekämpfung<br />

p sonstige Anwendungen, wie Landminenräumung und<br />

Schutz des kulturellen Erbes<br />

2 Rolle der <strong>International</strong>en Atomenergie<br />

Organisation<br />

Die Ziele der IAEO hinsichtlich der Anwendung von<br />

nuklearen Technologien sind ein Beitrag zur nachhaltigen<br />

Entwicklung und zum Umweltschutz in den Bereichen<br />

medizinische Versorgung, Ernährung, Landwirtschaft und<br />

Industrie sowie Versorgung mit Wasser. Die Aktivitäten<br />

der IAEO in diesen Bereichen haben Grundbedürfnisse der<br />

Menschheit im Fokus. Nukleare Technologien sollen dabei<br />

wesentlichere und konkurrenzfähigere Vorteile bieten als<br />

vergleichbare andere Technologien.<br />

Entsprechende nukleare Technologien und Anwendungen<br />

werden im Rahmen von koordinierten, angewandten<br />

Forschungsprojekten entwickelt. Dazu werden<br />

über mehrere Jahre Beiträge von potenziellen Anwendern<br />

aus Industriestaaten und Entwicklungsländern koordiniert.<br />

Bereits einsetzbare Technologien werden verfügbar<br />

gemacht und interessierten Ländern bereit gestellt; über<br />

technischer Kooperationsprojekte wird die technische<br />

Ausrüstung beschafft und es wird für eine entsprechende<br />

notwendige Ausbildung zum Betrieb von Analgen und<br />

Einsatz von Technologien gesorgt.<br />

Im derzeitigen Gesamtbild der weltweiten Entwicklung<br />

erkennt man folgende wichtige Problemfelder:<br />

p Die Weltbevölkerung hat fast 6 Milliarden Menschen<br />

erreicht und wächst pro Jahr um rund 80 Millionen<br />

an – dies entspricht etwa der deutschen Gesamtbevölkerung.<br />

p Fast eine Milliarde Menschen sind chronisch unterernährt<br />

und es fehlt ihnen zudem an einer zuver lässigen<br />

Wasserversorgung.<br />

p Sechs Millionen Kinder in der Dritten Welt sterben<br />

jedes Jahr aufgrund Mangelernährung; ebenso viele<br />

Erwachsene sterben in den Industriestaaten jährlich an<br />

Krebs. Weitere Millionen sind erkrankt oder sterben an<br />

bakteriellen oder von Viren verursachten Infektionen<br />

und Krankheiten. Zum Beispiel streben allein an den<br />

Folgen der Malaria weltweit mehr als 2 Millionen<br />

Menschen pro Jahr.<br />

Viele heutige Aktivitäten des Menschen verschmutzen und<br />

verändern die Umwelt und sind auf längere Sicht nicht<br />

tragbar.<br />

Die von der IAEO entwickelten und unterstützten<br />

nuklearen Anwendungen außerhalb des Energiesektors<br />

haben daher auch<br />

p die Verbesserung der Ernährungssituation und Wasserversorgung,<br />

p die Verbesserung des Gesundheitsschutzes sowie<br />

p den Schutz der Umwelt durch Analyse, Vorbeugung<br />

und Sanierung belasteter Bereiche<br />

zum Ziel. Trends der IAEO-Unterstützung von Projekten<br />

im Rahmen der technischen Kooperation sind in Tabelle 1<br />

dargestellt.<br />

3 Gesundheitsvorsorge<br />

Die meist verbreiteten diagnostischen Anwendungen von<br />

ionisierender Strahlung und nuklearer Technologie sind<br />

die Röntgendiagnose sowie die Diagnostik mit radioaktiven<br />

biologischen Indikatoren in der Nuklearmedizin.<br />

Die therapeutische Anwendung ist die Strahlungsonkologie,<br />

bei der mit einer wirksamen Strahlungsdosis an<br />

bestimmten krebsgefährdeten Stellen des Körpers mittels<br />

einer externen Strahlungsquelle (Teletherapie) oder<br />

mittels einer implantierten Strahlungsquelle (Brachytherapie)<br />

gearbeitet wird. Auch offene Strahlungsquellen,<br />

wie radioaktive Substanzen zur Behandlung von Schilddrüsenkrebs,<br />

werden manchmal verwendet. Neue Anwendungen<br />

von radioaktiven Substanzen werden er<strong>for</strong>scht<br />

und entwickelt.<br />

Special Topic | A Journey Through 50 Years AMNT<br />

Protection of Man and Environment – <strong>Nuclear</strong> Usage Outside of Energy Sector ı P.P. De Regge, Wien


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Projekte<br />

Eine revolutionäre Entwicklung der diagnostischen<br />

Radiologie war die Computertomografie, die genaue<br />

Schnittbilder von bestimmten Körperregionen in allen<br />

anatomischen und funktionellen Einzelheiten darstellt.<br />

Die Nuklearmedizin verwendet Radioisotope in über<br />

Hundert standardisierten Methoden für diagnostische,<br />

therapeutische und Forschungszwecke in den Bereichen<br />

Onkologie, Endokrinologie, Kardiologie, Neurologie und<br />

Nephrologie. Industriestaaten haben derzeit etwa<br />

20 Gammakameras pro einer Million Einwohner zur<br />

Verfügung, während in Entwicklungsländern nur eine<br />

pro eine Million Einwohnern verfügbar ist, und dies<br />

nur in den größeren Metropolen.<br />

Molekulare Nuklearmedizin ist ein neuer Anwendungsbereich.<br />

Erkrankungen werden auf zellulären oder<br />

gene tischem Niveau untersucht und identifiziert. Polymerase<br />

Kettenreaktionen bilden extrem empfindliche<br />

Diagnosetechniken für epidemische Infektionen, Krebs<br />

oder weiterer Symptome und sind z. B. von der US Food<br />

and Drug Administration als in vitro Tests genehmigt<br />

worden.<br />

Radioimmunoassay ist eine der wichtigsten Komponenten<br />

der in vitro Diagnosetechniken für die Quan tifizierung<br />

von Proteinänderungen oder anderer erkrankungsbedingter<br />

Metabolismusprodukte. Ein Zuwachs des<br />

Weltumsatzes für diese in vitro Diagnosetechniken von<br />

20 Milliarden Dollar im Jahr 1999 auf bis zu 26 Milliarden<br />

Dollar im Jahr 2004 wird prognostiziert und spiegelt die<br />

Bedeutung dieser Diagnosemöglichkeit wieder.<br />

Die Anzahl von Krebserkrankungen, eine der häufigsten<br />

Todesursachen in den Industriestaaten, nimmt auch aufgrund<br />

der steigenden Lebenserwartung in den Ländern<br />

der Dritten Welt ständig zu. Von etwa 10 Millionen neuen<br />

Krebsfällen im Jahr 2000 entfällt jeweils die Hälfte auf<br />

Industriestaaten und auf Entwicklungsländer. Es wird<br />

erwartet, dass 10 Millionen von 15 Millionen neuer<br />

Jahr<br />

1980 1985 1990 1995 2000<br />

Kernenergie 26 10 10 5 5<br />

Gesundheitsvorsorge <strong>11</strong> <strong>11</strong> <strong>11</strong> 16 23<br />

Physik und Chemie 25 25 20 15 10<br />

Hydrologie 2 3 5 6 <strong>11</strong><br />

Landwirtschaft und Ernährung 24 25 22 24 18<br />

Industrie 5 12 12 8 8<br />

Strahlenschutz 6 12 17 24 20<br />

Abfallverarbeitung 1 2 3 2 5<br />

| Tab. 1.<br />

IAEO Unterstützung von Projekten im Rahmen der Technischen Kooperation (Prozent der Gesamtfinanzierung)<br />

| Teletherapie.<br />

Krebsfälle im Jahr 2015 in Ländern der Dritten Welt auftreten<br />

werden. Fünfzig Prozent der Krebspatienten werden<br />

mittels Radiotherapie behandelt, mit einer Überlebenschance<br />

von ungefähr 45%.<br />

Am meisten wird noch Teletherapie verwendet, jetzt<br />

mit Präzisionsbestrahlung von komplexen und unregelmäßigen<br />

Tumor<strong>for</strong>men. Für Brachytherapie mittels<br />

implantierter Strahlungsquellen werden derzeit Caesium-<br />

137 und Iridium-192 anstelle von Radium-226 verwendet.<br />

Die IAEO, zusammen mit der Weltgesundheitsorganisation<br />

(World Health Organisation, WHO), verwaltet<br />

eine Datenbank von Spitälern und Kliniken, in denen<br />

Radio therapie zur Anwendung kommt. Länder der<br />

Dritten Welt mit 85 % der Bevölkerung verfügen über<br />

ein Drittel der Radiotherapieinstallationen, etwa 2200<br />

Teletherapie geräte mit Kobalt-60 Quellen und 850<br />

Brachytherapie geräten allerdings nur einem Fünftel der<br />

Elektronen beschleuniger. Mit jedem Gerät können<br />

600 Patienten pro Jahr behandelt werden oder insgesamt<br />

nur 1,9 Millionen von 2,5 Millionen Patienten in diesen<br />

Ländern. Mehr als 5000 zusätzliche Radiotherapiegeräte<br />

werden daher bis zum Jahr 2015 gebraucht werden. Am<br />

Rande sei erwähnt, dass die IAEO und die WHO jährlich<br />

die ordnungsgemäße Anwendung sowie die eingesetzte<br />

Strahlungsdosis von etwa 600 dieser Geräte in der Dritten<br />

Welt überprüfen.<br />

Nukleare Diagnose- und Forschungtechniken auf Basis<br />

stabiler oder radioaktiver Isotope werden auch zu Maßnahmen<br />

auf dem Ernährungssektor verwendet, insbesondere<br />

zum Nachweisen von Vitamin-, Spuren element- und<br />

Mikronährstoffmangel. Diese sind mit Ursache für Anämie,<br />

Beeinträchtigungen des Sehver mögens, Wachstum und<br />

geistige Entwicklung. Differen tielle Röntgenabsorp tiometrie<br />

wird derzeit auch als nicht intrusive Methode zur<br />

Knochendichtemessungen verwendet.<br />

4 Landwirtschaft<br />

Nukleare Technologien werden in der Landwirtschaft zur<br />

Verbesserung und zum Schutz von Nahrungsmitteln<br />

verwendet sowie zur Optimierung der Viehzucht und<br />

Vorbeugung gegen Seuchen und zur Bekämpfung von<br />

Schädlingen. Biodiversität ist entstanden durch zufällige<br />

Mutationen, induziert durch kosmische und natürliche<br />

Strahlung und auch durch Transkriptionsfehler im Erbgut.<br />

Durch Anwendung von Strahlung kann deshalb die<br />

Mutationsfrequenz erhöht und beschleunigt werden<br />

und man kann in kurzer Zeit eine große Anzahl von<br />

neuen Pflanzenvarianten bilden. Die meisten erzeugten<br />

Mutationen sind nicht weiter lebensfähig oder sogar<br />

schlechter überlebensfähig als die ursprüngliche Pflanze,<br />

aber einige sind unter Umständen besser angepasst und<br />

werden daher weiter gezüchtet. Einige Beispiele werden<br />

551<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

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| Züchtungen von Pflanzen:<br />

Höhere Erträge durch den Einsatz von Bestrahlungen.<br />

hier angegeben, da sie für die betreffenden Länder und<br />

Regionen eine erhebliche Bedeutung in der zukünftigen<br />

Versorgung besitzen. Reissorten, geeignet zum Wachsen<br />

im Wechsel von Salzwasser und Süßwasser der asiatischen<br />

Flussdeltas, dienen als Grundnahrungsmittel von<br />

Millionen Menschen in Pakistan, China und Bangladesh.<br />

Gerstevarianten ohne Hülse, dürretolerantes Sorghum<br />

und Reis werden durch strahlungsinduzierte Mutationen<br />

hergestellt und gezüchtet, weil sie den klimatologischen<br />

und geologischen Bedingungen in Peru oder auch in Mali<br />

besser angepasst sind. Am besten eignen sich strahlungsinduzierte<br />

Mutationen für die Optimierung von Pflanzen,<br />

die sich durch sterile Klonung vermehren, wie Bananen.<br />

Eine früh blühende, herrlich schmeckende und seuchenresistente<br />

Variante Novaria wurde auf diese Art in Malaysia<br />

entwickelt. Seuchenresistente Sorten bringen nicht nur<br />

bessere Ausbeute sondern benötigen auch weniger oder<br />

gar keinen Pestizideinsatz und sind deshalb sowohl billiger<br />

als auch umweltschonender. Bananen und Reis sind die<br />

wichtigsten Grundnahrungsmittel und Landwirtschaftsprodukte<br />

in der Dritten Welt, aber auch in der Industriestaaten<br />

werde fast zweitausend durch strahlungsinduzierte<br />

Mutationen verbesserte Produkte konsumiert.<br />

In Österreich wurde aus Golden Delicious die Variante<br />

„Golden Haidegg” gezüchtet, mit einer schöneren Farbe<br />

und längerer Haltbarkeit ohne Rostflecken. Japanische<br />

Birnen mit strahlungsinduzierter Resistenz gegen<br />

Schwarzfleckenseuche werden derzeit gezüchtet. Dies<br />

reduziert den früher notwendigen Pestizideinsatz auf ein<br />

Viertel.<br />

Nukleare Techniken werden nicht nur zur Verbesserung<br />

der Nahrungsmittel sondern auch zu deren Schutz<br />

und Konservierung verwendet. In 30 Ländern wird<br />

Bestrahlung von Nahrungsmitteln zur Gesundheits- und<br />

| Bestrahlung von Nahrungsmitteln.<br />

Qualitätssicherung genutzt sowie zur Einhaltung der<br />

Quaran täneverordnungen für bestimmte Nahrungsmittel,<br />

wie Fleisch, Früchte, Kräuter und getrocknete Gemüse.<br />

Seit Mitte 2000 wird die Bestrahlung von Hackfleisch in<br />

den Vereinigten Staaten aus mikrobiologischen Sicherheitsgründen<br />

eingesetzt. In mehr als 2000 Supermärkten<br />

werden diese Produkte ohne signifikante Vorbehalte der<br />

Verbraucher angeboten. Viele Länder, insbesondere in<br />

Asien, Lateinamerika, dem Mittler Osten und Afrika<br />

würden enorme Nutzen durch der Anwendung dieser<br />

Technologie zum Schutz der Nahrungsmittel erzielen.<br />

Die weltweit internationale Autorität für die Sicherheit<br />

der Nahrungsmittel, die Kodex Alimentarius Kommission,<br />

hat auch im Bereich der Strahlenbehandlung von Nahrungsmitteln<br />

Protokolle, Richtlinien und Empfehlungen<br />

veröffent licht.<br />

Im Landwirtschaftsbereich findet man auch die Anwendung<br />

von Radioisotopen durch den Radioimmunoassay<br />

bei Tieren für Hormonanalysen von Milch, Serum oder<br />

Plasma. Zweck ist die Optimierung der Planung und<br />

Diagnose von der Trächtigkeit, die frühzeitige Erkennung<br />

von Gesundheits- und Reproduktionsmängeln und die<br />

zeitgerechte Identifikation und Vorbeugung vor Seuchen,<br />

wie Rinderpest, Trypanosomosis, Brucellosis und Maulund-Klauen-Seuche.<br />

Nukleare Technologien und Isotopmarkierungen<br />

mit Stickstoff, Kohlenstoff und Phosphor<br />

finden auch verbreitet Anwendungen zur Untersuchung<br />

von biologischer Stickstofffixierung durch Bakterien, zur<br />

Optimierung der Anwendung von Düngemitteln nach<br />

Menge, Art und Jahreszeit in den verschiedensten Klimazonen,<br />

Bodentypen und Ernährungskulturen sowie zur<br />

Minimierung der ungenutzten Phosphat- und Stickstoffmengen<br />

in der Umwelt.<br />

| Insektensterilisierung: Schädlingsbekämpfung.<br />

Beim Ausrotten von Schädlingen werden Nukleare<br />

Technologien unter der Bezeichnung „Sterile Insekten-<br />

Technik“ verwendet. Diese Technik beinhaltet die<br />

großräumige Produktion und systematische Freisetzung<br />

von strahlungssterilisierten männlichen Fruchtfliegen,<br />

Tsetse-Fliegen, Holzwürmer usw. Freigelassen, verhalten<br />

sie sich auf ganz normale Weise, aber produzieren keinerlei<br />

Nachwuchs. Die schon geschwächte nächste Generation<br />

wird wieder mit neuen sterilen Männchen versorgt,<br />

bis sie lokal gezielt ausgerottet ist, ohne Pestizide, ohne<br />

Umwelt zerstörung und unter Erhalt des biologischen<br />

Gleich gewichtes. Insbesondere die Nachfrage und<br />

Produktionskapazität für sterilisierte Mittelmeerfruchtfliegen,<br />

notorische Schädlinge mit erheblichen Auswirkungen,<br />

sind im Lauf der letzten Dekade von einer auf<br />

drei Milliarden Fliegen pro Woche gestiegen. Am meisten<br />

Erfolg hatten die Ausrottungskampagnen in Mittel- und<br />

Special Topic | A Journey Through 50 Years AMNT<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

| Insektensterilisierung: Schädlingsbekämpfung.<br />

Lateinamerika, den Südlichen Vereinigten Staaten,<br />

Australien und Japan. Subtile genetisch eingebaute Unterschiede<br />

erlauben die Sterilisierung und Aussortierung der<br />

männlichen Insekten schon im Puppenstadium. Dementsprechend<br />

werden die Zuchtkosten gesenkt und die<br />

Effektivität wird erhöht. Auf diese Weise wird diese Technologie<br />

gleich einem selektiven Insektizid zur kompletten<br />

Ausrottung verwendet. Im Rahmen eines Projektes des<br />

IAEO Programms zur „Technischen Kooperation“ wurde in<br />

Tansania eine Produktionsanlage für Millionen sterilisierte<br />

Tsetse Fliegen errichtet, womit verschiedene afrikanische<br />

Regionen von diesen Schädlingen befreit werden konnten.<br />

Derzeitige Forschungsprojekte in diesem Bereich entwickeln<br />

weitere sterile Insekten-Techniken für Motten und<br />

Würmer. Die IAEO hat ein ehrgeiziges Projekt für die<br />

Ausrottung von Malaria Moskitos gestartet.<br />

Die wirtschaftlichen Erfolge dieser Technologie sind<br />

insgesamt herausragend. Die Bekämpfung von Fruchtfliegen<br />

in Mexiko mit einem Kapitaleinsatz von<br />

10 Millionen Dollar pro Jahr bewirkt einen Ertrag von<br />

einer Milliarde Dollar Wert an Zitrusfrüchten und Gemüse.<br />

Ähnliche Ergebnisse wurden in Chile erzielt, insbesondere<br />

da die Zitrusfrüchte dort während des nördlichen Winters<br />

produziert werden und somit einen maßgeblichen Einfluss<br />

auf den Export haben.<br />

5 Hydrologie<br />

Nukleare und Isotopentechniken werden auch in der<br />

Hydrologie und Klimatologie genutzt, wo sich aufgrund<br />

der natürlich entstandenen Fingerabdrücke der Isotopenzusammensetzung<br />

die Herkunft und das Alter der Wasservorräte<br />

nachweisen lässt. Isotopentechniken, basierend<br />

auf Messungen von stabilen und radioaktiven Wasserkomponenten,<br />

werden zur Modellierung von Wassersystemen<br />

verwendet, um die in vielen Regionen<br />

| Isotopentechniken in der Hydrologie: Wasservorräte erweitern.<br />

beschränkten und kostbaren Wasserressourcen optimal zu<br />

nutzen und ein lokales Gleichgewicht zwischen verbrauchten<br />

und erneuerbaren Wassermengen zu erzielen.<br />

Diese Methoden werden auch zur Identifizierung und zum<br />

Nachweis von unerwünschten Vermischungen der Wasserversorgung<br />

mit Abwässern verwendet, wie es in dichtbesiedelten<br />

Großmetropolen öfter vorkommt. Die<br />

Temperaturabhängigkeit der Wasserisotopenzusammensetzung<br />

beim Verdampfen und Kondensieren wird in<br />

weltweiten Forschungen von Klima- und Treibhauseffekt<br />

genutzt. So ist die Klimageschichte von zehntausenden<br />

Jahren aufgrund der Isotopenanalysen von Gletschereisschichten<br />

oder antarktischen Ablagerungen rekonstruiert<br />

worden.<br />

Beispiele zur Angabe der wirtschaftlichen und humanitären<br />

Bedeutung dieser nuklearen Techniken in der<br />

Hydrologie sind:<br />

p In Venezuela wurde das Trinkwasserdefizit um 30 %<br />

verringert durch Inbetriebnahme von 50 mit Isotopentechniken<br />

georteten Frischwasserquellen.<br />

p Durch Grundwassermessungen im Niltal und in den<br />

ägyptischen und äthiopischen Wüsten wurden erneuerbare<br />

Wasserreserven in Nubischen Sandsteinschichten<br />

geortet womit die Wasserversorgung um 20 % erhöht<br />

werden konnte.<br />

p Ebenfalls in Venezuela wurde aufgrund von Isotopenmessungen<br />

die Zuverlässigkeit eines Stauseedammes<br />

nachgewiesen. Eingeplante Reparaturkosten in<br />

Höhe von 6 Mio. US-Dollar konnten somit eingespart<br />

werden.<br />

p Mit dem Ausbau von geothermischen Energiequellen<br />

werden in Mittelamerika und Südostasien Millionen<br />

Dollar an Ölimporten gespart. Isotopentechniken<br />

dienen dabei zur Suche von geeigneten Standorten und<br />

zur Optimierung der Anlageneinrichtungen.<br />

6 Umweltschutz<br />

Nukleare Technologien und Messtechniken werden weit<br />

verbreitet zur Überwachung und Er<strong>for</strong>schung der Umwelt<br />

eingesetzt. Überhaupt werden künstliche und natürliche<br />

Radioisotope in der Atmosphäre, in der Hydrosphäre und<br />

im Boden für radiologische Zwecke überwacht. Sie sind<br />

zudem ausgezeichnete Indikatoren für atmosphärische<br />

und ozeanografische Transportprozesse, für das Verhalten<br />

und den Transport von nicht radioaktiven Umweltgiften,<br />

für die geochronologische Altersbestimmung und für<br />

ökologische und biologische Forschungszwecke. Einfache<br />

tragbare Röntgenfluoreszenzgeräte mit Radioisotopenquellen,<br />

Forschungsreaktoren für die Neutronenaktivierungsanalyse<br />

und Teilchenbeschleuniger zur<br />

Charakterisierung von mikroskopischen Teilchen <strong>for</strong>men<br />

zusammen ein Arsenal von nuklearen Techniken zum<br />

Identifizieren von schädliche Substanzen in der Umwelt<br />

und zur genauer Identifizierung ihrer Zusammensetzung<br />

und Herkunft.<br />

Nicht nur zur Diagnose von schädlichen Substanzen,<br />

sondern auch zur Vorbeugung oder Eliminierung ihrer<br />

Effekte werden nukleare Technologien verwendet.<br />

Elektronenbeschleuniger werden in Kohlekraftwerken<br />

verschiedener Länder zur Abgasbehandlung eingesetzt.<br />

Schwefel- und Stickstoffoxiden werden im Rauchgas<br />

umgewandelt und mit zugefügtem Ammoniak zu Düngemittel<br />

umgesetzt. Abwässer, organisch verseuchtes Wasser<br />

und Klärschlamm werden mit ionisierender Strahlung<br />

behandelt und desinfiziert.<br />

Zur Umweltsanierung gehört auch die Ortung und<br />

Entfernung von rund 60 Millionen Landminen, die in<br />

553<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

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| Lokalisierung von Landminen.<br />

62 Ländern Jahr für Jahr Ursache für Tausende Tote und<br />

Verletzte sind. Nukleare Technologien werden auch für<br />

diese Anwendungen entwickelt. Derzeit verfügbare Geräte<br />

benutzen Neutronengeneratoren und Gammadetektoren<br />

zur Auswertung von Elementverhältnissen, insbesondere<br />

Wasserstoff, Stickstoff, Kohlenstoff und Sauerstoff, in<br />

minenverdächtigen Gegenständen im Boden.<br />

7 Industrie und Archäologie<br />

Unzählige Anwendungen von nuklearen Technologien,<br />

ionisierender Strahlung und Isotopen werden auch in<br />

allgemeinen Bereichen der Industrie genutzt, so in<br />

Transport- und Bauunternehmen für radiografische<br />

Materialkontrolle, in der Polymerverarbeitung (z. B.<br />

Autoreifen), für die Sterilisierung sowie das Vernichten<br />

von Viren, Bakterien und Pilzen im medizinischen<br />

Bereich.<br />

Minuten quadratmillimeter kleine Einzelheiten von<br />

Gemälden, Skulpturen, Polychromen, Münzen oder<br />

Keramiken zu charakterisieren und auszuwerten. Ein<br />

Prototyp kam im Kunsthistorischen Museum in Wien für<br />

die Identifizierung von Gemälden des 16ten Jahrhunderts<br />

und zur Katalogisierung von Etruskischen Bronzen und<br />

Münzen zum Einsatz.<br />

8 Zusammenfassung<br />

Nukleare Technologien, Isotopentechniken und Strahlungsanwendungen<br />

sind weit verbreitete Anwendungen<br />

und liefern unersetzbare Beiträge zur Verbesserung und<br />

Erhaltung des heutigen Lebensstandards sowie zum<br />

Schutz des Menschen und seiner Umwelt. Das Ausmaß<br />

dieser Anwendungen für diagnostische und thera peutische<br />

Zwecke im Gesundheitsbereich wurde dargestellt und<br />

Erfolge und Erwartungen der Entwicklungen von<br />

Computertomografie und molekularer Nuklearmedizin<br />

wurden erwähnt. Im therapeutischen Bereich ist ohne<br />

Zweifel eine Erweiterung der Anwendungen zu erwarten,<br />

inbesondere in der Dritten Welt. Anwendungen in der<br />

Landwirtschaft, der Nahrungsversorgung und dem Schutz<br />

der Nahrung, wie Pflanzenoptimierung, Ausrottung von<br />

Schädlingen, Vorbeugung vor Seuchen und Konservierung<br />

von Nahrungsmittel sind vollständig ausgereift und<br />

werden akzeptiert. Bei komplexeren Problemen kommen<br />

ebenfalls erfolgreiche Maßnahmen zum Einsatz, wie<br />

im Zusammenhang mit der Bekämpfung von Malaria,<br />

Trypanosomosis und Myasis. Anwendungen in der Hydrologie<br />

und im Umweltschutz sind seit Jahren etabliert und<br />

anerkannt. Im Bereich der Klimatologie und in der meeresoder<br />

landesbezogenen Ökologie werden sie erweitert<br />

genutzt. Nukleare Techniken und Technologien in den<br />

hier beschriebenen Anwendungsbereichen werden nicht<br />

aus reinem Interesse an der Sache oder aufgrund<br />

popukärer Überlegungen eingesetzt, oder weil besondere<br />

Förderungen zu erwarten wären, sondern weil sie nach<br />

vergleichender Begutachtung und Kosten-Nutzen-Analyse<br />

optimalere Ergebnisse versprechen als vergleichbare<br />

verfügbare Technologien.<br />

Literatur<br />

1. Annual Report 2000. <strong>International</strong> Atomic Energy Agency, Vienna, 2001<br />

2. <strong>Nuclear</strong> Technology Review 2002. Document GOV/2002/7, <strong>International</strong> Atomic Energy Agency,<br />

Vienna, February 2002<br />

3. The <strong>International</strong> Atomic Energy Agency‘s Laboratories, Meeting the Challenges of Research and<br />

<strong>International</strong> Cooperation in the Application of <strong>Nuclear</strong> Techniques. Document IAEA/PI/A67E<br />

99-02059, August 1999<br />

| Kulturelles Erbe: Analysen zur Bestätigung der Herkunft.<br />

Interessant und vielleicht weniger bekannt ist die<br />

Anwendung von zerstörungsfreien nuklearen Technologien<br />

und Messmethoden im Bereich der Archäologie und<br />

der Konservierung des Kulturellen Erbes. Vor kurzer Zeit<br />

hat die IAEO ein Projekt in Lateinamerika zur Förderung<br />

von interdisziplinärer Forschung und Unterstützung in<br />

diesem Bereich koordiniert. Jedes teilnehmende Land<br />

hat sich mit bestimmten ungelösten Fragen bezüglich<br />

Keramiken befasst und mittels Spurenanalyse durch<br />

Röntgenfluoreszenz- und Neutronenaktivierung das<br />

erwünschte Ergebnis endgültig erhalten.<br />

In Zusammenarbeit mit polnischen, kubanischen und<br />

österreichischen Wissenschaftlern hat die IAEO ein tragbares<br />

Röntgenfluoreszenzgerät entwickelt, das ideal für<br />

die zerstörungsfreie Analyse von Kunstgegenstände<br />

eingesetzt werden kann. Es ist mit Computer- und Lasersteuerung<br />

ausgestattet und in der Lage, innerhalb von<br />

Verfasser<br />

Peter P. De Regge<br />

Head, PCI Laboratorien,<br />

<strong>International</strong>e Atomenergie Organisation (IAEO),<br />

Wagramer Straße 5, 1400 Wien, Austria<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Inside<br />

Women in <strong>Nuclear</strong> bei Framatome/ANF in Karlstein<br />

Die deutsche Sektion des international agierenden Frauennetzwerks<br />

WiN (Women in <strong>Nuclear</strong>) traf sich am 10. und<br />

<strong>11</strong>. Oktober 2019 bei Framatome in Karlstein. Nach der<br />

Mitgliederversammlung von WiN Germany wurde das<br />

Treffen mit Teilnehmerinnen aus Schweden, Finnland und<br />

der Schweiz international.<br />

Unter dem Dach des gemeinnützigen Vereins Women in<br />

<strong>Nuclear</strong> (WiN) Germany – gegründet 2008 – möchten<br />

Frauen, die auf den Gebieten Kernenergie, Strahlenschutz,<br />

Nuklearmedizin und nukleare Wissenschaften arbeiten,<br />

dazu beitragen, einen transparenten Dialog über die Notwendigkeit<br />

der nuklearen Kompetenzen für Deutschland<br />

zu führen.<br />

Chantal Greul, Präsidentin von WiN Germany, erklärt:<br />

„Auch wenn sich unser Land mehrheitlich für den Ausstieg<br />

aus der Stromerzeugung durch Kerntechnik entschieden<br />

hat, bleiben das Know-how und das hohe Sicherheitsverständnis<br />

Deutschlands nach dem Ausstieg 2022<br />

wichtig: für den Rückbau, die Endlagerung – und nicht<br />

zuletzt als Beitrag zum Erhalt des höchstmöglichen Sicherheitsniveaus<br />

weltweit. Wir ermutigen junge Frauen, eine<br />

berufliche Laufbahn in der Nukleartechnik zu wählen.<br />

Denn nur durch gut ausgebildete Nachwuchskräfte bleiben<br />

die notwendigen Ressourcen in Deutschland langfristig<br />

gesichert.“<br />

| Im Foyer der Framatome GmbH in Karlstein mit den Standortleitern<br />

Stefan Rosenberger der Framatome GmbH (links) und Matthias Gutjahr<br />

der Advanced <strong>Nuclear</strong> Fuels GmbH (Mitte).<br />

| v.l. Chantal Greul, WiN Germany Präsidentin,<br />

Preisträgerin Jenny Jessat, Martina Ezmuß,<br />

Vorstand WiN Germany.<br />

Einen wichtigen Beitrag zu diesem Vereinsziel leistet<br />

die jährliche Verleihung des WiN-Germany Preises. Für<br />

diesen mit 500 Euro dotierten Preis gab es in diesem Jahr<br />

gleich fünf Bewerberinnen. Beiträge aus Wissenschaft und<br />

Forschung sowie aus Unternehmen ließen eine hohe<br />

fachliche Kompetenz erkennen, die die Entscheidung für<br />

eine Bewerberin schwer fallen ließ. Jenny Jessat überzeugte<br />

den Kreis mit dem Thema „Studies on the interaction<br />

of plant cells with uranium (VI) and europium (III)<br />

an on stress-induced metabolite release” („Studien zur<br />

Wechselwirkung von Pflanzenzellen mit Uran (VI) und<br />

Europium (III) und zur stressinduzierten Metabolitfreisetzung“).<br />

Es handelte sich um die Präsentation<br />

ihrer Masterarbeit in Chemie an der TU Dresden. Jenny<br />

Jessat ist derzeit wissenschaftliche Mitarbeiterin am<br />

Helmholtz-Zentrum Dresden-Rossendorf, Institut für<br />

Ressourcenökologie und bereitet dort ihre Promotion vor.<br />

Neben weiteren Fachvorträgen über kerntechnische<br />

Themen fanden die Werksführungen sehr viel Anklang bei<br />

den Teilnehmerinnen. „Wir möchten uns herzlich bei den<br />

beiden Standortleitern der Framatome GmbH und ANF<br />

bedanken. Insbesondere die Standortbesuche waren ein<br />

wirkliches Highlight und die Teilnehmerinnen waren<br />

damit sehr zufrieden“, lobte Chantal Greul den Einsatz der<br />

Framatome für die Veranstaltung.“<br />

| Aus der Komponentenfertigung der ANF<br />

Karlstein stellte Petra Denner die<br />

Entwicklungs historie des neuen GAIA<br />

Abstandhalters vor und zeigte an diesem<br />

Beispiel auf, wie die hohen Qualitätsan<strong>for</strong>derungen<br />

der Kunden erfüllt werden.<br />

Karin Reiche<br />

WiN<br />

KTG Inside<br />

Verantwortlich<br />

für den Inhalt:<br />

Die Autoren.<br />

Lektorat:<br />

Natalija Cobanov,<br />

Kerntechnische<br />

Gesellschaft e. V.<br />

(KTG)<br />

Robert-Koch-Platz 4<br />

10<strong>11</strong>5 Berlin<br />

T: +49 30 498555-50<br />

F: +49 30 498555-51<br />

E-Mail:<br />

natalija.cobanov@<br />

ktg.org<br />

555<br />

KTG INSIDE<br />

Herzlichen Glückwunsch!<br />

www.ktg.org<br />

Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag<br />

und wünscht ihnen weiterhin alles Gute!<br />

Dezember 2019<br />

50 Jahre | 1969<br />

13. Bernd Gulich, Tiefenbach-Ast<br />

55 Jahre | 1964<br />

30. Thomas Schmidt, Lörrach<br />

60 Jahre | 1959<br />

15. Axel Lenzen, Titz<br />

20. Martin Schlieck-Weber, Hausen<br />

65 Jahre | 1954<br />

24. Reinhold Paul, Hanau<br />

70 Jahre | 1949<br />

2. Dipl.-Ing. Berndt Standfuß, Dresden<br />

28. Fritz Grimm, Alzenau<br />

76 Jahre | 1943<br />

8. Dr. Dieter Herrmann, Brandis<br />

77 Jahre | 1942<br />

8. Karl Georg Weber, Neckarwestheim<br />

14. Günter Breiling, Weinheim<br />

78 Jahre | 1941<br />

13. Dipl.-Ing. Klaus-Dieter Hnilica,<br />

Rodenbach/Hanau<br />

Wenn Sie künftig eine<br />

Erwähnung Ihres<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

556<br />

79 Jahre | 1940<br />

8. Dipl.-Ing. Wolfgang Heess, Laudenbach<br />

21. Dr. Jürgen Wehmeier, Springe<br />

94 Jahre | 1925<br />

10. Dr. Arthur Pilgenröther, Kleinostheim<br />

83 Jahre | 1937<br />

9. Dipl.-Ing. Werner Rossbach,<br />

Bergisch Gladbach<br />

NEWS<br />

79 Jahre | 1940<br />

16. Dipl.-Ing. Wolfgang Breyer, Buckenhof<br />

19 Prof. Dr. Wernt Brewitz, Wolfenbüttel<br />

80 Jahre |1939<br />

1. Dipl.-Ing. Georg Dumsky, Gräfelfing<br />

6. Dipl.-Ing. Hans-Henn. Kuchenbuch,<br />

Laboe-Brodersdorf<br />

27 Dr. Horst Bauer, Sigless/AT<br />

81 Jahre | 1938<br />

1. Dr. Gert Spannagel,<br />

Linkenheim-Hochstetten<br />

82 Jahre | 1937<br />

30. Dipl.-Ing. Wilhelm Weiss, Weinheim<br />

83 Jahre | 1936<br />

7. Dipl.-Ing. Aurel Badics, Bad Kreuznach<br />

17. Prof. Dr.-Ing. Rolf Theenhaus, Linnich<br />

86 Jahre | 1933<br />

10. Prof. Dr. Jürgen Vollradt,<br />

Unna-Königsborn<br />

Januar 2020<br />

55 Jahre | 1965<br />

31. Eckhard Stengert, Worms<br />

60 Jahre | 1960<br />

26. Dr. Friedhelm Funke, Dormitz<br />

70 Jahre | 1950<br />

15. Dipl.-Ing. Andreas Hüttmann, Oering<br />

78 Jahre | 1942<br />

31. Dipl.-Phys. Werner Scholtyssek,<br />

Stutensee<br />

79 Jahre | 1941<br />

15. Dipl.-Ing. Ulf Rösser,<br />

Heiligkreuzsteinach<br />

81 Jahre | 1939<br />

16. Dr. Wolfgang Kersting, Blieskastel<br />

82 Jahre | 1938<br />

7. Dipl.-Ing. Manfred Schirra, Stutensee<br />

12. Dipl.-Ing. Hans Dieter Adami, Rösrath<br />

22. Dr. Franz Müller, Erlangen<br />

84 Jahre | 1936<br />

5. Obering. Peter Vetterlein, Oberursel<br />

23. Prof. Dr. Hartmut Schmoock,<br />

Norderstedt<br />

30. Dipl.-Phys. Wolfgang Borkowetz,<br />

Rüsselsheim<br />

30. Dipl.-Ing. Friedrich Morgenstern, Essen<br />

85 Jahre |1935<br />

10. Dipl.-Ing. Walter Diefenbacher,<br />

Karlsruhe<br />

17. Dipl.-Ing. Helge Dyroff, Alzenau<br />

24. Theodor Himmel, Bad Honnef<br />

87 Jahre |1933<br />

9. Prof. Dr. Hellmut Wagner, Karlsruhe<br />

88 Jahre | 1932<br />

3. Dipl.-Ing. Fritz Kohlhaas, Kahl/Main<br />

91 Jahre | 1929<br />

20. Dr. Devana Lavrencic-Cannata, Rom/I<br />

93 Jahre | 1927<br />

1. Prof. Dr. Werner Oldekop,<br />

Braunschweig<br />

Top<br />

World Energy Outlook 2019<br />

highlights deep disparities in<br />

the global energy system<br />

(iea) Deep disparities define today’s<br />

energy world. The dissonance<br />

between well-supplied oil markets<br />

and growing geopolitical tensions and<br />

uncertainties. The gap between the<br />

ever-higher amounts of greenhouse<br />

gas emissions being produced and the<br />

insufficiency of stated policies to<br />

curb those emissions in line with<br />

international climate targets. The gap<br />

between the promise of energy <strong>for</strong><br />

all and the lack of electricity access<br />

<strong>for</strong> 850 million people around the<br />

world.<br />

The World Energy Outlook 2019,<br />

the <strong>International</strong> Energy Agency’s<br />

flagship publication, explores these<br />

widening fractures in detail. It explains<br />

the impact of today’s decisions<br />

on tomorrow’s energy systems, and<br />

describes a pathway that enables the<br />

world to meet climate, energy access<br />

and air quality goals while maintaining<br />

a strong focus on the reliability<br />

and af<strong>for</strong>dability of energy <strong>for</strong> a<br />

growing global population.<br />

As ever, decisions made by<br />

governments remain critical <strong>for</strong> the<br />

future of the energy system. This is<br />

evident in the divergences between<br />

WEO scenarios that map out different<br />

routes the world could follow over the<br />

coming decades, depending on the<br />

policies, investments, technologies<br />

and other choices that decision<br />

makers pursue today. Together, these<br />

scenarios seek to address a fun damental<br />

issue – how to get from where<br />

we are now to where we want to go.<br />

The path the world is on right now<br />

is shown by the Current Policies<br />

Scenario, which provides a baseline<br />

picture of how global energy systems<br />

would evolve if governments make no<br />

changes to their existing policies. In<br />

this scenario, energy demand rises by<br />

1.3 % a year to 2040, resulting in<br />

strains across all aspects of energy<br />

markets and a continued strong<br />

upward march in energy-related<br />

emissions.<br />

The Stated Policies Scenario,<br />

<strong>for</strong>merly known as the New Policies<br />

Scenario, incorporates today’s policy<br />

intentions and targets in addition to<br />

existing measures. The aim is to hold<br />

up a mirror to today’s plans and<br />

illustrate their consequences. The<br />

future outlined in this scenario is still<br />

well off track from the aim of a secure<br />

and sustainable energy future. It<br />

describes a world in 2040 where<br />

hundreds of millions of people still go<br />

without access to electricity, where<br />

pollution-related premature deaths<br />

remain around today’s elevated levels,<br />

and where CO 2 emissions would<br />

lock in severe impacts from climate<br />

change.<br />

The Sustainable Development<br />

Scenario indicates what needs to be<br />

done differently to fully achieve<br />

climate and other energy goals<br />

that policy makers around the world<br />

have set themselves. Achieving this<br />

scenario – a path fully aligned with<br />

the Paris Agreement aim of holding<br />

the rise in global temperatures to well<br />

below 2° C and pursuing ef<strong>for</strong>ts to<br />

limit it to 1.5° C – requires rapid and<br />

widespread changes across all parts of<br />

the energy system. Sharp emission<br />

cuts are achieved thanks to multiple<br />

fuels and technologies providing<br />

efficient and cost-effective energy<br />

services <strong>for</strong> all.<br />

“What comes through with crystal<br />

clarity in this year’s World Energy<br />

News


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Outlook is there is no single or simple<br />

solution to trans<strong>for</strong>ming global energy<br />

systems,” said Dr Fatih Birol, the IEA’s<br />

Executive Director. “Many technologies<br />

and fuels have a part to play<br />

across all sectors of the economy. For<br />

this to happen, we need strong leadership<br />

from policy makers, as governments<br />

hold the clearest responsibility<br />

to act and have the greatest scope to<br />

shape the future.”<br />

In the Stated Policies Scenario,<br />

energy demand increases by 1% per<br />

year to 2040. Low-carbon sources, led<br />

by solar PV, supply more than half of<br />

this growth, and natural gas accounts<br />

<strong>for</strong> another third. Oil demand flattens<br />

out in the 2030s, and coal use edges<br />

lower. Some parts of the energy sector,<br />

led by electricity, undergo rapid<br />

trans<strong>for</strong>mations. Some countries,<br />

notably those with “net zero”<br />

aspirations, go far in reshaping all<br />

aspects of their supply and consumption.<br />

However, the momentum behind<br />

clean energy is insufficient to offset<br />

the effects of an expanding global<br />

economy and growing population.<br />

The rise in emissions slows but does<br />

not peak be<strong>for</strong>e 2040.<br />

Shale output from the United<br />

States is set to stay higher <strong>for</strong> longer<br />

than previously projected, reshaping<br />

global markets, trade flows and<br />

security. In the Stated Policies<br />

Scenario, annual US production<br />

growth slows from the breakneck<br />

pace seen in recent years, but the<br />

United States still accounts <strong>for</strong> 85 % of<br />

the increase in global oil production to<br />

2030, and <strong>for</strong> 30 % of the increase in<br />

gas. By 2025, total US shale output<br />

(oil and gas) overtakes total oil and<br />

gas production from Russia.<br />

“The shale revolution highlights<br />

that rapid change in the energy<br />

system is possible when an initial push<br />

to develop new technologies is<br />

complemented by strong market<br />

incentives and large-scale investment,”<br />

said Dr Birol. “The effects<br />

have been striking, with US shale now<br />

acting as a strong counterweight to<br />

ef<strong>for</strong>ts to manage oil markets.”<br />

The higher US output pushes down<br />

the share of OPEC members and<br />

Russia in total oil production, which<br />

drops to 47 % in 2030, from 55 % in<br />

the mid-2000s. But whichever pathway<br />

the energy system follows, the<br />

world is set to rely heavily on oil supply<br />

from the Middle East <strong>for</strong> years to<br />

come.<br />

Alongside the immense task of<br />

putting emissions on a sustainable<br />

trajectory, energy security remains<br />

paramount <strong>for</strong> governments around<br />

the globe. Traditional risks have not<br />

gone away, and new hazards such as<br />

cybersecurity and extreme weather<br />

require constant vigilance. Meanwhile,<br />

the continued trans<strong>for</strong>mation<br />

of the electricity sector requires policy<br />

makers to move fast to keep pace with<br />

technological change and the rising<br />

need <strong>for</strong> the flexible operation of<br />

power systems.<br />

“The world urgently needs to put a<br />

laser-like focus on bringing down<br />

global emissions. This calls <strong>for</strong> a grand<br />

coalition encompassing governments,<br />

investors, companies and everyone<br />

else who is committed to tackling<br />

climate change,” said Dr Birol. “Our<br />

Sustainable Development Scenario<br />

is tailor-made to help guide the<br />

members of such a coalition in their<br />

ef<strong>for</strong>ts to address the massive climate<br />

challenge that faces us all.”<br />

A sharp pick-up in energy efficiency<br />

improvements is the element that<br />

does the most to bring the world<br />

towards the Sustainable Development<br />

Scenario. Right now, efficiency<br />

improvements are slowing: the 1.2 %<br />

rate in 2018 is around half the average<br />

seen since 2010 and remains far below<br />

the 3 % rate that would be needed.<br />

Electricity is one of the few energy<br />

sources that sees rising consumption<br />

over the next two decades in the<br />

Sustainable Development Scenario.<br />

Electricity’s share of final consumption<br />

overtakes that of oil, today’s<br />

leader, by 2040. Wind and solar PV<br />

provide almost all the increase in<br />

electricity generation.<br />

Putting electricity systems on a<br />

sustainable path will require more<br />

than just adding more renewables.<br />

The world also needs to focus on<br />

the emissions that are “locked in”<br />

to existing systems. Over the past<br />

20 years, Asia has accounted <strong>for</strong><br />

90 % of all coal-fired capacity built<br />

worldwide, and these plants potentially<br />

have long operational lifetimes<br />

ahead of them. This year’s WEO considers<br />

three options to bring down<br />

emissions from the existing global<br />

coal fleet: to retrofit plants with carbon<br />

capture, utilisation and storage or<br />

biomass co-firing equipment; to<br />

repurpose them to focus on providing<br />

system adequacy and flexibility; or to<br />

retire them earlier.<br />

| www.iea.org<br />

New report and webinar<br />

on the supply of medical<br />

radioisotopes<br />

(nea) The NEA hosted a webinar on<br />

18 November 2019 to present findings<br />

from a new report on the supply of<br />

medical radioisotopes, jointly produced<br />

with the Organisation <strong>for</strong><br />

Economic Co-operation and Development<br />

(OECD) Health Committee.<br />

Technetium-99m (Tc-99m) is the<br />

most commonly used medical<br />

radioisotope, essential <strong>for</strong> 85 % of the<br />

nuclear medicine diagnostic scans<br />

per<strong>for</strong>med worldwide. There are no<br />

comparable substitutes available <strong>for</strong><br />

diagnoses of various cancers and <strong>for</strong> a<br />

range of diagnostics in children.<br />

Un<strong>for</strong>tunately, the global supply of<br />

Tc-99m is not technically and economically<br />

robust, and the existing<br />

supply-chain continues to experience<br />

chronic shortages. This new study<br />

analyses the current market structure<br />

and identifies barriers <strong>for</strong> the implementation<br />

of full cost recovery.<br />

Report and the webinar recording<br />

are available at: <br />

oe.cd/nea-med-rad-webinar-2019.<br />

| www.oecd-nea.org<br />

World <strong>Nuclear</strong> Per<strong>for</strong>mance<br />

Report 2019 Asia Edition<br />

launched<br />

(wna) The World <strong>Nuclear</strong> Per<strong>for</strong>mance<br />

Report 2019 Asia Edition shows<br />

that nuclear energy in Asia is meeting<br />

the growing demand <strong>for</strong> electricity,<br />

whilst helping to tackle<br />

air pollution and climate change. The<br />

report, published by World <strong>Nuclear</strong> Association,<br />

was launched today at Singapore<br />

<strong>International</strong> Energy<br />

Week.<strong>Nuclear</strong> generation in Asia<br />

continued its rapid growth in 2018, increasing<br />

by 12 %. By replacing<br />

coal-fired generation nuclear energy<br />

avoided the emission of over<br />

500 million tonnes of carbon dioxide<br />

last year.Agneta Rising, Director<br />

General of World <strong>Nuclear</strong> Association<br />

said, “<strong>Nuclear</strong> is fast, scalable and a<br />

long-lasting way to add clean<br />

557<br />

NEWS<br />

News


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Operating Results July 2019<br />

558<br />

NEWS<br />

Plant name Country Nominal<br />

capacity<br />

Type<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Month Year Since<br />

commissioning<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Month Year Month Year<br />

OL1 Olkiluoto 1) BWR FI 910 880 744 681 827 4 427 294 266 082 502 100.00 95.66 100.00 94.49 99.61 94.60<br />

OL2 Olkiluoto BWR FI 910 880 744 678 000 4 090 942 255 987 484 100.00 88.26 99.94 87.58 99.05 87.41<br />

KCB Borssele 3) PWR NL 512 484 662 326 759 4 445 853 166 167 541 88.61 80.62 88.66 80.51 85.71 76.95<br />

KKB 1 Beznau 7) PWR CH 380 365 744 265 244 1 583 283 128 917 393 100.00 82.35 100.00 82.10 93.63 81.77<br />

KKB 2 Beznau 7) PWR CH 380 365 744 266 954 1 925 0<strong>11</strong> 136 275 418 100.00 100.00 100.00 100.00 94.22 99.56<br />

KKG Gösgen 3,7) PWR CH 1060 1010 6<strong>11</strong> 634 890 4 627 036 318 502 564 82.06 86.72 82.03 86.04 80.50 85.81<br />

KKM Mühleberg BWR CH 390 373 744 274 120 1 942 150 129 346 465 100.00 100.00 99.89 99.73 94.47 97.89<br />

CNT-I Trillo PWR ES 1066 1003 744 784 619 4 605 674 251 897 342 100.00 86.01 100.00 85.50 98.03 84.41<br />

Dukovany B1 PWR CZ 500 473 744 363 425 2 485 594 <strong>11</strong>4 715 087 100.00 99.86 99.31 99.61 97.70 97.72<br />

Dukovany B2 1,2) PWR CZ 500 473 0 0 1 610 472 109 844 643 0 65.17 0 64.65 0 63.32<br />

Dukovany B3 PWR CZ 500 473 744 357 397 1 966 445 108 464 485 100.00 79.26 100.00 78.81 96.07 77.31<br />

Dukovany B4 PWR CZ 500 473 744 365 370 2 533 257 108 976 525 100.00 100.00 100.00 99.92 98.22 99.60<br />

Temelin B1 PWR CZ 1080 1030 744 798 674 3 926 675 <strong>11</strong>8 287 717 100.00 72.28 99.96 71.97 99.21 71.34<br />

Temelin B2 1,2) PWR CZ 1080 1030 0 0 4 477 198 <strong>11</strong>3 749 715 0 80.87 0 80.81 0 81.34<br />

Doel 1 PWR BE 454 433 744 340 008 1 582 284 137 026 746 100.00 67.09 99.98 66.66 97.84 67.15<br />

Doel 2 PWR BE 454 433 744 334 234 1 915 142 135 717 081 100.00 83.89 99.29 82.19 98.53 82.55<br />

Doel 3 2) PWR BE 1056 1006 181 144 566 4 063 502 259 195 987 24.37 75.48 20.70 74.72 18.01 75.13<br />

Doel 4 PWR BE 1084 1033 744 778 346 5 427 730 265 801 140 100.00 100.00 99.94 97.96 94.76 96.96<br />

Tihange 1 PWR BE 1009 962 744 725 978 5 120 495 303 951 352 100.00 100.00 99.99 99.99 96.68 99.91<br />

Tihange 2 2) PWR BE 1055 1008 693 659 639 659 639 255 3<strong>11</strong> 569 93.13 13.62 87.24 12.76 84.43 12.35<br />

Tihange 3 PWR BE 1089 1038 744 780 007 5 404 413 276 631 686 100.00 99.96 100.00 99.02 96.67 98.05<br />

Plant name<br />

Type<br />

Nominal<br />

capacity<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Since Month Year Month Year Month Year<br />

commissioning<br />

KBR Brokdorf 1,2) DWR 1480 1410 544 722 179 5 437 455 356 005 265 73.08 81.<strong>11</strong> 67.73 76.05 65.25 71.95<br />

KKE Emsland DWR 1406 1335 744 1 033 507 5 701 515 352 520 484 100.00 81.41 100.00 81.27 98.81 79.71<br />

KWG Grohnde DWR 1430 1360 744 1 000 783 5 717 420 383 291 634 100.00 82.88 99.55 82.50 93.39 78.10<br />

KRB C Gundremmingen SWR 1344 1288 744 983 969 5 481 876 336 423 631 100.00 81.32 100.00 80.60 97.93 79.74<br />

KKI-2 Isar 1,2) DWR 1485 1410 389 518 222 6 788 871 360 514 684 52.28 93.02 50.22 92.71 46.55 89.53<br />

GKN-II Neckarwestheim 4) DWR 1400 1310 744 910 310 6 858 810 336 685 644 100.00 100.00 100.00 99.88 86.99 96.47<br />

KKP-2 Philippsburg 1,2) DWR 1468 1402 448 580 690 6 339 374 372 500 529 60.22 88.81 59.74 88.52 52.23 83.60<br />

*)<br />

Net-based values<br />

(Czech and Swiss<br />

nuclear power<br />

plants gross-based)<br />

1)<br />

Refueling<br />

2)<br />

Inspection<br />

3)<br />

Repair<br />

4)<br />

Stretch-out-operation<br />

5)<br />

Stretch-in-operation<br />

6)<br />

Hereof traction supply<br />

7)<br />

Incl. steam supply<br />

8)<br />

New nominal<br />

capacity since<br />

January 2016<br />

9)<br />

Data <strong>for</strong> the Leibstadt<br />

(CH) NPP will<br />

be published in a<br />

further issue of <strong>atw</strong><br />

BWR: Boiling<br />

Water Reactor<br />

PWR: Pressurised<br />

Water Reactor<br />

Source: VGB<br />

elec tricity generation.“For the 90<br />

reactors that have started operating<br />

from 2000 to today, the typical construction<br />

time is 5 to 7 years. Of those<br />

90 reactors, 27 % were built in less<br />

than five years – and they will provide<br />

clean and reliable electricity <strong>for</strong> more<br />

than 60 years or more.”Many reactors<br />

in operation today are planned to<br />

operate <strong>for</strong> 60-80 years. Reactors are<br />

already demonstrating high per<strong>for</strong>mance<br />

irrespective of how long<br />

they have been in operation, with<br />

capacity factors of around 80 %<br />

maintained egardless of age.The<br />

report profiles Tarapur 1, a reactor<br />

located in Palghar, India, which<br />

marked 50 years of operation in<br />

April 2019. Four other reactors will<br />

match this achievement in 2019,<br />

the first year in which reactors have<br />

passed this milestone. Worldwide<br />

nuclear generation in 2018 increased<br />

<strong>for</strong> the sixth successive year, reaching<br />

2563 TWh. This is more than<br />

10 % of global electricity demand.<br />

Overall, capacity additions <strong>for</strong> the<br />

period 2016-2020 are expected to<br />

reach the targets of the nuclear<br />

industry’s Harmony programme.<br />

But build rates will have to increase<br />

significantly to achieve the overall goal<br />

of supplying 25 % of global electricity<br />

demand be<strong>for</strong>e 2050. Agneta Rising<br />

said, “<strong>Nuclear</strong> energy is key to Asia<br />

meeting the twin challenges of a growing<br />

demand <strong>for</strong> electricity, and an<br />

urgent need to switch to less polluting,<br />

low-carbon generation sources. More<br />

and more organizations are recognizing<br />

that nuclear energy is vital to the<br />

goal of a sustainable future <strong>for</strong> people<br />

and the planet.”<br />

| www.world-nuclear.org<br />

World Science Day:<br />

<strong>International</strong> <strong>Nuclear</strong><br />

In<strong>for</strong>mation System<br />

highlights open access<br />

(iaea) Public awareness of developments<br />

in scientific research is crucial<br />

<strong>for</strong> building a more in<strong>for</strong>med global<br />

society, and the IAEA is providing<br />

access to a trove of research on the<br />

peaceful uses of nuclear science and<br />

technology.<br />

The IAEA’s <strong>International</strong> <strong>Nuclear</strong><br />

In<strong>for</strong>mation System (INIS) is one of the<br />

world’s largest repositories of published<br />

research on the peaceful uses of nuclear<br />

science and technology, with more<br />

than 4.2 million bibliographic records<br />

and access to more than 1.6 million<br />

full-text documents. Each year, more<br />

than 1 million unique users per<strong>for</strong>m<br />

around 2 million searches and download<br />

3 million pages.<br />

News


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

Operating Results August 2019<br />

Plant name Country Nominal<br />

capacity<br />

Type<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Month Year Since<br />

commissioning<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Month Year Month Year<br />

OL1 Olkiluoto 1) BWR FI 910 880 744 680 095 5 107 389 266 762 596 100.00 96.22 99.94 95.18 99.36 95.21<br />

OL2 Olkiluoto BWR FI 910 880 744 678 578 4 769 520 256 666 062 100.00 89.76 100.00 89.16 99.14 88.91<br />

KCB Borssele PWR NL 512 484 693 336 333 4 782 186 166 503 874 91.08 81.95 91.06 81.86 88.20 78.39<br />

KKB 1 Beznau 7) PWR CH 380 365 733 272 559 1 855 842 129 189 952 100.00 84.60 100.00 84.38 96.27 83.62<br />

KKB 2 Beznau 1,2,7) PWR CH 380 365 239 86 668 2 0<strong>11</strong> 679 136 362 086 32.12 91.34 31.92 91.31 30.14 90.70<br />

KKG Gösgen 7) PWR CH 1060 1010 495 5<strong>11</strong> 876 5 138 912 319 014 440 66.47 84.14 66.04 83.49 64.91 83.14<br />

KKM Mühleberg BWR CH 390 373 744 280 420 2 222 570 129 626 885 100.00 100.00 99.93 99.75 96.64 97.73<br />

CNT-I Trillo PWR ES 1066 1003 744 785 304 5 390 978 252 682 646 100.00 87.79 99.87 87.33 98.16 86.16<br />

Dukovany B1 3) PWR CZ 500 473 379 176 941 2 662 535 <strong>11</strong>4 892 028 50.94 93.62 50.34 93.32 47.57 91.32<br />

Dukovany B2 2) PWR CZ 500 473 0 0 1 610 472 109 844 643 0 56.85 0.10 56.39 0 55.24<br />

Dukovany B3 PWR CZ 500 473 744 355 841 2 322 286 108 820 326 100.00 81.91 99.95 81.50 95.66 79.65<br />

Dukovany B4 PWR CZ 500 473 733 355 048 2 888 305 109 331 573 98.52 99.81 97.61 99.62 95.44 99.07<br />

Temelin B1 PWR CZ 1080 1030 744 793 753 4 720 428 <strong>11</strong>9 081 470 100.00 75.82 99.86 75.52 98.60 74.82<br />

Temelin B2 1) PWR CZ 1080 1030 5<strong>11</strong> 537 389 5 014 587 <strong>11</strong>4 287 104 68.68 79.32 66.85 79.03 66.76 79.48<br />

Doel 1 PWR BE 454 433 744 341 195 1 923 479 137 367 941 100.00 71.29 99.97 70.94 98.34 71.16<br />

Doel 2 PWR BE 454 433 744 330 999 2 246 141 136 048 080 100.00 85.95 99.28 84.37 97.52 84.46<br />

Doel 3 PWR BE 1056 1006 744 775 861 4 839 363 259 971 848 100.00 78.61 99.74 77.91 98.19 78.07<br />

Doel 4 PWR BE 1084 1033 744 696 829 6 124 559 266 497 969 100.00 100.00 87.42 96.62 84.39 95.36<br />

Tihange 1 PWR BE 1009 962 744 725 185 5 845 679 304 676 537 100.00 100.00 100.00 99.99 96.57 99.48<br />

Tihange 2 PWR BE 1055 1008 744 756 385 1 416 023 256 067 953 100.00 24.64 99.92 23.88 97.<strong>11</strong> 23.16<br />

Tihange 3 PWR BE 1089 1038 744 778 719 6 183 131 277 410 404 100.00 99.96 100.00 99.14 96.50 97.86<br />

559<br />

NEWS<br />

Plant name<br />

Type<br />

Nominal<br />

capacity<br />

gross<br />

[MW]<br />

net<br />

[MW]<br />

Operating<br />

time<br />

generator<br />

[h]<br />

Energy generated, gross<br />

[MWh]<br />

Time availability<br />

[%]<br />

Energy availability<br />

[%] *) Energy utilisation<br />

[%] *)<br />

Month Year Since Month Year Month Year Month Year<br />

commissioning<br />

KBR Brokdorf DWR 1480 1410 744 994 792 6 432 247 357 000 057 100.00 83.52 94.05 78.35 89.88 74.24<br />

KKE Emsland DWR 1406 1335 744 1 026 542 6 728 057 353 547 026 100.00 83.78 100.00 83.66 98.<strong>11</strong> 82.06<br />

KWG Grohnde DWR 1430 1360 744 990 908 6 708 328 384 282 542 100.00 85.06 100.00 84.73 92.41 79.93<br />

KRB C Gundremmingen SWR 1344 1288 744 978 185 6 460 061 337 401 816 100.00 83.70 99.40 83.00 97.38 81.99<br />

KKI-2 Isar 4) DWR 1485 1410 744 1 031 644 7 820 515 361 546 328 100.00 93.91 98.93 93.51 92.84 89.95<br />

GKN-II Neckarwestheim 1,2,4) DWR 1400 1310 207 223 200 7 082 010 336 908 844 100.00 100.00 21.19 89.84 21.19 86.86<br />

KKP-2 Philippsburg 1,2,5) DWR 1468 1402 409, 536 941 6 876 315 373 037 470 55.01 84.50 54.82 84.22 48.33 79.10<br />

In line with the theme of this year’s<br />

World Science Day <strong>for</strong> Peace and<br />

Development – “Open Science, leaving<br />

no one behind” – on 10 November,<br />

INIS plays a vital role in providing<br />

open access to in<strong>for</strong>mation and<br />

research.<br />

“The IAEA has been an advocate of<br />

open access to knowledge since the<br />

inception of INIS in 1970,” said<br />

Dobrica Savic, Head of the <strong>Nuclear</strong><br />

In<strong>for</strong>mation Section. All the documents,<br />

presentations and records are<br />

free of charge and accessible to anyone<br />

via the Internet.<br />

Open science is defined by<br />

UNESCO as making scientific research<br />

and data accessible to all. It includes<br />

publishing open scientific research,<br />

promoting open access and making<br />

open source software available.<br />

Libraries and repositories are key <strong>for</strong><br />

open science as many play an active<br />

role in the preservation, curation,<br />

publication and dissemination of<br />

digital scientific materials, according<br />

to the Organisation <strong>for</strong> Economic<br />

Co-operation and Development.<br />

The INIS repository includes<br />

research on a wide range of nuclearrelated<br />

topics, from environmental science<br />

and energy storage to radio logy<br />

and many more. It’s made possible<br />

through an ongoing colla borative<br />

ef<strong>for</strong>t between the IAEA and experts<br />

from 132 countries and 24 international<br />

organizations. Member States<br />

submit peer-reviewed research from<br />

their country/organization to INIS,<br />

and approximately 120,000 bibliographic<br />

records and 13,000 full-text<br />

PDF documents are added each year.<br />

To be sure, most publications are<br />

not <strong>for</strong> the lay reader. A quick glance<br />

turns up titles such as “Method and<br />

apparatus <strong>for</strong> the nondestructive<br />

assay of bulk nuclear fuel using 1 keV<br />

to Mev range neutrons” or “Regional<br />

comparison of nuclear and fossil electric<br />

power generation costs”. Many<br />

were once only available in hard copy<br />

but are now online after a massive<br />

ef<strong>for</strong>t to convert millions of microfiche<br />

pages to fully searchable electronic<br />

files.<br />

Improving INIS inputs<br />

To continually improve the quantity<br />

and quality of submissions to the INIS<br />

repository, the IAEA periodically<br />

offers trainings and e-learning courses<br />

to specialists, librarians and scientific<br />

officers involved in knowledge<br />

management. Last month, experts<br />

gathered in Vienna to discuss different<br />

aspects of INIS operations, such as<br />

News


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

560<br />

NEWS<br />

selection criteria and descriptive<br />

cataloguing, and improve their skills<br />

<strong>for</strong> the preparation of high-quality<br />

input and use of the repository. The<br />

participants, more than half of them<br />

women, came from Africa, Asia, Europe<br />

and the Americas.<br />

Full-text documents available in<br />

the INIS repository represent almost<br />

entirely nuclear-related non-conventional<br />

literature. Non-conventional<br />

literature includes any literature<br />

which is not normally available<br />

through commercial distribution<br />

channels and which is generally<br />

difficult to locate. The depth and<br />

breadth of the INIS repository and<br />

the large number of daily users<br />

demonstrate how a well-planned and<br />

implemented international cooperation<br />

project can make a significant<br />

contribution to open science.<br />

| www.iaea.org<br />

Company News<br />

Framatome successfully<br />

implements innovative<br />

maintenance technique<br />

on reactor vessel<br />

component underwater<br />

(framatome) Framatome applied a<br />

cutting-edge maintenance technique<br />

on reactor vessel primary nozzles at<br />

Dominion Energy’s Millstone <strong>Power</strong><br />

Station during the plant’s spring 2019<br />

outage. This was the first application<br />

of Framatome’s ultra-high pressure<br />

(UHP) cavitation peening process on<br />

reactor pressure vessel nozzles to<br />

primary pipe welds. Because it is<br />

deployed directly to the inner surface,<br />

it is uniquely suited to remediate the<br />

component regardless of external<br />

space restrictions or dose constraints.<br />

“Framatome’s innovative solutions<br />

are ensuring the efficient and reliable<br />

operation of today’s reactor fleet,” said<br />

Catherine Cornand, Framatome’s<br />

senior executive vice president in<br />

charge of the Installed Base Business<br />

Unit. “This new underwater application<br />

of UHP cavitation peening on a<br />

primary nozzle is another example of<br />

our team’s expertise and dedication to<br />

innovation and continuous improvement<br />

in servicing our customers<br />

worldwide.”<br />

To prepare <strong>for</strong> the work, Framatome<br />

demonstrated the qualified reactor<br />

vessel primary nozzle cavitation<br />

peening technology on a full-scale<br />

mock-up at the company’s world-class<br />

Technical Training Center in<br />

Lynchburg, Virginia, in early 2019.<br />

UHP cavitation peening is designed<br />

to prevent primary water stress<br />

corrosion cracking. The process uses<br />

ultra-high-pressure water jets to<br />

generate vapor bubbles that collapse<br />

with enough <strong>for</strong>ce to create beneficial<br />

compression of the components’<br />

surfaces. This surface compression<br />

improves components’ material<br />

properties and enhances resistance<br />

to corrosion and other types of<br />

degradation, which reduces the<br />

effects of aging.<br />

UHP cavitation peening can extend<br />

the life of nuclear reactor primary<br />

components, including the hot leg<br />

primary nozzles, <strong>for</strong> up to 40 additional<br />

years. Additionally, the process<br />

reduces outage time and saves money<br />

by eliminating the need to replace<br />

components or address indications<br />

with traditional repair methods. UHP<br />

cavitation peening can be used <strong>for</strong><br />

several different applications in most<br />

reactor designs.<br />

“Cavitation peening is an industry<br />

game-changer that was recognized in<br />

2017 as one of the Top Innovative<br />

Practices <strong>for</strong> work completed on the<br />

Byron and Braidwood reactor vessel<br />

closure heads,” said Craig Ranson,<br />

senior vice president of Framatome’s<br />

North America Installed Base Business<br />

Unit. “We are proud to work with<br />

Dominion to expand our proven<br />

capabilities and engineer a solution<br />

<strong>for</strong> this unique primary nozzle repair.”<br />

Located in Water<strong>for</strong>d, Connecticut,<br />

the Millstone <strong>Power</strong> Station’s two<br />

pressurized water reactors produce<br />

enough electricity to power 2.1 million<br />

homes.<br />

| www.framatome.com<br />

UK Government and industry<br />

champion new compact<br />

nuclear power station<br />

(rolls-royce) UK Research and Innovation<br />

(UKRI) has confirmed it has<br />

provided initial match funding to the<br />

consortium of companies designing a<br />

new type of nuclear power station in<br />

the UK.<br />

| Artist's view of the SMR design of a new UK nuclear power station project.<br />

The initial joint investment of<br />

£18million from UKRI will be matched<br />

by nuclear, civil engineering construction<br />

and manufacturing industry<br />

firms, who have been working on the<br />

preliminary design <strong>for</strong> four years.<br />

The power station is a compact<br />

design, the components <strong>for</strong> which are<br />

manufactured in sections in regional<br />

UK factories, be<strong>for</strong>e being transported<br />

to existing nuclear sites <strong>for</strong> rapid<br />

assembly inside a weatherproof<br />

canopy. This cuts costs by avoiding<br />

weather disruptions and secures<br />

gradual efficiency savings by using<br />

streamlined and standardised manufacturing<br />

processes <strong>for</strong> its components.<br />

By 2050 a full UK programme of up<br />

to 16 of these power stations could<br />

create:<br />

p Up to 40,000 jobs<br />

p £ 52 bn of value to the UK<br />

economy<br />

p £ 250 bn of exports<br />

Paul Stein, Chief Technology Officer<br />

<strong>for</strong> Rolls-Royce, which leads the<br />

consortium, said: “Tackling climate<br />

change requires collaboration across<br />

industries and governments to find<br />

effective, af<strong>for</strong>dable and sustainable<br />

ways of achieving net zero by 2050.<br />

“The consortium’s work with the<br />

Government shows that action is<br />

being taken to decarbonise our<br />

economy and meet our society’s vital<br />

and growing power needs. This is a<br />

very positive step <strong>for</strong>ward to this next<br />

phase of the programme.”<br />

The partners in the consortium<br />

are Assystem, BAM Nuttall, Laing<br />

O’Rourke, National <strong>Nuclear</strong> Laboratory<br />

(NNL), Rolls-Royce, Atkins,<br />

Wood, The Welding Institute (TWI)<br />

and <strong>Nuclear</strong> AMRC.<br />

The target cost <strong>for</strong> each station is<br />

£1.8 billion by the time five have been<br />

built, with further savings possible.<br />

Each power station will be able to<br />

operate <strong>for</strong> 60 years and provide<br />

440 MW of electricity, which is<br />

enough to power a city the size of<br />

Leeds.<br />

News


<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

The shared initial investment will<br />

be used to progress the significant<br />

opportunities presented by the programme;<br />

prepare it <strong>for</strong> the UK’s<br />

regulatory Generic Design Assessment<br />

process; and make final decisions on<br />

which innovations to pursue and realise.<br />

It will also generate the valuable<br />

confidence that the supply chain needs<br />

to begin to prepare <strong>for</strong> a programme<br />

that could create around £52 billion of<br />

value <strong>for</strong> the UK economy.<br />

When licensed and supported by<br />

the required enabling legislation and<br />

siting processes, the power station<br />

could provide reliable low carbon<br />

energy from the early 2030s.<br />

The Government’s intent to<br />

support the programme was announced<br />

in July 2019.<br />

| www.rolls-royce.com<br />

Uranium<br />

Prize range: Spot market [USD*/lb(US) U 3O 8]<br />

140.00<br />

120.00<br />

100.00<br />

80.00<br />

60.00<br />

40.00<br />

20.00<br />

0.00<br />

1980<br />

Yearly average prices in real USD, base: US prices (1982 to1984) *<br />

1985<br />

1990<br />

1995<br />

2000<br />

2005<br />

) 1<br />

2010<br />

2015<br />

2019<br />

Year<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2019<br />

* Actual nominal USD prices, not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2019<br />

| Uranium spot market prices from 1980 to 2019 and from 2008 to 2019. The price range is shown.<br />

In years with U.S. trade restrictions the unrestricted uranium spot market price is shown.<br />

Separative work: Spot market price range [USD*/kg UTA]<br />

Conversion: Spot conversion price range [USD*/kgU]<br />

180.00<br />

22.00<br />

) 1<br />

160.00<br />

140.00<br />

120.00<br />

100.00<br />

80.00<br />

60.00<br />

40.00<br />

Uranium prize range: Spot market [USD*/lb(US) U 3O 8]<br />

140.00<br />

120.00<br />

100.00<br />

80.00<br />

60.00<br />

40.00<br />

20.00<br />

0.00<br />

20.00<br />

18.00<br />

16.00<br />

14.00<br />

12.00<br />

10.00<br />

Jan. 2008<br />

8.00<br />

6.00<br />

4.00<br />

Jan. 2009<br />

Jan. 2010<br />

) 1<br />

Jan. 20<strong>11</strong><br />

Jan. 2012<br />

Jan. 2013<br />

Jan. 2014<br />

Jan. 2015<br />

Jan. 2016<br />

Jan. 2017<br />

Jan. 2018<br />

Jan. 2019<br />

Jan. 2020<br />

561<br />

NEWS<br />

) 1 Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2019<br />

Westinghouse signs<br />

agreement to acquire Rolls-<br />

Royce Civil <strong>Nuclear</strong> Systems<br />

and Services Business<br />

(west) Westinghouse Electric Company<br />

signed a definitive agreement to<br />

acquire Rolls-Royce’s Civil <strong>Nuclear</strong><br />

Systems and Services business in North<br />

America, expanding Westinghouse’s<br />

global capabilities in digital, engineering<br />

services, plant automation and<br />

monitoring systems, field services and<br />

manufacturing.<br />

“Creating customer value and<br />

supporting our customers’ operations<br />

is a key driver <strong>for</strong> Westinghouse.<br />

Acquiring Rolls-Royce will strengthen<br />

our ability to serve the nuclear<br />

operating fleet through an expanded<br />

presence in our core business while<br />

adding new digital offerings,” said<br />

Patrick Fragman, Westinghouse president<br />

and chief executive officer. “This<br />

acquisition is an important step in our<br />

growth strategy. We look <strong>for</strong>ward to<br />

welcoming the employees of Rolls-<br />

Royce to Westinghouse.”<br />

The acquisition of Rolls-Royce will:<br />

p Expand Westinghouse’s operating<br />

plant services capabilities<br />

p Enhance the company’s digital<br />

innovation ef<strong>for</strong>ts to optimize<br />

customer planning and maintenance,<br />

and provide engineering<br />

solutions to maximize cost<br />

effectiveness and obsolescence risk<br />

p Support both Westinghouse’s and<br />

Rolls-Royce’s global customer base<br />

through an expanded presence<br />

and synergies between both<br />

companies, enhancing customer<br />

offerings and experience in field<br />

services and plant automation<br />

p Further enable Westinghouse’s<br />

growth while supporting<br />

20.00<br />

0.00<br />

Jan. 2008<br />

Jan. 2009<br />

Jan. 2010<br />

Jan. 20<strong>11</strong><br />

Jan. 2012<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

customers in the North American<br />

and European nuclear markets<br />

Rolls-Royce operates <strong>11</strong> sites in<br />

Canada, France, the United Kingdom<br />

and the United States.<br />

| www.westinghousenuclear.com<br />

Market data<br />

(All in<strong>for</strong>mation is supplied without<br />

guarantee.)<br />

<strong>Nuclear</strong> Fuel Supply<br />

Market Data<br />

In<strong>for</strong>mation in current (nominal)<br />

U.S.-$. No inflation adjustment of<br />

prices on a base year. Separative work<br />

data <strong>for</strong> the <strong>for</strong>merly “secondary<br />

market”. Uranium prices [US-$/lb<br />

U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 =<br />

0.385 kg U]. Conversion prices [US-$/<br />

kg U], Separative work [US-$/SWU<br />

(Separative work unit)].<br />

Jan. 2013<br />

Jan. 2014<br />

Jan. 2015<br />

Jan. 2016<br />

2017<br />

p Uranium: 19.25–26.50<br />

p Conversion: 4.50–6.75<br />

p Separative work: 39.00–50.00<br />

2018<br />

p Uranium: 21.75–29.20<br />

p Conversion: 6.00–14.50<br />

p Separative work: 34.00–42.00<br />

2019<br />

January 2019<br />

p Uranium: 28.70–29.10<br />

Jan. 2017<br />

Jan. 2018<br />

Jan. 2019<br />

Jan. 2020<br />

Sources: Energy Intelligence, Nukem; Bild/Figure: <strong>atw</strong> 2019<br />

2.00<br />

0.00<br />

* Actual nominal USD prices, not real prices referring to a base year. Year<br />

| Separative work and conversion market price ranges from 2008 to 2019. The price range is shown.<br />

)1<br />

In December 2009 Energy Intelligence changed the method of calculation <strong>for</strong> spot market prices. The change results in virtual price leaps.<br />

* Actual nominal USD prices, not real prices referring to a base year<br />

Sources: Energy Intelligence, Nukem; Bilder/Figures: <strong>atw</strong> 2019<br />

Jan. 2008<br />

Jan. 2009<br />

Jan. 2010<br />

Jan. 20<strong>11</strong><br />

Jan. 2012<br />

p Conversion: 13.50–14.50<br />

p Separative work: 41.00–44.00<br />

February 2019<br />

p Uranium: 27.50–29.25<br />

p Conversion: 13.50–14.50<br />

p Separative work: 42.00–45.00<br />

March 2019<br />

p Uranium: 24.85–28.25<br />

p Conversion: 13.50–14.50<br />

p Separative work: 43.00–46.00<br />

April 2019<br />

p Uranium: 25.50–25.88<br />

p Conversion: 15.00–17.00<br />

p Separative work: 44.00–46.00<br />

May 2019<br />

p Uranium: 23.90–25.25<br />

p Conversion: 17.00–18.00<br />

p Separative work: 46.00–48.00<br />

June 2019<br />

p Uranium: 24.30–25.00<br />

p Conversion: 17.00–18.00<br />

p Separative work: 47.00–49.00<br />

July 2019<br />

p Uranium: 24.50–25.60<br />

p Conversion: 18.00–19.00<br />

p Separative work: 47.00–49.00<br />

August 2019<br />

p Uranium: 24.90–25.60<br />

p Conversion: 19.00–20.00<br />

p Separative work: 47.00–49.00<br />

September 2019<br />

p Uranium: 24.80–26.00<br />

p Conversion: 20.00–21.00<br />

p Separative work: 47.00–50.00<br />

| Source: Energy Intelligence<br />

www.energyintel.com<br />

Jan. 2013<br />

Jan. 2014<br />

Jan. 2015<br />

Jan. 2016<br />

Jan. 2017<br />

Jan. 2018<br />

Jan. 2019<br />

Jan. 2020<br />

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<strong>atw</strong> Vol. 64 (2019) | Issue <strong>11</strong>/12 ı November/December<br />

562<br />

NUCLEAR TODAY<br />

John Shepherd is a<br />

freelance journalist<br />

and communications<br />

consultant who has<br />

covered the nuclear<br />

industry <strong>for</strong> the past<br />

20 years and is<br />

currently editor-in-chief<br />

of UK-based Energy<br />

Storage Publishing.<br />

Sources<br />

World <strong>Nuclear</strong><br />

Association<br />

https://cutt.ly/weEfVJC<br />

IEA Sweden report<br />

https://cutt.ly/TeEfZIs<br />

Cryospheric Sciences<br />

Division blog<br />

https://cutt.ly/1eEfLyD<br />

Taking a Leaf out of Greta’s<br />

Climate Change Book<br />

The media loves a ‘warrior underdog’ and so journalists around the world have largely fallen in love with Swedish eco<br />

champion Greta Thunberg. The diminutive teenager has captured the imagination of a new generation of environmental<br />

activists since she burst onto the international scene around a year ago, when she protested on school days outside the<br />

Swedish parliament. She said her school strike was to draw attention to global warming.<br />

Her actions caught on and soon led to other students<br />

making similar protests. The rest, to coin a phrase, is<br />

already modern history. Thunberg has been feted worldwide<br />

and she even addressed the 2018 United Nations<br />

Climate Change Conference.<br />

Whatever one thinks of Thunberg’s approach, it’s hard<br />

to deny that she has reignited interest in the climate change<br />

debate and raised the political temperature on all sides.<br />

So could nuclear’s proponents take a leaf out of the<br />

young environmentalist’s book? Earlier this year, a group<br />

of reactor physicists, operators and politicians were among<br />

those who followed Thunberg’s lead and gathered outside<br />

the Swedish parliament to raise awareness of the benefits<br />

of nuclear energy.<br />

One of the organisers of the ‘Stand up <strong>for</strong> <strong>Nuclear</strong>’ event<br />

was Swedish <strong>Nuclear</strong> Society president Marcus Eriksson.<br />

He told me that, on the same day of the event, the group<br />

published an opinion article in one of Sweden’s major<br />

newspapers, Svenska Dagbladet, which explained the<br />

country’s energy situation in Sweden, making a parallel to<br />

Germany’s energy sector.<br />

The Stand up <strong>for</strong> <strong>Nuclear</strong> event is an important development<br />

in Sweden and one that could and should have<br />

wider implications <strong>for</strong> Europe. Sweden has a chequered<br />

past where nuclear is concerned.<br />

According to the World <strong>Nuclear</strong> Association, Sweden<br />

generated a combined 156 terawatt hours (TWh) of<br />

electricity in 2016, of which 63 TWh (40%) was from its<br />

eight nuclear power reactors and 62 TWh (40%) from<br />

hydro. Wind provided 15.5 TWh (10%), various fossil fuels<br />

2 TWh, and biofuels and waste 13 TWh (8%). Total<br />

installed generating capacity at the end of 2016, as<br />

recorded by the <strong>International</strong> Energy Agency‘s (IEA)<br />

‘ Electricity In<strong>for</strong>mation 2018’, was 40 gigawatts electric.<br />

In 1980, the Swedish government decided to phase out<br />

nuclear power but this policy was repealed by lawmakers<br />

in June 2010. The country‘s 1997 energy policy had<br />

allowed 10 reactor units to operate longer than envisaged<br />

by the 1980 phase-out policy. However, the policy also led<br />

to the premature closure of a two-unit (1,200 megawattelectric)<br />

plant, although some 1,600 MWe was subsequently<br />

added in uprates to the remaining 10 units.<br />

In 2015, decisions were made to close four older<br />

reactors by 2020. A year later, a tax discriminating against<br />

nuclear power was abolished.<br />

<strong>Nuclear</strong>’s impact on Sweden’s environmental credentials<br />

is impressive. The country has almost fully decarbonised its<br />

electricity generation, which the IEA has described as “a<br />

feat which is quite unique among the IEA member countries”.<br />

In a review of Swedish energy policy published last<br />

April, the IEA said the country should assess the contribution<br />

of nuclear power over the next 20 years and the impact<br />

of further potential early closures on energy security.<br />

I have no scientific basis <strong>for</strong> saying this, only a gut<br />

feeling (from reporting on nuclear and the wider energy<br />

industry over the past 20 years), that there is often a silent<br />

majority who would speak up in favour of nuclear –<br />

provided they are given the facts.<br />

It appears such a majority could exist in Sweden, if a<br />

survey conducted last summer by Novus Opinion is<br />

anything to go by. That survey indicated a majority of<br />

Swedes now believe that nuclear power could be a means<br />

to tackle the climate crisis. Which brings me back to the<br />

Stand up <strong>for</strong> <strong>Nuclear</strong> event organised in Stockholm.<br />

Marcus Eriksson told journalists covering the event:<br />

“Two years ago none of the political parties wanted to talk<br />

about nuclear power, now everyone is talking about it. It<br />

reflects a stronger opinion that the technology has an<br />

important role to play to combat climate change.”<br />

The Swedish <strong>Nuclear</strong> Society should be praised <strong>for</strong><br />

raising awareness of the benefits of nuclear in this way.<br />

Every day, when we are out and about, there is a manifestation<br />

of some sort or another with campaigners competing<br />

<strong>for</strong> our attention on issues. So why not something<br />

that speaks up <strong>for</strong> nuclear? In these gloomy winter nights,<br />

perhaps there is an inbuilt advantage <strong>for</strong> the nuclear cause<br />

in pointing out just how dark and cold many of our communities<br />

would be without the benefit of atomic power.<br />

As we head into a new year, it’s a good time to start<br />

thinking about the resolutions we will make <strong>for</strong> the months<br />

ahead. In addition to resolution favourites, such as going<br />

on a diet, or giving up smoking, those involved in the<br />

nuclear sector should pledge to make a special ef<strong>for</strong>t to<br />

stand up <strong>for</strong> nuclear in 2020.<br />

The initial unassuming campaign launched by Greta<br />

Thunberg just a year ago is, I suggest, a template <strong>for</strong> what<br />

could be a ‘2020 Vision <strong>for</strong> <strong>Nuclear</strong>’ campaign <strong>for</strong> the year<br />

ahead. Perhaps with the advent of social media we’ve<br />

<strong>for</strong>gotten about the impact of ‘grassroots’ campaigning.<br />

Social media is of course tremendously valuable in<br />

reaching millions. But let’s not <strong>for</strong>get the impact that can<br />

be had by being out on the streets and chatting to passersby<br />

about the technology that keeps the lights on and can<br />

help clean up our local and global environments.<br />

There’s no time like the present to get out there and<br />

follow Greta’s footprint. In fact, the beloved Santa Clause,<br />

a figure who returns to prominence at this time of the year,<br />

could be recruited to underline the importance of the<br />

nuclear cause.<br />

As the Cryospheric Sciences Division of the European<br />

Geosciences Union has pointed out, Santa might well have<br />

to think about moving from his fabled home of Lapland<br />

because of global warming and polar amplification. In the<br />

absence of snow, Santa does have the possibility of<br />

converting his sleigh to an all-electric model. But even if<br />

there were enough charging points around the globe, they<br />

would need reliable supplies of clean electricity, which<br />

brings us back to nuclear’s importance <strong>for</strong> our planet’s<br />

energy mix.<br />

In wrapping up this festive article, it only remains to<br />

wish everyone the compliments of the season and, if not a<br />

white Christmas, at least an increasingly green one.<br />

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

Taking a Leaf out of Greta’s Climate Change Book ı John Shepherd


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