atw - International Journal for Nuclear Power | 11/12.2019

nucmag.com 

2019 

11/12 

Performance Shaping 

Factors for Human Error 

Reduction 

Decommissioning & Dismantling 

of the Rossendorf Research 

Reactor RFR 

First On-site Demonstration 

of Laser- based Decontamination 

Technology in Germany 

ISSN · 1431-5254 

24.– € 

Register Now!

Kommunikation und 

Training für Kerntechnik 

Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik? 

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

3 Atom-, Vertrags- und Exportrecht 

Atomrecht – Ihr Weg durch Genehmigungs- und 

Aufsichtsverfahren 

RA Dr. Christian Raetzke 18.02.2020 Berlin 

Atomrecht – Was Sie wissen müssen 

RA Dr. Christian Raetzke 

Akos Frank LL. M. 

11.11.2020 Berlin 

Atomrecht – Das Recht der radioaktiven Abfälle RA Dr. Christian Raetzke 10.03.2020 Berlin 

Export kerntechnischer Produkte und Dienstleistungen – 

Chanchen und Regularien 

3 Kommunikation und Politik 

RA Kay Höft M.A. (BWL) 17.06.2020 Berlin 

Public Hearing Workshop – 

Öffentliche Anhörungen erfolgreich meistern 

Dr. Nikolai A. Behr 10.11. - 11.11.2020 Berlin 

3 Rückbau und Strahlenschutz 

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

Das neue Strahlenschutzgesetz – 

Folgen für Recht und Praxis 

Stilllegung und Rückbau in Recht und Praxis 

Dr. Maria Poetsch 


Dr. Stefan Kirsch 


28.01. - 29.01.2020 

16.06. - 17.06.2020 

23.09. - 24.09.2020 

Berlin 

17.03. - 18.03.2020 Berlin 

3 Nuclear English 

English for Nuclear Business Angela Lloyd 01.04. - 02.04.2020 Berlin 

3 Wissenstransfer und Veränderungsmanagement 

Veränderungsprozesse gestalten – Heraus forderungen 

meistern, Beteiligte gewinnen 

Erfolgreicher Wissenstransfer in der Kerntechnik – 

Methoden und praktische Anwendung 

Dr. Tanja-Vera Herking 

Dr. Christien Zedler 

Dr. Tanja-Vera Herking 

Dr. Christien Zedler 

21.01. - 22.01.2020 Berlin 

24.03. - 25.03.2020 Berlin 

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

Kontakt 

INFORUM Verlags- und Verwaltungs gesellschaft mbH ı Robert-Koch-Platz 4 ı 10115 Berlin 

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

Die INFORUM-Seminare können je nach 

Inhalt ggf. als Beitrag zur Aktualisierung 

der Fachkunde geeignet sein.

atw Vol. 64 (2019) | Issue 11/12 ı November/December 

TMI and Lessons Learned – Afterwards 

and for the Future 

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

technology therefore requires a responsible and forward-looking approach to its risks, the best possible protection and, 

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

absolute safety. 

With a view to criticism of technologies, technology sceptics 

are provided with a comfortable approach for transporting 

their respective concerns by anniversaries. One such event 

is the 1979 Three-Mile-Island (TMI 2) accident. 

On 28 March 1979, shortly after 4 a.m., the feed water 

supply for the steam generators in unit 2 of the Three Mile 

Island (TMI 2) nuclear power plant near Harrisburg, 

Pennsylvania, USA, failed. Such a malfunction was taken 

into account in the operating procedures with a view to 

the future and can be easily dealt with by the plant and 

operators. The event is one of the design cases. But what 

began as a normal incident later developed into the most 

serious accident to date in a Western nuclear power plant 

due to a combination of technical defects, human error 

and organisational inadequacies. 

The technical aspects of the TMI-2 accident have been 

extensively reviewed and can be traced in detail in the 

literature: After the failure of the feed-water supply, the 

safety systems of the nuclear power plant reacted as 

designed. However, due to valves in the emergency feedwater 

system that were closed manually on site, the steam 

generators were not supplied with water and a link in the 

chain for the residual heat dissipation of the reactor, which 

was switched off but still supplies heat due to the decay 

heat, was missing. This error was corrected 8 minutes 

after the start of the fault. Subsequently, the increase in 

pressure in the primary circuit caused the safety relief 

valve on the pressure holder to open, the escape of steam 

from the primary circuit and the intended pressure 

reduction. However, the opened blow-off valve did not 

close as designed. This caused large quantities of coolant 

to escape from the primary circuit – a loss of coolant 

accident “ LOCA” had occurred. Only 2 hours later it was 

detected by the operators and corrected by closing other 

existing shut-off valves. After further misinterpretations 

about the condition of the reactor core and the primary 

circuit, the core was no longer cooled by water and the 

exothermic reaction between water vapour and the 

cladding tube material of the fuel elements then set on, 

releasing additional heat. About half of the fuel rod mass 

in the core melted or was severely damaged. Core masses 

shifted and partially reached the bottom of the reactor 

pressure vessel where they solidified. After it had been 

recognised that a core melt was in full swing, measures for 

core cooling were taken and it was possible to ensure the 

heat removal of the TMI-2 core again in the long term. 

Of the radioactive substances released in the plant, a 

small part was released into the environment. Triggered 

by fears of the additional radiation exposure associated 

with the release of a calculated average of 0.02 mSv – i.e. 

about 1 % of the natural annual radiation exposure – the 

Pennsylvania Department of Health kept a register for 

18 years containing the data of more than 30,000 persons 

who had lived within a five-mile radius of the nuclear 

power plant at the time of the accident. This state register 

was closed in June 1997 after no unusual health 

developments had been identified. More than a dozen 

comprehensive studies on the physical health effects of the 

accident did not provide evidence of an abnormally high 

number of cancer cases in the region around the nuclear 

power plant years after the event. 

Despite all the shortcomings and shortcomings that 

led to the TMI-2 accident, the basic safety concept of 

Western nuclear power plants that had been established at 

an early stage was confirmed: The amount of radioactive 

substances released from the plant was low and there were 

no fatalities or injuries. On the other hand, the accident 

sequence underlines the suitability of successively 

staggered safety barriers, which, taking even serious 

accidents into account, aims at the confinement of the 

radioactive substances in the plant itself: The reactor 

building of TMI-2 withstood, the containment remained 

intact. 

In addition, research and development for reactor 

safety further minimised the risks in an international 

context during the past 40 years: Expressed in figures, 

even for running plants in Germany and elsewhere, a 

safety level was achieved, which is 100 times higher than 

the original international reference level. The reactor 

disaster at Fukushima in 2011 will not change this point 

of view either, as other foreseeable risks and specific 

boundary conditions were the causes. 

Safety is and remains an international task for all 

concerned. Nuclear safety and the maintenance and 

promotion of competences, especially when they are 

available with excellent know-how in research and 

development and are internationally recognised – as in 

Germany – are part of the overall social and political 

responsibility at the same level as other goals of 

environmental protection. Today's and future safety and 

the promotion of safety culture and technology, also for 

nuclear technology, cannot and must not be the subject 

of restricted or restrictive action. Research and 

development should always be allowed to advance in 

the spirit of freedom of science and technology. 

Christopher Weßelmann 

– Editor in Chief – 

507 

EDITORIAL 

Editorial 

TMI and Lessons Learned – Afterwards and for the Future


508 

EDITORIAL 

TMI und Lessons Learned – danach 

und für die Zukunft 

Keine technische Entwicklung ist von vornherein perfekt, jede birgt spezifische Risiken. Ein verantwortungsvoller 

Umgang mit Technik erfordert daher einen verantwortungsvollen und vorausblickenden Umgang mit ihren 

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

absolut perfekte Technik nicht gibt, ebenso keine absolute Sicherheit. 

Mit Blick auf Kritik an Technologien liefern Jahrestage gerne 

Technologieskeptikern einen komfortablen Ansatz, ihre 

jeweiligen Anliegen zu transportieren. Ein solches Ereignis 

ist auch der Three-Mile-Island (TMI 2)-Unfall von 1979. 

Am 28. März 1979, kurz nach 4 Uhr morgens, fiel im 

Block 2 des US-amerikanischen Kernkraftwerks Three Mile 

Island (TMI 2) nahe Harrisburg, Pennsylvania, die 

Speisewasser versorgung der Dampferzeuger aus. Eine 

solche Störung ist in den Betriebsabläufen vorausblickend 

berücksichtigt und kann von Anlage und Operateuren ohne 

Weiteres bewältigt werden. Das Ereignis gehört zu den 

Auslegungsfällen. Doch was als normale Störung begann, 

entwickelte sich im weiteren Verlauf aufgrund einer Kombination 

aus technischen Mängeln, mensch lichen Fehlern und 

organisatorischen Unzulänglichkeiten zum bis dahin schwerwiegendsten 

Unfall in einem west lichen Kernkraftwerk. 

Die technischen Aspekte zum TMI-2-Unfall sind ausgiebig 

aufgearbeitet und können in der Literatur detailliert 

nachvollzogen werden: Nach dem Ausfall der Speisewasserversorgung 

reagierten die Sicherheitssysteme des Kernkraftwerks 

auslegungsgemäß. Aufgrund versehentlich 

manuell Vor-Ort geschlossener Ventile im Notspeisewassersystem 

fehlte aber die Bespeisung der Dampferzeuger und 

damit ein Glied in der Kette für die Nachwärmeabfuhr des 

zwar abgeschalteten aber durch die Nachzerfalls wärme 

immer noch Wärme liefernden Reaktors. Dieser Fehler 

wurde 8 Minuten nach Störungsbeginn korrigiert. In 

weiterer Folge kam es durch den Druckanstieg im Primärkreislauf 

zum Öffnen des für diesen Fall vorgesehenen 

Sicherheits-Abblaseventils am Druckhalter, dem Entweichen 

von Dampf aus dem Primärkreislauf und der 

beabsichtigten Druckabsenkung. Das geöffnete Abblaseventil 

schloss aber folgend nicht auslegungsgemäß. Über 

diesen Weg entwichen große Mengen Kühlmittel aus dem 

Primärkreislauf – ein Loss of coolant accident „LOCA“ war 

eingetreten. Erst 2 Stunden später wurde er durch die 

Operateure erkannt und durch Schließen weiterer vorhandener 

Absperrarmaturen korrigiert. Nach weiteren 

Fehlinterpretationen über den Zustand von Reaktorkern 

und Primärkreislauf war der Kern nicht mehr durch Wasser 

gekühlt und die dann einsetzende exotherme Reaktion 

zwischen Wasserdampf und dem Hüllrohrmaterial der 

Brenn elemente setzte zusätzliche Wärme frei. Etwa die 

Hälfte der Brennstabmasse im Kern schmolz bzw. wurde 

schwer beschädigt. Kernmassen verlagerten sich und 

erreichten teilweise den Boden des Reaktordruckbehälters, 

wo sie erstarrten. Nachdem erkannt worden war, dass eine 

Kernschmelze in vollem Gange war, wurden Maßnahmen 

zur Kernkühlung ergriffen und es gelang, die Wärmeabfuhr 

des TMI-2-Kerns wieder langfristig zu gewährleisten. 

Von den in der Anlage frei gesetzten radioaktiven 

Stoffen gelangte ein geringer Teil in die Umgebung. 

Ausgelöst durch die Ängste der mit der Freisetzung 

verbundenen zusätzlichen Strahlenbelastung von berechneten 

im Mittel 0,02 mSv – also etwa 1 % der natürlichen 

jährlichen Strahlenbelastung – führte das Pennsylvania 

Department of Health während 18 Jahren ein Register mit 

den Daten von mehr als 30.000 Personen, die zum Zeitpunkt 

des Unfalls im Umkreis von fünf Meilen um das 

Kernkraftwerk gelebt hatten. Dieses staatliche Register 

wurde im Juni 1997 geschlossen, nachdem keine ungewöhnlichen 

Entwicklungen bei der Gesundheit festgestellt 

worden waren. Mehr als ein Dutzend Studien über die 

Auswirkungen des Unfalls auf die physische Gesundheit 

gaben auch Jahre nach dem Ereignis keine Hinweise auf 

eine abnormal hohe Zahl von Krebsfällen in der Region um 

das Kernkraftwerk. 

Trotz aller Unzulänglichkeiten und Fehler die zum TMI- 

2-Unfall führten, bestätigte sich einerseits das grundlegende 

frühzeitig etablierte Sicherheitskonzept westlicher 

Kernkraftwerke: Die aus der Anlage freigesetzte Menge an 

radioaktiven Stoffen war gering und weder Todesopfer 

noch Verletzte waren zu beklagen. Und andererseits 

unterstreicht der Unfallablauf den Sinn hintereinander 

gestaffelter Sicherheitsbarrieren, die unter Berücksichtigung 

auch schwerer Unfälle, den Einschluss der radioaktiven 

Stoffe in der Anlage selbst zum Ziel haben: Das 

Reaktorgebäude von TMI-2 hielt stand, der Sicherheitsbehälter 

blieb intakt. 

Zudem haben Forschung und Entwicklung für die 

Reaktor sicherheit in den letzten 40 Jahren die Risiken 

im internationalen Kontext weiter minimiert: In Zahlen 

ausgedrückt wurde selbst für laufende Anlagen in 

Deutschland und anderswo ein Sicherheitsniveau erreicht, 

das um den Faktor 100 höher liegt, als das ursprüngliche 

inter nationale Referenz niveau. Daran ändert auch die 

Reaktorhavarie von Fukushima 2011 nichts, da andere 

absehbare Ursachen und spezifische Rand bedingungen 

Auslöser waren. 

Sicherheit ist und bleibt dabei eine international zu 

lebende Aufgabe für alle Beteiligten. Kerntechnische 

Sicherheit und der Erhalt und die Förderung von 

Kompetenzen, vor allem dann, wenn sie mit exzellentem 

Know-how in Forschung und Entwicklung vorhanden und 

international anerkannt sind – wie in Deutschland – 

gehören auf gleicher Ebene zur gesamtgesellschaftlichen 

und politischen Verantwortung wie andere Ziele des 

Umweltschutzes. Heutige und künftige Sicherheit und 

Förderung von Sicherheitskultur und -technik, auch für 

die Kerntechnik, können und dürfen nicht Gegenstand 

eingeschränkten oder beschränkenden Handelns sein. 

Forschung und Entwicklung sollte immer im Geist der 

Freiheit von Wissenschaft und Technik vorgetrieben 

werden dürfen. 

Christopher Weßelmann 

– Chefredakteur – 

Editorial 

TMI und Lessons Learned – danach und für die Zukunft


Did you know...? 

Growing Popular Support for Nuclear Power in Belgium 

Forum Nucléaire, the association of the nuclear sector in Belgium, 

commissioned the 2019 edition of its regular opinion poll 

with Kantar TNS. Between 15 July and 6 September 2019 

756 Belgians over the age of 16 were interviewed by phone 

about their opinion on nuclear power and the nuclear sector 

in Belgium. The opinion poll shows a clear increase in support 

for nuclear power in Belgium compared to the 2017 edition 

of the poll, in particular with regard to the security of supply and 

the reduction of CO 2 -emissions. Below you can find a selection 

of results. The complete poll was published 18 October 2019 

on www.forumnucleaire.be and is available there in Dutch and 

French. 

For further details 

please contact: 

Nicolas Wendler 

KernD 

Robert-Koch-Platz 4 

10115 Berlin 

Germany 

E-mail: presse@ 

KernD.de 

www.KernD.de 

DID YOU EDITORIAL KNOW...? 

509 

For our country the production of nuclear energy in Belgium means … 

2017 

(n=1027) 

49 % 

13 % 

38 % 

2019 

(n=756) 

57 % 

More advantages than inconvenients 9 % 

Advantages 

and inconvenients equally 

32 % 

More inconvenients than advantages 

Are you favourable to keeping nuclear in Belgium for the production of electricity? 

2017 

(n=1027) 

30 % 

50 % 

19 % 2 % 

2019 

(n=756) 

46 % 

Yes, in long term too 

37 % 

Yes, but in short term (2025) 

16 % 

No 

1 % 

Don't know 

The current legislation foresees the closure of all nuclear power plants by 2025. 

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

69 % 

No 

28 % 

Yes 

3 % 

Undecided, 

don't know 

It is good to replace nuclear power plants with gas fired power plants 

even if they emit much more CO 2 . (n=756) 

26 % 

I don't agree at all 

51 % 

I rather don't agree 

1 % 

I don't know 

11 % 

I neither agree or disagree 

7 % 

I rather agree 

5 % 

I completely agree 

Figures in percent. Rounded values. 

Source: 

Forum Nucléaire 

Did you know...?


510 

Issue 11/12 | 2019 

November/December 

CONTENTS 

Contents 

Editorial 

TMI and Lessons Learned – Afterwards 

and for the Future E/G 507 

Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . .509 

Inside Nuclear with NucNet 

Key to Unlocking Bulgaria’s Belene Project 

is Finding Right Financing Package 512 

Feature | Environment and Safety 

Development of Performance Shaping Factors for Human Error 

Reduction during Reactor Decommissioning Activities through 

the Task Analysis Method 515 

Environment and Safety 

Root Causes of the Three Mile Island Accident 521 

Spotlight on Nuclear Law 

The New Radiation Protection Law – 

Protection Against Radon G 525 

Research and Innovation 

Evaluation of a Double-Ended Guillotine LBLOCA Transient 

in a Generic Three-Loops PWR-900 with TRACE Code 

Coupled with DAKOTA Uncertainty Analysis 526 

Experiment Research on the Insurge Transient Behavior 

of Gas-steam Pressurizer under Various Pressure 533 

Decommissioning and Waste Management 

Decommissioning & Dismantling of the Rossendorf 

Research Reactor RFR | Part 1 G 537 

First On-site Demonstration of Laser- based Decontamination 

Technology in Germany 543 

Report 

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

on Nuclear Reactors “Physics, Fuels and Systems“ 549 

Special Topic | A Journey Through 50 Years AMNT 

Protection of Man and Environment – Nuclear Usage 

Outside of Energy Sector G 550 

KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .555 

Cover: 

Korea Kori NPP. 

Copyright: ©Korea Kori NPP 

News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .556 

Nuclear Today 

Taking a Leaf out of Greta’s Climate Change Book 562 

G 

E/G 

= German 

= English/German 

Imprint 542 

Insert: INFORUM – Seminarprogramm 2020 

Contents


Feature 


515 Development of Performance Shaping 

Factors for Human Error Reduction 

during Reactor Decommissioning Activities 

through the Task Analysis Method 

511 

CONTENTS 

Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam 


521 Root Causes of the Three Mile Island Accident 

Zoltan R. Rosztoczy 


526 Evaluation of a Double-Ended Guillotine LBLOCA Transient 

in a Generic Three-Loops PWR-900 with TRACE Code 

Coupled with DAKOTA Uncertainty Analysis 

Andrea Bersano and Fulvio Mascari 


537 Decommissioning & Dismantling 

of the Rossendorf Research Reactor RFR | Part 1 

Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz 

543 First On-site Demonstration 

of Laser- based Decontamination Technology in Germany 

Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, 

Wolfgang Lippmann and Antonio Hurtado 

Contents


512 

INSIDE NUCLEAR WITH NUCNET 

Key to Unlocking Bulgaria’s Belene 

Project is Finding Right Financing Package 

Last year the government decided to formally revive the nuclear power station project. The process is now 

underway to select an investor. 

The History of Belene 

The project goes back to the mid-1970s when a 

two-unit nuclear station at Belene, on the 

Danube river in the north of the country, was planned by 

Bulgaria’s communist government. Site work began in 

1981 and construction of the first Russian-supplied 

VVER-1000 began in 1987. This was aborted in 1991 due to 

lack of funds and environmental and financial concerns. 

The project was revived in 2008, but formally abandoned 

in 2012 after failing to find investors and with Sofia facing 

pressure from Washington and Brussels to limit its energy 

dependence on Russia, which was under contract to build 

the facility. The estimated €10bn cost of the project was 

too high to be funded by Bulgaria alone. In June 2018, the 

government decided to formally revive the project. A call 

for interest was published and seven companies, three of 

which are Bulgarian, applied to invest in the project. 

Russia’s state-owned Rosatom, China’s state-owned CNNC 

and state-run Korea Hydro and Nuclear Power Company 

all filed applications confirming their interest. Press 

reports said that among the three Bulgarian companies 

there are two consortia with interests in energy that have 

been formed specifically for the Belene project and a third 

company, IPK-UP, which is registered in Kozloduy, but 

lacks any registered assets or income. The seventh 

company that filed its interest is Karlsruhe-registered 

Becktron-Liaz Technical Engineering. 

Bulgaria’s Nuclear Status 

Bulgaria, which joined the EU in 2007, closed four 440-MW 

VVER reactors at the Kozloduy nuclear station for safety 

reasons as part of its accession treaty with Brussels. 

Kozloduy has two newer 1,000-MW VVER reactors in commercial 

operation, supplying about 34 % of the country’s 

electricity. These two units have been renovated with 

financing from Euratom, the US ExportImport Bank, 

Citibank and Russia’s Roseximbank. Kozloduy-6 recently 

received a licence to operate for another 10 years until 

2029; Kozloduy-5 received a 10-year extension in 2017. 

Reviving the Belene Project 

The idea to revive the Belene project was born during 

negotiations with the European Union on Bulgaria’s 

accession. The centrist government at the time said Belene 

could replace generation that would be lost with the 

retirement of the four older Kozloduy units. In 2005, a 

socialist-led government revived Belene and a tendering 

procedure for a 2000-MW station was approved. A year 

later, Bulgaria chose an offer by a proposed consortium 

between Atomstroyexport (ASE) of Russia, Areva (now 

Framatome) of France and Siemens of Germany for the 

deployment of two Generation III VVER-1000 PWRs. In 

2008, Bulgaria signed a contract with ASE for the 

design, construction and commissioning of the two units. 

However, in 2012 a new government headed by current 

prime-minister Boyko Borissov cancelled Belene after 

failing to find financing. Russia took Bulgaria to international 

arbitration over the cancellation. 

Bulgaria again began looking for ways to revive Belene 

after it lost the arbitration and paid € 600 m in 2016 in 

compensation to ASE for components which had already 

been ordered. The 2016 decision by the Geneva-based 

International Court of Arbitration of the International 

Chamber of Commerce said Bulgarian state energy 

company Nationalna Elektricheska Kompania (NEK) had to 

pay € 620 m to ASE and assume ownership of the components. 

The price was reduced after interest adjustments. 

By the beginning of 2018 Russia had delivered most of the 

equipment that was at the centre of the court ruling, 

including two reactor pressure vessels, the RPV heads, the 

full steam generator sets and the main pipelines. The 

equipment is stored at the Belene site. 

A Question Of Financing 

According to energy minister Temenuzhka Petkova Bulgaria 

wants to build the two units within 10 years and at a cost of 

up to € 10 bn. However, the exact nature of how the station 

will be financed remains unclear. Ms Petkova told parliament 

that the government would like state energy company 

NEK to have a “blocking stake” – potentially giving it a veto if 

necessary – in the project, but only by contributing existing 

assets and infrastructure at the Belene site, valued at 

€ 1.5 bn. The government’s policy is to attract private investment 

for the project with no state guarantees or long-term 

electricity purchase contracts. Critics have said that would 

be difficult given the magnitude of nuclear new-build 

projects. 

The Need For Nuclear 

Bulgaria says it needs nuclear because it is heavily dependent 

on imported primary energy resources and uses the most 

electricity relative to GDP in the EU. Bulgaria imports almost 

100 % of its oil and gas, 100 % of its nuclear fuel and about 

35 to 40 % of its coal. It exports electricity to neighbouring 

countries, but the electricity sector is dependent on imports 

of primary energy, mainly from Russia. Russia sees Bulgaria 

as a transit point into the rest of Europe for Russian energy 

sources because it bypasses Ukraine. 

Bulgaria is under pressure by the EU to reduce carbon 

emissions in line with the bloc’s climate ambitions. Bulgaria 

generates about 40 % of its electricity from coal and new 

nuclear could replace a major proportion of this. The government 

has also hinted that new reactors might be needed 

to replace the two Kozloduy units, which are scheduled to 

be retired in 2027 and 2029. A report by the Bulgarian 

Academy of Science said electricity demand in Bulgaria is 

expected to increase after 2030 and new nuclear could be 

economically viable if construction costs are kept in check. 

What Happens Next? 

Mr Borissov’s ruling party and its allies command a 

comfortable majority so plans to build new units at Belene 

would not appear to pose a political problem. The 

opposition socialists also support nuclear energy. Public 

support is high, with almost two-thirds of Bulgarians 

voting in support of nuclear in a 2013 referendum, 


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


although the result was not binding because of the low 

turnout. The EU and the US have urged Bulgaria to take 

stock of its dependence on Russian gas and oil and 

encouraged it to seek Western investors for Belene. The 

key is to find the right financing package. China is said to 

be asking for financial guarantees from Sofia in return for 

its participation, but Sofia favours the creation of some 

form of public- private partnership. Any financing scheme 

will need to be approved by the EU’s state aid watchdog. 

A risk analysis of the Belene project, prepared by the 

Vienna International Centre for Nuclear Competence for 

the Bulgarian Atomic Forum, recommended that the 

government reconsider its position of non-participation in 

the project. Globally, there is no precedent in which nuclear 

projects are implemented without state involvement, the 

analysis said. Different forms and levels of participation are 

possible, including direct sponsorship, loan guarantees, tax 

credits, long-term energy purchase agreements and price 

differences, the analysis noted. Another option is to use the 

Belene equipment for a new unit at Kozloduy, but this 

would incur additional costs. Abandoning the project and 

selling the equipment as scrap would result in a loss of 

about $ 1.5 m. The sale of the project is also a possibility to 

companies from Russia or China. 

Whoever the project owner and investors are, the 

government has said the main contractor for the nuclear 

part of the project will be Russia’s Rosatom, because of the 

nature of the technology already paid for by Bulgaria. 

Turbine hall equipment and instrumentation and control 

systems could be supplied by US-based General Electric 

and France’s Framatome. Both companies participated in 

the recent tender, but as equipment suppliers rather than 

investors. 

Author 

David Dalton 

Advertisement 

INSIDE NUCLEAR WITH NUCNET 513 

Radioaktive Abfälle bis zu ihrer Endlagerung nach höchsten Sicherheitsstandards aufzubewahren – das ist unser Auftrag. Als Gesellschaft des Bundes gewährleisten 

wir den sicheren und zuverlässigen Betrieb von Zwischenlagern für hochradioaktive Abfälle. Ab 2020 werden wir zudem zwölf Zwischenlager mit 

schwach- und mittelradioaktiven Abfällen an den Standorten deutscher Kernkraftwerke führen. Dadurch liegt die Verantwortung für die Zwischenlagerung 

der radioaktiven Abfälle der Energieversorgungsunternehmen zentral in unserer Hand. Neben bewährter und solider Technik sind es vor allem unsere 

Mitarbeiterinnen und Mitarbeiter, die eine sichere und zuverlässige Zwischenlagerung an zahlreichen Standorten in Deutschland gewährleisten. Für diese 

verantwortungsvolle Aufgabe suchen wir Verstärkung. 

Verantwortung übernehmen. 

Anlagensicherheit 

Qualität 

Schutz 

Zuverlässigkeit 

Für unsere Organisationseinheit Betrieb an unserem Standort in Grafenrheinfeld suchen wir zum nächstmöglichen Zeitpunkt einen 

Abteilungsleiter/Leiter des Zwischenlagers (m/w/d) 

Wir bieten Ihnen die Möglichkeit, Ihr Wissen in ein interessantes und abwechslungsreiches Tätigkeitsfeld einzubringen. Spannende Projekte und 

verantwortungsvolle Aufgaben warten auf Sie. 

Fühlen Sie sich angesprochen? Unter https://karriere.bgz.de finden Sie die Details zu Ihren Aufgaben und Anforderungen sowie die Möglichkeit, 

sich direkt online zu bewerben. Bitte geben Sie dabei Ihre Gehaltsvorstellung sowie Ihren frühestmöglichen Eintrittstermin an. 

BGZ Gesellschaft für Zwischenlagerung mbH 

Personalwesen • Jens Abker • Frohnhauser Straße 67 • 45127 Essen • Telefon 0201 2796-1611 

www.bgz.de 


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


CALENDAR 514 

Calendar 

2020 

12.01. – 16.01.2020 

Power Plant Simulation Conference. Chattanooga, 

Tennessee United States, Society for Modeling & 

Simulation International, www.scs.org 

13.01. – 14.01.2020 

ICNPPS 2020 – 14 th International Conference on 

Nuclear Power Plants Safety. Zurich, Switzerland, 

Waset, www.waset.org 

20.01. – 21.01.2020 

6 th Central & Eastern Europe Nuclear Industry 

Congress 2020. Prague, Czech Republic, SZWGroup, 

www.szwgroup.com 

22.01.2020 

31.03. – 03.04.2020 

ATH'2020 – International Topical Meeting on 

Advances in Thermal Hydraulics. Paris, France, 

Société Francaise d’Energie Nucléaire (SFEN), 

www.sfen-ath2020.org 

19.04. – 24.04.2020 

International Conference on Individual 

Monitoring. Budapest, Hungary, EUROSAFE, 

www.eurosafe-forum.org 

20.04. – 22.04.2020 

World Nuclear Fuel Cycle 2020. Stockholm, 

Sweden, WNA World Nuclear Association, 

www.world-nuclear.org 

06.06. – 12.06.2020 

ATALANTE 2020. Montpellier, France, CEA, 

www.atalante2020.org 

08.06. – 12.06.2020 

20 th WCNDT – World Conference on 

Non-Destructive Testing. Seoul, Korea, EPRI, 

www.wcndt2020.com 

15.06. – 19.06.2020 

International Conference on Nuclear Knowledge 

Management and Human Resources Development: 

Challenges and Opportunities. Moscow, 

Russian Federation, IAEA, www.iaea.org 

15.06. – 20.07.2020 

Nuclear Fuel Supply Forum. Washington, D.C., USA, 

NEI, www.nei.org 

WNU Summer Institute 2020. Japan, World Nuclear 

University, www.world-nuclear-university.org 

10.02. – 14.02.2020 

01.09. – 04.09.2020 

37 th Short Courses on Multiphase Flow. Zurich, 

Switzerland, Swiss Federal Institute of Technology 

ETH, www.lke.mavt.ethz.ch 

IGORR – Standard Cooperation Event in the International 

Group on Research Reactors Conference. 

Kazan, Russian Federation, IAEA, www.iaea.org 

10.02. – 14.02.2020 

09.09. – 10.09.2020 

ICONS2020: International Conference on Nuclear 

Security. Vienna, Austria, The International Atomic 

Energy Agency (IAEA), www.iaea.org 

02.03. – 06.03.2020 

International Workshop on Developing a National 

Framework for Managing the Response to 

Nuclear Security Events. Madrid, Spain, IAEA, 

www.iaea.org 

08.03. – 12.03.2020 

WM Symposia – WM2019. Phoenix, AZ, USA, 

www.wmsym.org 

08.03. – 13.03.2020 

IYNC2020 – The International Youth Nuclear 

Congress. Sydney, Australia, IYNC, www.iync2020.org 

18.03. – 20.03.2020 

12. Expertentreffen Strahlenschutz. Bayreuth, 

Germany, TÜV SÜD, www.tuev-sued.de 

25.03. – 27.03.2020 

H2020 McSAFE Training Course. Eggenstein- 

Leopoldshafen, Germany, Karlsruher Institute für 

Technologie (KIT), www.mcsafe-h2020.eu 

29.03. – 02.04.2020 

PHYSOR2020 — International Conference on 

Physics of Reactors 2020. Cambridge, United 

Kingdom, Nuclear Energy Group, 

www.physor2020.com 

31.03. – 02.04.2020 

4 th CORDEL Regional Workshop on 

Harmonization to support the Operation and 

New Build fo NPPs including SMRs. Lyon, France, 

NUGENIA, www.nugenia.org 

KERNTECHNIK 2020. 

Berlin, Germany, KernD and KTG, 

www.kerntechnik.com 

05.05. – 06.05.2020 

10.05. – 15.05.2020 

ICG-EAC Annual Meeting 2020. Helsinki, Finland, 

ICG-EAC, www.icg-eac.org 

11.05. – 15.05.2020 

International Conference on Operational Safety 

of Nuclear Power Plants. Beijing, China, IAEA, 

www.iaea.org 

12.05. – 13.05.2020 

INSC — International Nuclear Supply Chain 

Symposium. Munich, Germany, TÜV SÜD, 

www.tuev-sued.de 

17.05. – 22.05.2020 

BEPU2020, Giardini Naxos. Sicily, Italy, NINE, 

www.nineeng.com 

18.05. – 22.05.2020 

SNA+MC2020 – Joint International Conference on 

Supercomputing in Nuclear Applications + Monte 

Carlo 2020, Makuhari Messe. Chiba, Japan, Atomic 

Energy Society of Japan, www.snamc2020.jpn.org 

20.05. – 22.05.2020 

Nuclear Energy Assembly. Washington, D.C., USA, 

NEI, www.nei.org 

31.05. – 03.06.2020 

13 th International Conference of the Croatian 

Nuclear Society. Zadar, Croatia, Croatian Nuclear 

Society, www.nuclear-option.org 

VGB Congress 2020 – 100 Years VGB. Essen, 

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

09.09. – 11.09.2020 

World Nuclear Association Symposium 2020. 

London, United Kingdom, WNA World Nuclear 

Association, www.world-nuclear.org 

16.09. – 18.09.2020 

3 rd International Conference on Concrete 

Sustainability. Prague, Czech Republic, fib, 

www.fibiccs.org 

16.09. – 18.09.2020 

International Nuclear Reactor Materials 

Reliability Conference and Exhibition. 

New Orleans, Louisiana, USA, EPRI, www.snetp.eu 

28.09. – 01.10.2020 

NPC 2020 International Conference on Nuclear 

Plant Chemistry. Antibes, France, SFEN Société 

Française d’Energie Nucléaire, 

www.sfen-npc2020.org 

28.09. – 02.10.2020 

Jahrestagung 2020 – Fachverband Strahlenschutz 

und Entsorgung. Aachen, Germany, Fachverband 

für Strahlenschutz, www.fs-ev.org 

26.10. – 30.10.2020 

NuMat 2020 – 6 th Nuclear Materials Conference. 

Gent, Belgium, IAEA, www.iaea.org 

09.11. – 13.11.2020 

International Conference on Radiation Safety: 

Improving Radiation Protection in Practice. 

Vienna, Austria, IAEA, www.iaea.org 

30.03. – 01.04.2020 

INDEX International Nuclear Digital Experience. 

Paris, France, SFEN Société Française d’Energie 

Nucléaire, www.sfen-index2020.org 

This is not a full list and may be subject to change. 

Calendar


Feature | Environment and Safety 

Development of Performance Shaping 

Factors for Human Error Reduction during 

Reactor Decommissioning Activities 

through the Task Analysis Method 

Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam 

In this study, we present the process of deriving Performance Shaping Factors (PSFs) and developing a classification 

system for them to account for human errors that may occur during the decommissioning activities due to the shutdown 

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

errors was performed on the Reactor Pressure Vessel Internal (RPVI) cutting work. Task Analysis was conducted on the 

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

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

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

classification system was developed. 

1 Introduction 

In recent years, industries have experienced a reduction in 

the rate of accidents due to technical problems because of 

the development of technologies for accident prevention. 

However, it is difficult to evaluate the reliability of such 

systems without considering the impact of human errors 

by their operators. In particular, according to real statistical 

accident data, about 60 to 90 % of all accidents are due to 

human errors, while the remainder is due to technical 

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

minor operator errors in real life can severely undermine 

the operational performance of the system. Furthermore, 

in large or crucial systems, such as large power plants 

( including nuclear power plants), air transportation, and 

railways, human error can lead to large and more harmful 

accidents; therefore, the reduction of human errors in the 

case of such systems requires significant attention [4]. 

At present, Korea is preparing to safely and economically 

decommission Unit 1 of the Kori Nuclear Power Plant, 

whose operation was permanently suspended in June 

2017. Various technological developments are being 

conducted for the decommissioning of Kori Unit 1 with the 

primary objective of ensuring the safety of human life and 

property and improving economic efficiency. Therefore, it 

is important to establish countermeasures to decrease 

human errors during this decommissioning, because these 

human errors not only hinder the safety of the operators, 

but also affect the economics of the project by causing 

delays in the decommissioning schedule. In order to do so, 

human error trends and influencing factors need to be 

identified through Human Reliability Analysis (HRA) [1]; 

in particular, HRA is used to identify, model, and quantify 

the likelihood of human error [3]. Figure 1 shows the 

flowchart for HRA that needs to be applied for the cutting 

of the Reactor Pressure Vessel Internal (RPVI) components, 

which is the decommissioning activity targeted in this 

study. 

It is important to conduct an HRA for the cutting 

operation of RPVI components, because they have the 

highest levels of radioactivity among all the other parts of 

the unit, and this operation is highly susceptible to human 

errors because of the use of remote equipment, among 

others. Therefore, in particular, the objective of this study 

is to obtain the Performance Shaping Factors (PSFs) that 

| Fig. 1. 

Procedure to perform the HRA in the decommissioning activity [5]. 

affect RPVI cutting and develop a classification system 

for these PSFs. 

Task characteristics, procedures, and information on 

the decommissioning of RPVI components were identified 

by reviewing related domestic and foreign literature as 

well as by examining previous cases of decommissioning of 

overseas nuclear power plants. In addition, task analysis 

for the decommissioning of the RPVI was conducted 

to identify probable human errors. First, potential PSFs 

were identified for this study based on a combination of 

the expected human errors from the detailed task 

analysis and the general PSFs from the evaluation of 

PSFs in nuclear facilities. Then, the selection criteria for 

the PSFs affecting the decommissioning activities were 

established. Finally, the PSFs were derived to construct 

the classification system. 

FEATURE | ENVIRONMENT AND SAFETY 515 

Feature 

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



Task 

Task Information 

and Requirements 

Task to Consider 

Working Time 

Teamwork and 

Communication 

Workload 

Task Support 

Workplace Factors 

Hazard Identification 

Expected PSFs 

2 Task Analysis 

Task Analysis (TA) is the initial step in the case of human 

error assessment. It provides the characteristics of different 

tasks, their vulnerabilities, and properties by understanding 

the objectives, performance methods, scopes of 

the tasks, and procedures involved [6]. In addition, TA 

helps eliminate the conditions that can cause errors before 

they occur by providing detailed information about the 

tasks as well as other information for predicting and 

preventing errors. 

2.1 TA Method 

Although several TA methods exist, considering the 

physical characteristics (complexity) and Human Machine 

Interface (HMI) aspect (using remote devices) of the 

decommissioning activities, we utilized the Hierarchical 

TA (HTA) method. 

HTA is a systematic and detailed TA method that is used 

to achieve task objectives. It is appropriate for not only 

identifying detailed task configuration and conditions, but 

also expressing complex task steps as a simple, hierarchical 

structure. Furthermore, the HTA method was used because 

it can easily describe the characteristics of the work 

involved and identify significant information about the 

Title 

| Tab. 1. 

Details of tasks to be performed for TA. 

• Task influencing factors 

(cutting size, number of cutting operations, precision, etc.) 

• Task output requirements 

• Record feedback to indicate adequacy of action taken 

• Alarms and warnings 

• Actions to be taken 

• Equipment needed (type, size, performance, 

required utility, equipment usage constraint, etc.) 

• Task frequency and required accuracy 

• Physical position of the operator 

(standing, sitting, squatting, etc.) 

• Biomechanics 

• Movement 

(lifting, pushing, rotating, pulling, swaying, etc.) 

• Force required 

• Unit working time based on “work contents” 

• Additional work hours taking into account “work support” 

and “workplace environmental conditions” 

• Number of work shifts and workers per work shift 

• Cooperation required between the teams performing 

the work 

• Personal communication for monitoring or 

taking control actions 

• Cognitive workload 

• Physical workload 

• Overlap of task requirements (serial versus parallel task 

elements) 

• Special and protective clothing for work 

• Job aids, procedures, or reference materials needed 

• Required auxiliary tools and equipment 

• Ingress and egress paths to the work site 

• Workspace required to perform the task 

• Typical workplace environmental conditions 

(e.g. lighting, temperature, noise, etc.) 

• Work breaks taking into account “work support” and 

“ workplace environmental conditions” 

• Identify work-related hazards, 

e.g. potential personal injury hazards 

Examples include: 

• Stress 

• Time pressure (critical path operations) 

• Extreme environmental conditions 

• Reduced staffing 

HMI, communication and decision making processes, as 

well as possible accidents. The HTA method involves 

describing the manner in which tasks need to be performed 

after establishing their overall objectives and classifying 

them into their sub-tasks [6]. 

Using a tabular format to perform HTA allows one to 

express complex tasks that require significant skill in a 

suitable manner, because one can include detailed notes, if 

necessary. In this study, we comprehensively reviewed 

various items, such as HMI, Communication, Time, and 

Accident, for the decommissioning activity in a tabular 

format; this is shown in Table 1 [5]. 

2.2 Target Decommissioning Activity for TA 

In general, one of the most challenging tasks during plant 

decommissioning is believed to be the removal of the 

highly radioactive internal components of the reactor pressure 

vessel (RPV); this is true for Kori Unit 1 as well. In 

addition, another reason that this is one of the most 

difficult activities is because these radioactive components 

must be dismantled and cut underwater owing to the 

severe radiological conditions of the RPVI components 

[7-8]. Therefore, it is recommended that the reactor 

internals be removed as early as possible in the plant 

dismantling sequence, so that these water systems and 

their associated support systems can be released for 

decommissioning, which minimizes the costs associated 

with maintaining these systems in operation after 

permanent plant shutdown [8]. 

The cross-section of the RPV with the primary internal 

components at Kori Unit 1 is shown in Figure 2. As can be 

seen from the figure, the internal structures adjacent to the 

core barrel active region are the most highly activated, and 

in most cases, include intermediate level waste components 

that might require removal prior to the disposal of the 

remainder of the RPV and reactor internal components 

[7]. Thus, in this study, this RPV internal segmentation 

activity is selected from among the various dismantling 

activities in Kori Unit 1. 

Furthermore, in this study, the TA was performed for 

the most complex and highly radioactive RPVI cutting task 

among various disassembly activities by using the HTA 

method. The sequence of operations for each sub-activity 

in this target task is listed in Table 2. 

2.3 TA Results 

The TA for the RPVI Dismantling Activity was performed as 

per the activities listed in Table 2 based on the items listed 

in Table 1. In our study, this analysis was performed for 

each of the 10 sub-activities. The summary of the TA for 

the RPVI Dismantling Activity is given as follows. 

p In general, in the decommissioning of nuclear power 

plants, the cutting of parts in the RPVI is the most 

complicated and difficult task during the dismantling 

process. Therefore, the influence on the internal factors 

of the workers was evaluated in order to have a relatively 

high value in terms of operator internal response. 

p This kind of dismantling operation, which is complex 

and requires accuracy and reliability, is significantly 

influenced by the internal and external characteristics 

of the worker. Therefore, sufficient education and 

training is required. However, as the driving principle 

and operation method of these cutting equipment and 

accessories ( cutting equipment, remote control device, 

display, auxiliary equipment, etc.) are not complicated 

and operation is relatively simple, an operator is not 

required to have considerable experience in using them. 

Feature 



No Sub activity Activity Description 

1 Control Rod Guide Tube Upper Area Cutting 

and Packaging 

2 Control Rod Guide Tube Lower Area Cutting 

and Packaging 

1) Lift up the Upper Internal Assembly and fix it to the turntable 

2) Cut the control rod guide tube upper area 

3) Transfer and package the cut section to a storage container 

1) Cut the control rod guide tube lower area 

2) Transfer and package the cut section to a storage container 

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

2) Cut the Upper Core Plate fixed to the turntable 

3) Fix the Upper Core Plate to the turntable and cut the Upper Support Plate 

4) Transfer and package the cut pieces to a storage container 

4 Baffle Fixed Bolt Head Cutting 1) Install the mechanical drill for Baffle Separation 

2) Cut the Baffle Fixing Bolt Head by placing the mechanical drill inside the RPV 

5 Baffle Cutting and Packaging 1) Lift the Baffle and remove it from the Former 

2) After fixing it to the turntable, cut the Baffle 

3) Transfer and package the cut Baffle pieces to a storage container 

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

2) Rotate the turntable to cut the Lower Internal Assembly 

3) Lift the upper area of the Lower Internal Assembly into the Vessel 

7 Lower Internal Structure Assembly Cutting 

and Packaging 

8 Thermal Shield Separation, Cutting and 

Packaging 

1) Cut the Instrument Nozzle from the Core Support Structure Assembly 

2) Transfer and package the cut nozzle to a storage container 

3) Cut and package the tie plate fixed to the Turntable 

4) Fix the Lower Core Plate to the Turntable and cut it 


1) Lift the upper area of the Lower Internal Assembly and fix it to the turntable 

2) Separate the Thermal Shield by cutting the Bolt Head 

3) Release the removed Core barrels from the Thermal Shield inside the vessel 

4) After fixing the Thermal Shield to the Turntable, remove the Irradiation Specimen Guide 

5) Cut the Turntable Thermal Shield Upper and Lower Panels 

9 Former Separation 1) Lift the Core Barrel to the turntable and fix it 

2) Cutting the Former fixing bolt head outside the Core Barrel 

3) Separate the Former from the Core Barrel 

4) Transfer and package the separated Former to a storage container 

10 Core Barrel Cutting and Packaging 1) Fix the Core Barrel to the turntable and cut it 

2) Temporarily release the cut Upper Core Barrel in the Vessel 

3) Segment the Lower part of the Core Barrel fixed to the turntable 


5) Repeat the procedure for the cutting the remaining Core Barrel 


| Tab. 2. 

Task Description of RPVI Dismantling Activity. 

p Considering the characteristics of the work (equipment 

and facilities, the object to be cut, and clothing for 

radiation protection in the work environment), a 

detailed work plan must be established in advance. 

Further, as this cutting work is time-consuming, the 

psychological and physical influences that the supervisor 

and the worker can receive are considerable. 

p If the cutting activity is dangerous, takes a long time, 

and has associated time constraints, it should be 

performed during the day/night time. In this case, 

various difficulties (such as break time, clothing 

discomfort, and physio logical factors) are generated. 

Because these difficulties have a significant impact on 

the internal and external factors affecting the worker, 

much cooperation and communication is required 

between the worker and the supervisor in this working 

environment. 

p The radiation and the physical environment of the 

workplace are the major risk factors for the workers, 

and the influence of these working environments on the 

internal and external factors of the workers was 

con siderable. 

p Nuclear decommissioning work is not a frequently 

performed task. Therefore, workers may have insufficient 

experience and education/training. Therefore, 

after the decommissioning activity has been carried 

out, it is necessary to feedback the results of the work to 

be reflected in the necessary work procedures and to be 

managed as experience data. 

| Fig. 2. 

Internal components for the RPV in Kori Unit 1 [7]. 

3 PSFs in the Decommissioning Activities 

Because the workers’ activities are the fundamental factors 

that renders the system vulnerable, it is necessary to 

identify, model, and quantify the possibility of human 

error using HRA [3]. In particular, the nominal Human 

Error Probability (HEP) used in the HRA is estimated 

Feature 




based on the workers’ activity, most often based on the 

PSFs. This is because PSFs characterize the important 

aspects of human error and provide numerical criteria for 

adjusting the average HEP level [9]. Therefore, the key 

step in the HRA for decommissioning activities of a nuclear 

power plant is to select appropriate PSFs, develop the 

classification system, and quantify their impacts on the 

decommissioning activity. 

3.1 Criteria for selecting PSFs 

When selecting PSFs that affect the decommissioning 

activities of nuclear plants, it is important to carefully 

select all the important situational factors that affect the 

hazard level so as to not miss any of the factors. In addition, 

it should be ensured that there is no overlap in the meaning 

and scope of the PSFs; furthermore, the factors should be 

selected based on their actual effect on human error 

analysis. Therefore, the PSFs are selected for the HRA of 

the decommissioning activities of the nuclear power plant 

considering the following criteria. 

p The PSFs should be selected to appropriately reflect the 

work characteristics. Therefore, the TA results of the 

work should be fully reflected by the PSFs. 

p The internal and external factors that might affect 

the workers’ performance should be comprehensively 

considered. 

p When a worker encounters an abnormal event, the 

internal factors should be included in the PSFs 

appro priately to account for the reactions that occur 

naturally in workers, including physical, cognitive, and 

emotional factors. 

p The external factors should comprehensively include 

those that directly or indirectly affect the workers’ 

response to performance, i.e., the business organization 

and work environment factors, among others. 

p Based on previous research results, in our case, the 

PSFs will consist of three levels (namely, Level 1, Level 

2, and Level 3). Though there are direct and indirect 

dependencies between the PSFs in Level 1 and Level 2, 

the PSFs are selected in a manner that there is no 

dependency between the Level 3 PSFs. 

p In practice, the PSFs are selected based on factors 

that directly affect the trends of the ongoing events. 

Therefore, factors such as workers’ training and work 

shifts, among others, indirectly affecting the work 

performance of the workers, are excluded from the PSF 

Classification Direct effect Indirect effect 

Personal Factors 

System factors 

Task factors 

| Tab. 3. 

Derived PSF factor. 

• Duration of mental stress 

• Mental tension 

• Pain or discomfort 

• Hunger or thirst 

• Emotional state 

• Duration of physical stress 

• Disruption of circadian rhythm 

• Lack of sleep 

• Work hours 

• Shift rotation 

• Suddenness of onset 

• State of current practice or skill 

• Motivation and attitudes 

• Personality and 

intelligence variables 

knowledge of required 

performance previous 

training/experience 

• Sensory deprivation 

• Distractions (noise, glare, 

movement, flicker. color) 

• Complexity 

• Movement constriction 

• Workplace layout 

• Threats of failure 

• Lack of physical exercise 

• High jeopardy risk 

• Conflicts of motives 

about job performance 

selection because they are considered to be inherent in 

the other PSFs themselves. 

p The PSFs should have clear definitions so that their 

meanings and roles do not overlap. Therefore, as much 

as possible, the scale of the PSFs should be reduced by 

grouping all the PSFs and reducing the number of tasks 

involved. 

3.2 Selection and definition of PSFs 

Based on our review of existing literature related to the 

work performances of individual workers from various 

industrial fields, we observed that it is not easy to find 

a consensus on the factors influencing the workers’ 

performance. However, considering the results of these 

previous studies, it is deemed that the possibility for human 

error can be determined based on the degradation of the 

workers’ human error factor, i.e., their task performance. 

The Institute of Nuclear Power Operations (INPO) 

presents 85 error precursors that can lead to possible 

human errors considering the business requirements, personal 

abilities, work environments, and human nature in 

terms of the operation of the nuclear power plants [10]. 

These precursors are considered as the risk variables for 

human errors made by workers. 

Therefore, it is realistic to utilize an Error Precursor, 

i.e., PSF, that affects work performance. While estimating 

the probability of human error, it is desirable to add 

psychological factors that cannot be directly managed 

after the possibility of human error for manageable 

factors is reduced through safety management or accident 

prevention activities. The PSFs extracted for use in the 

human error assessment model in this study are listed 

in Table 3 below. 

3.3 Development of a classification system 

for PSFs 

TA was conducted to derive the PSFs for RPVI cutting, 

which is one of the primary tasks involved in the 

decommissioning of nuclear power plants. Because it is 

difficult to select PSFs such that their meanings and roles 

are clearly delineated, the PSFs are classified into three 

major categories, which are defined by their task analysis 

results. In other words, the human factors of the workers 

themselves, the operating system factors related to the 

work, and the ergonomic factors linking the worker and 

the dismantling activities were suitably classified. 

In particular, an important aspect of the TA is “Human,” 

i.e., a worker who performs the task of decommissioning 

the RPV and can cause human errors. It is noteworthy 

that human factors are important to consider not only 

in the field of nuclear power generation, but also, in 

other industrial fields, such as the railways and aviation 

[11-16]. 

p Supervisors and workers might be affected psychologically 

while performing the tasks described above. 

Because the cutting process takes considerable time, it 

is important to consider the various stresses and 

emotional conditions that supervisors and workers 

might experience. 

p In particular, the work environment and conditions for 

the cutting task in normal or abnormal situations affect 

the workers’ physical and physiological factors. Thus, to 

perform the RPVI cutting tasks, wearing the appropriate, 

protective work clothes and protective cap 

and having good physical and health conditions to 

carry or handle the equipment are considered important. 

Furthermore, while performing the tasks, the 

Feature 



physiological factors of the workers, such as fatigue, 

hunger, and excretion, is equally important. 

p When the nuclear power plant is dismantled for the 

first time, there might be a lack of personal skills and 

experiences, because the workers are not accustomed 

to the decommissioning work. Therefore, information 

about the complete decommissioning task and equipment 

used for each task should be taught through 

training and education. 

Next, there is a need for an operating system that can be 

managed, supervised, and overseen to efficiently perform 

the decommissioning tasks. Therefore, the “operation” 

elements, such as organization, including the supervisors 

and workers; task management for managing work 

schedules and workload efficiently; and documents 

include detailed information on decommissioning, such as 

procedure and scenarios, are important. 

p In order to ensure that the cutting work is performed 

smoothly, the supervisors and workers need to form 

appropriate teams. Organizational factors, such as the 

objectives of these team, necessary team-building 

elements (cooperation, role, etc.), and precise and 

detailed communication of the supervisors’ judgement/decisions 

to the workers are important. 

p Next, the tasks should be well-designed including the 

preparation, methods and strategy, as well as processes 

involved in the tasks. Because the work is actively 

con ducted throughout the day, the work assignment 

Level 1 Level 2 Level 3 

Human 

Operation 

Ergonomic 

System 

Psychological 

State 

Physical 

State 

Performance 

Capability 

Organizational 

Factors 

Task Management 

Procedure and 

Information 

HMI 

(Human Machine 

Interface) 

Workplace Design 

Workplace Physical 

Environment 

Stress 

and shift work schedule should be coordinated and 

managed. 

p Finally, because supervisors and workers understand 

and perform their work based on the procedures and 

scenarios related to the decommissioning of the RPVI 

components, information on regulations, equipment 

and facilities, as well as accuracy and details of the 

procedures are important. 

Furthermore, because the cutting work is human- centric, 

the surrounding situations and conditions must be 

designed in a manner that is suitable for human beings. 

Therefore, “ergonomic system” factors, such as worker 

interaction with the necessary equipment and facilities, 

which are based on the decommissioning task characteristics 

and requirements for RPVI components, workplace 

design based on the path of the supervisor and worker and 

task types, and optimal workplace environment for 

workers, are important points to consider. 

p The systems and equipment the supervisors and 

workers operate must be interactive to ensure that they 

are not difficult to use to ensure the convenience and 

safety of the workers. 

p Because the RPV is located under water and is not 

directly visible and accessible to humans, the cutting 

processes under the RPVI decommissioning task 

are generally performed using remotely controlled 

devices. Therefore, the configuration, condition, 

and per formance of these remote systems, such as 

Emotional State (Excitement, Boredom, Accomplishment, Frustration, Dissatisfaction) 

Safety Awareness 

Fatigue 

Physical Capability 

Discomfort 

Task Knowledge and Memory (Diagnoses, Action, Results, Information) 

Experience 

Personal Capability 

Overall Planning 

Supervision of Management including Decision Making 

Team Factors(Roles, Coordination, Risk Management, Atmosphere, Assisting Other) 

Work Process Design (considering Task Characteristic) 

Workload Management 

Problem Identification and Solution 

Communication Availability and Quality 

Procedure and Information Availability 

Procedure and Information Complexity 

Procedure and Information Accuracy and Completeness 

Procedure and Information Feedback and Recency 

Interaction Element (Menu, Push Button, Direct Manipulation, Special Symbol, Shape Type) 

Familiarity of Equipment and Facility 

Complexity of Equipment and Facility 

Maintenance 

Physical Access to Work Items 

Warning Sign (Alarm Location, Quantity, Intensity, Importance and Easy of Identification) 

Arrangement of Functional Areas 

Safety Device 

Noise 

Lighting 

Temperature 

Radiation Level 


| Tab. 4. 

PSF Classification System in the Reactor Decommissioning Activity. 

Feature 




remote control, camera, and display should be well 

established, and designers should consider functional 

spatial arrangements and accessibility in order to use 

them. 

p An optimal working environment is required to ensure 

the safety of workers during the decommissioning 

work. In particular, it is important to establish an 

optimal physical environment for the workers by 

checking for factors, such as the noise of the equipment, 

temperature experienced by the workers wearing 

protective clothing, level of radiation exposure, cleanliness 

of the workplace, and turbidity of the water in 

which the cutting operation is being performed. 

In particular, the PSFs are classified into three categories 

per the level of detail and importance required for the 

decommissioning activities. Level 1 includes “Human 

Factors,” “Operation Factors,” and “Ergonomic System 

Factors.” Next, Level 2 of Human Factors includes the 

basic elements responsible for human error, namely 

the psychological and physical state and performance 

capability of the supervisors and workers. Level 2 of 

Operation Factors is composed of Organizational Factors, 

Task Management, Procedures, and Information. Finally, 

Level 2 for human-oriented Ergonomic Factors is composed 

of HMI, Workplace Design, and Workplace Physical 

Environment. The PSFs under Level 3 are selected in a 

manner so as to not overlap with each other; in particular, 

these include a total of 44 factors as shown in Table 4. 

4 Concluding Remarks 

In this work, we study and present the methodology for 

HRA for the cutting task of the Nuclear RPVI components 

in order to reduce human errors during the nuclear decommissioning 

activities. In order to do so, the HRA implementation 

procedures were reviewed to identify the components 

of the HRA and the most important TA was performed 

in preparation for the HRA. The task characteristics, 

procedures, and information on the decommissioning 

of the RPVI components were identified by reviewing relevant 

literature in the field and examining experiences in 

decommissioning at overseas nuclear power plants. 

In addition, PSFs required for the nuclear decommissioning 

activities were identified using TA results and 

PSFs selection criterion for RPVI cutting (10 tasks in total). 

The results of the PSF review applied to nuclear facilities 

and types of human errors identified through the detailed 

TA were synthesized. Furthermore, the selection criteria 

for the PSFs affecting the decommissioning activities were 

set. Finally, PSFs were selected and classified using a 

classification system. 

In a future study, we will examine the interrelationship 

among the PSFs and consider methods for assessing 

the PSFs. Moreover, we will develop a framework to model 

the mutual influences that exist among the PSFs with 

appropriate consideration of the relationships and 

dependencies among them. The collection of experience 

data for decommissioning nuclear power plants will also 

reduce the uncertainty in the information used to perform 

HRA. 

References 

[1] Madonna, M., et al.: Il Fattore Umano Nella Valutazione Dei Rischi: Confronto Metodologico Fra 

Le Tecniche Per L’analisi Dell’affidabilità Umana. Prevenzione Oggi. 5 (n. 1/2), 67–83 (2009). 

[2] Hollnagel, E.: Cognitive reliability and error analysis method (CREAM). Elsevier, (1998). 

[3] Griffith, C.D., Mahadevan, S.: Inclusion of fatigue effects in human reliability analysis. Reliability 

Engineering & System Safety, 96 (11), 1437–1447 (2011). 

[4] Lee, Y., et al.: Research Activities and Techniques for the Prevention of Human Errors during the 

Operation of Nuclear Power Plants. Journal of the Ergonomics Society of Korea, 30 (1), 75-86 

(2011). 

[5] O’Hara, J. M., et al.: Human Factors Engineering Program Review Model, US NRC, NUREG-0711, 

Rev. 3, 2012 

[6] Center for Chemical Process Safety : Guidelines for preventing human error in process safety 

(1994) 

[7] Byung-Sik Lee: Optimization of reactor pressure vessel internals segmentation in KoreaATW 

Vol.62 (2017), Issue 11, 654~658, November, 2017 

[8] Boucau, J., et.al.: Best practices for preparing vessel internals segmentation projects. No. 

NEA-PREDEC--2016. February 16–18, Lyon, France (2016). 

[9] Boring, R.L.: Modelling human reliability analysis using MIDAS. In: International Workshop on 

Future Control Station Designs and Human Performance Issues in Nuclear Power Plants (2006). 

[10] Human performance reference manual, Institute of Nuclear Power Operations, INPO-06-003, 

2006 

[11] Chang, Y. H. J., Mosleh, A.: Cognitive modeling and dynamic probabilistic simulation of operating 

crew response to complex system accidents. Part 2: IDAC performance influencing factors model. 

Reliability Engineering & System Safety, 92(8), 1014-1040 (2006). 

[12] Blackman, H. S., Gertman, D. I., Boring, R. L.: Human error quantification using performance 

shaping factors in the SPAR-H method. In Proceedings of the human factors and ergonomics 

society annual meeting (Vol. 52, No. 21, pp. 1733-1737). Sage CA: Los Angeles, CA: SAGE 

Publications (2008). 

[13] Li, P. C., Chen, G. H., Dai, L. C., Zhang, L.: A fuzzy Bayesian network approach to improve the 

quantification of organizational influences in HRA frameworks. Safety science, 50(7), 1569-1583 

(2012). 

[14] Jung, K., Byun, S., Kim, J., Heo, E., Park, H.: An empirical study on evaluation of performance 

shaping factors on AHP. Journal of the Ergonomics Society of Korea, 30(1), 99-108 (2011). 

[15] Mindock, J.: Development and Application of Spaceflight PSFs for HRA. PhD Thesis , University of 

Colorado at Boulder (2012). 

[16] Baek, D., Koo, L., Lee, K., Kim, D., Yoon, W., Jung, M.: Taxonomy of PSFs for human error analysis 

of railway accidents. Business Administration, Hanyang University (2007). 

Authors 

Byung-Sik Lee 

Hyun-Jae Yoo 

Chang-Su Nam 

Dankook University 

119, Dandae-ro, Dongnam-gu, Cheonan-si 

Chungnam, 31116 

Republic of Korea 

Acknowledgments 

This work was supported by the National Research 

Foundation of Korea (NRF), granted financial resource 

from the Ministry of Science, ICT and Future Planning 

(No. 2017M2A8A5015148 and No.2016M2B2B1945086), 

Republic of Korea. 

Feature 



Root Causes of the 

Three Mile Island Accident 


The accident at Unit 2 of the Three Mile Island nuclear power plant, at that time operated and partly owned by 

Metropolitan Edison Company, occurred 40 years ago, on March 28, 1979. Following the accident, two major 

investigations were conducted, one by the President’s Commission on the Accident at Three Mile Island [1], appointed 

by President Carter, and the other by the Nuclear Regulatory Commission’s Special Inquiry Group. [2] The investigations 

documented the timeline of the accident and the availability and failure of equipment, and addressed operator actions 

during the accident, the training of operators, and NRC procedures that applied to the event. The design process for the 

plant and the designer’s responsibilities, including the plant’s safety analysis, were not addressed. Many additional 

studies and papers have been published over the past 40 years, none of which have addressed the design process or the 

safety analysis of the plant. The only effort specifically addressing the design of the plant and responsibility for the 

accident was Metropolitan Edison’s lawsuit against Babcock & Wilcox (B&W), the designer of the plant. A trial began 

but was terminated, and the case was settled out of court. The court records are sealed; information is not available. 

More than 10 years prior to the TMI-2 

accident, B&W was designing its first 

nuclear power plant. In the designation 

of safety systems and in the safety 

analysis of the plant, there were two 

relatively minor but important omissions. 

These omissions turned out to 

be the root causes of the accident. If 

just one of them had been corrected 

during the intervening years, the 

accident would have been avoided. 

The TMI design was reviewed by 

utilities purchasing plants from B&W 

and by the NRC. The omissions 

remained undetected. The safety role 

of the pilot-operated relief valve 

(PORV) and the PORV block valve 

were not fully appreciated. The manufacturer 

of the PORV was not notified 

of the valve’s safety function, namely 

that it has to be able to close after 

being exposed to accident loads. [2] 

Also, the plant’s safety analysis report 

(SAR) did not address loss-of-coolant 

accidents (LOCA) initiated by very 

small breaks. Unfortunately, the plant 

responds very differently to an event 

initiated by a stuck PORV than to the 

small-break events presented in the 

SAR. At the time, this was unknown. 

Lessons learned from the omissions 

in the TMI design are timely today, 

when new types of reactors, such 

as small modular reactors, are on the 

drawing board. The designers of these 

new systems can learn from the TMI 

experience. 

The initiating event 

Operators attempting to clean a 

condensate polisher tripped the steam 

generator feedwater pumps. Then, 

the plant safety system tripped the 

turbine. The turbine was no longer 

removing heat from the reactor coolant 

system (RCS), the temperature 

and pressure of the RCS started rising 

rapidly, and the PORV opened, as 

designed. 

Upon shutdown of the feedwater 

pumps, the plant’s safety system 

turned on the emergency feedwater 

pumps. Due to a maintenance error, 

both emergency feedwater block 

valves, which are supposed to be open 

when the plant is operating, were 

closed, so no emergency feed-water 

reached the steam generators. The 

closed valves caused the RCS to heat 

up faster than in the case of a normal 

turbine trip, and the PORV was 

exposed to a larger load than normal, 

most likely a heavy two-phase flow 

(steam and water mixture) or water 

discharge. Thus, the closed valves 

could have played a role in causing 

the accident. This possibility is not 

addressed in the literature. 

As RCS pressure increased, the 

reactor protection system shut the 

reactor down, after which the RCS 

pressure dropped. The PORV should 

have closed, but instead it stuck open, 

and the plant faced a LOCA. The 

obvious question is, “Why did the 

PORV fail to close?” 

Designers of nuclear power plants 

have a dual responsibility. They must 

design the plant not only for normal 

operation of generating electricity, 

but also for safe performance in case 

of events that might occur during the 

lifetime of the plant and in case of 

postulated accidents. 

Components of systems that have 

both an operating function and a 

safety function have to be identified 

and designed to perform both functions 

in a reliable manner. The PORV 

is a good example of such a component. 

During normal operation, the 

PORV maintains RCS pressure below 

specified limits by opening and closing 

and by discharging steam from the 

pressurizer. During abnormal events, 

as in the TMI-2 case, the PORV could 

be discharging two-phase flow or 

water. The valve must be designed to 

perform its safety function – namely, 

to close following a two-phase flow or 

water discharge. Apparently, this was 

not the case at TMI. The PORV was 

not designed to perform its safety 

function. The purchase order failed to 

specify this requirement, and the 

supplier of the valve did not know that 

the valve had a safety function and 

that it had to close following twophase 

flow or water discharge. [2] 

Designers are also responsible for 

incorporating operating experience 

into their design. Prior to the TMI-2 

accident, PORVs failed to close seven 

times at B&W plants. [2] Despite 

this record, the PORV itself was not 

modified or replaced. Instead, an 

indicator light was installed to show 

whether the block valve upstream of 

the PORV had received a signal to 

close, but there was no indication in 

the control room that the valve had 

actually closed. 

Reprinted with 

permission from the 

March 2019 issue of 

Nuclear News 

Copyright © 2019 by 

the American Nuclear 

Society 

| From right to left: President Jimmy Carter, Pennsylvania Gov. Richard 

Thornburgh, and the NRC’s Harold Denton tour the TMI-2 control room on 

April 1, 1979. Photo: The Jimmy Carter Presidential Library and Museum 

521 

ENVIRONMENT AND SAFETY 


Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy


ENVIRONMENT AND SAFETY 522 

PORVs opened relatively frequently 

on B&W-designed pressurized 

water reactors. The thermal hydraulic 

design of the reactor core was closer 

to acceptable limits than other PWR 

cores, and the amount of water 

contained in the secondary side of 

the steam generators was very small – 

only 25 percent of some other PWRs’ 

water content. [2] These differences 

made the system react faster to 

changes. With quick plant response, 

the PORV came into action relatively 

frequently. More frequent use of the 

PORV led to more frequent failures. 

The PORV failure at TMI-2 was the 

eighth at a B&W plant, more than 

an order of magnitude higher than 

PORV failures with other reactor 

designs. [2] 

The NRC has specific requirements 

for equipment related to safety. Equipment 

essential to accident mitigation 

and equipment whose failure can 

cause or aggravate an accident are 

considered “safety related.” A stuckopen 

PORV causes a breach in the 

boundary of the RCS, creating a 

LOCA. Among postulated accidents, 

LOCAs are considered to be the 

most serious, and therefore they 

receive special attention. Nuclear 

power plants are designed with three 

barriers to protect the public from 

radioactive material release: The 

fuel is enclosed in a sealed cladding, 

the reactor core is within the closed 

RCS, and the RCS is covered by a 

containment building. Among all 

postulated accidents, there is only 

one – the LOCA – where two of the 

barriers are predicted to be damaged. 

In the case of a LOCA, the event itself 

breaches the RCS, and the predicted 

consequences of the accident are 

expected to damage some of the 

fuel cladding. Protection of the public 

is reduced to a single barrier, the 

containment building. Furthermore, 

valve failures are more likely than 

pipe breaks. Thus, the most likely 

LOCA is the stuck-open PORV. 

Surprisingly, the PORV was not 

identified by the designer as safetyrelated 

equipment. The design was 

reviewed by Metropolitan Edison 

and evaluated by the NRC. Neither 

objected to the PORV not being 

designated as safety related, and 

the NRC approved the construction 

permit application. Had the PORV 

been designated as safety-related 

equipment, it would have had to 

meet reliability requirements and be 

tested under accident conditions. If 

the TMI-2 PORV had been tested, it 

most likely would not have passed. 

Following the accident, the manufacturer 

of the valve stated that the 

TMI-2 PORV was not qualified to close 

following a two-phase flow or water 

discharge. [2] If the PORV had been 

designated as safety related, it would 

have been replaced or modified. 

The reason given for not designating 

the PORV as safety related was 

the presence of a block valve upstream 

of the PORV. If the PORV is stuck 

open, the block valve can be closed, 

terminating the accident. Thus, the 

block valve is essential for the mitigation 

of a PORV failure accident, and 

it is also considered safety-related 

equipment. It must have automatic 

safety-grade actuation initiated from 

the stuck-open PORV or, if the initiation 

is manual, safety-grade position 

indication must be available in the 

control room with sufficient time 

for operator action. Neither of these 

conditions existed at TMI-2. 

Consequences of PORV failure 

Part of the designer’s responsibility is 

to conduct a complete and detailed 

safety analysis of the plant. The 

analysis must include transients that 

might occur in the plant. The analysis 

of transients must show that continued 

operation of the plant following 

these events is justified. The 

plant’s safety analysis also has to 

address all potential accidents, both 

system failures and operator errors 

that the plant could be subject to, 

unless they are considered to be 

extremely unlikely (severe accidents). 

It is the designer’s responsibility to 

identify all accident types specific 

to the design of the plant. In the case 

of water-cooled reactors, one of these 

accident types is a breach in the RCS 

– a LOCA. 

For PWRs such as TMI-2, it is an 

NRC requirement that a complete 

spectrum of breaches in the RCS be 

analyzed, starting from the doubleended 

break of the largest pipe in the 

RCS down to the break size that the 

makeup water system can keep up 

with. Unfortunately, it was not emphasized 

that a breach in the system 

includes stuck-open valves if the 

valve’s discharge area is within the 

size range of the postulated accident. 

The PORV falls within the size range. 

Complete spectrum also means all 

possible break locations. The consequences 

of a stuck-open valve on 

the top of the pressurizer could be 

different from a same-size break at a 

lower elevation. 

Typically, prior to the TMI accident, 

the large-break LOCA analysis 

| A six-page special report – Nuclear News’s 

initial coverage of the TMI accident – 

was mailed separately to subscribers and 

ANS members in April 1979. 

included break sizes in both the hot 

and cold legs of the RCS, starting 

from a double-ended break down 

to a 0.5 square-foot break. Usually, 

the consequences were most severe 

at one of the larger breaks. From 

there on, smaller sizes resulted in 

more favorable consequences. The 

small-break LOCA analysis ran from 

0.5 square foot down to about 

0.1 square foot. The trend was the 

same; smaller breaks had less severe 

consequences. Breaks even smaller 

were not analyzed for two reasons: 

(1) the calculations ran long on the 

computer and the analyses were 

expensive, and (2) the trend was 

already established. Instead, the 

assumption was made that the 

trend would continue down to the 

smallest required size. Also, smallbreak 

LOCA analysis was assumed 

to be independent of break location. 

Thus, breaks less than 0.5 square 

foot were not analyzed at different 

locations, and breaks less than 

0.1 square foot were not analyzed 

at all. 

The safety analysis of the plant 

serves many purposes. It provides 

both the designer and the operator of 

the plant with an understanding of 

how the plant responds to a specific 

event or accident, indicates potential 

damage if mitigating actions are 

not taken, guides the designer in the 

design of the needed safety systems, 

and provides information for training 

the operating staff. The analysis 

shows how reactor operators can 

recognize a specific event and what 

actions they must take and provides 

the needed information for the 

preparation of emergency procedures. 

The results of the analysis also show 




compliance with applicable regulations. 

It shows that potential damage 

has been mitigated and the safety 

of plant personnel and the public is 

ensured. 

The TMI-2 LOCA analysis was 

performed by the designer. It was 

rather elaborate but was incomplete 

in the sense that it failed to show that 

very small breaks behave differently 

and have more serious consequences 

than small breaks. As part of the 

TMI-2 licensing process, no LOCA 

analysis was performed by the 

designer, and no LOCA analysis was 

submitted to the NRC by the utility 

for a break size anywhere close to 

the size of the discharge opening of 

the PORV (about 2 square inches). No 

analyses were performed for any size 

break at the top of the pressurizer or 

for a LOCA caused by a stuck-open 

PORV. Unfortunately, as was learned 

from the TMI accident and from 

analysis performed after the accident, 

a stuck-open PORV. 

LOCA is very different from the 

small-break LOCA analyses presented 

for TMI-2. Analysis of smaller breaks 

showed that the trend reverses 

and the consequences increase with 

decreasing break size. The plant 

responds differently, reliance on 

safety systems and instrumentation 

changes, and different operator actions 

are required. The consequences 

of a PORV failure are even more 

different. RCS pressure can drop while 

the water level in the pressurizer is 

rising. Void can form in the reactor 

core and accumulate at high locations 

of the RCS while the pressurizer water 

level is high. Furthermore, the water 

level can drop below the top of the 

core, resulting in core damage, while 

the pressurizer water level is still high. 

Obviously, pressurizer water level 

indication is not a useful tool for the 

handling of this accident. 

A special design feature of B&W 

plants further aggravates this effect. 

The pressurizer surge line is designed 

with a loop seal to prevent steam from 

entering the pressurizer. Eliminating 

steam flow to the pressurizer prevents 

water level drops in the pressurizer, 

keeping the water level high, while 

void is accumulating in the RCS. 

Due to the lack of analysis, the 

consequences of a PORV failure 

were unknown. As it turned out, the 

actual consequences without proper 

mitigation were a lot worse than 

one would expect. The assumption 

that consequences get better with 

decreasing break size was incorrect. 

The actual consequences of the 

accident equally surprised the designers, 

the owner/operator of the 

plant, and the regulators. The plant’s 

response to the PORV failure was 

totally unexpected. 

Accident management 

Early in the morning of March 28, 

1979, four young operators at 

TMI-2 realized that something had 

happened, but they had no idea what 

it was. The turbine shut down, the 

reactor scrammed, and a cascade of 

alarms sounded and flashed. The 

plant was acting strangely. RCS 

pressure was decreasing while the 

pressurizer water level was increasing. 

The operators had not faced this 

situation before. It was not covered in 

their training. They did not know 

what to do. 

The event facing the operating 

crew was a stuck-open PORV and a 

very small LOCA. They did not know 

that was the case. There was no direct 

indication of PORV position in the 

control room. They could not see that 

the PORV was stuck open. 

Not knowing what was going on 

and not having familiarity with the 

event, the operators were improvising, 

trying to maintain water level 

in the RCS within prescribed limits. 

They relied on the pressurizer water 

level reading, as they were trained to 

do. Unfortunately, they took a few 

inappropriate actions, which included 

turning off the high-pressure emergency 

core cooling system, opening 

the letdown line, ignoring signs of 

overheating of the reactor core, and 

pumping radioactive water to the 

auxiliary building. All of this occurred 

before they learned – two hours and 

20 minutes into the accident – that the 

PORV was stuck open. Then they 

took corrective action and closed 

the block valve. 

The obvious question is, “Why 

were the operators in the dark, and 

why did they lack familiarity with 

this event?” Their training covered 

mitigation of postulated accidents, including 

LOCAs. There was only one 

set of accidents missing, very small 

LOCAs, including PORV failure. Since 

the designer did not analyze this 

event, it was not included in operator 

training. Not knowing the plant’s 

response to a PORV failure, the 

designers and the training staff 

instructed the operators to always 

rely on the pressurizer water level 

indication for water level measurements 

in the RCS. The operators 

followed their training on that 

morning. 

Despite the total lack of training 

for a stuck-open PORV event, could 

the operators have realized what 

was going on and taken appropriate 

action? The answer is yes. [1] The 

temperature of the PORV drain pipe 

was monitored and showed high 

readings, an alarm signaled high 

water level in the containment building 

sump, high neutron level indications 

were observed in the reactor 

core, temperature and pressure were 

rising in the containment building, 

and the reactor coolant pumps were 

vibrating. Any of these observations, 

typical of a LOCA, could have brought 

attention to a stuck-open PORV. The 

remedy should have been obvious: 

Close the block valve. 

Once the block valve was closed, 

the LOCA was terminated. The next 

step was to cool the core by natural 

circulation of the water in the RCS. 

This was not possible, however, due to 

the large amount of void that had 

accumulated in the RCS. The operating 

staff had to improvise again to 

reduce the void and the bubble in 

the RCS, and then to establish neutral 

circulation. It took a couple of days’ 

work for them to accomplish this. 

Both the plant’s designer and 

operator lacked the knowledge of how 

the plant would respond to a stuckopen 

PORV. What they did not know, 

they could not pass on to the operators. 

The operators’ training was 

misleading, and the emergency procedure 

was incorrect for the incident 

they were facing. 

Industry practice and 

oversight 

B&W’s two omissions – safety-related 

classification of the PORV and the 

PORV block valve, and the lack of 

PORV failure analysis – were not 

unique to B&W. The other U.S. PWR 

designers, Westinghouse and Combustion 

Engineering, made the same 

omissions. How could three independent 

sets of engineers make the 

same mistake? Licensing of the plants 

was a major consideration. The SAR 

was the centerpiece of the licensing 

review. Precedent provided guidance 

for the preparation of the report. 

Analyses presented in previous 

applications were included in the 

report; analysis that was not 

required was ignored. Dozens of 

utilities received SARs with the 

same omissions. The omissions 

had a direct and major effect on 

the training of reactor operators. 

The operators received training on 

a plant simulator, with postulated 






| TMI II – Late accident phase. 


was Manager of the 

Safety Analysis 

Department of 

Combustion Engineering’s 

Nuclear Division 

in its formative years. 

He later joined the 

Nuclear Regulatory 

Commission and was 

a charter member of 

the commission’s 

Senior Executive 

Service. 

accidents programmed into the 

simulator. One accident, the PORV 

failure, was missing. Nobody noticed 

it or took corrective action. After 

PORVs failed seven times at B&W 

plants, this accident was still missing 

from the operator training program 

and from the simulator. 

In the case of PWR evaluations, 

the NRC had the distinct advantage 

of reviewing SARs from three 

independent designers. Comparisons 

among the three designs frequently 

helped in the reviews. The NRC, 

however, failed to recognize its own 

effect on plant design and analysis. 

A nuclear power plant is a complex 

system. A regulatory review and 

evaluation cannot address all aspects 

of the design, and priorities have to 

be set. There was a tendency not to 

require more from an applicant than 

was required from previous ones. 

Spending time reviewing areas of the 

design that weren’t reviewed in the 

past was discouraged. Consequently, 

regulators and designers addressed 

the same areas of the design and the 

safety analyses over and over again 

and ignored other areas. 

Conclusions 

Failure to incorporate the safety 

function of the PORV and the block 

valve in the design of the plant created 

the condition for the TMI accident. 

With no positive indication in the 

control room of an open PORV and no 

positive position indication of the 

block valve, the operators were left to 

guess what was going on and what 

needed to be done. 

Not having addressed PORV failure 

in the plant safety analysis, the 

designers, as well as the training and 

operating staff, were unfamiliar with 

the plant’s response to this type of 

accident. They did not know that 

| TMI II – Late accident phase. 

the plant conditions the operators 

were facing were possible, and as 

a result, training and instructions 

were inadequate. When similar plant 

designs are being reviewed or evaluated 

one after the other, there is a 

tendency to address the same issues 

in each case. Plants are very complex, 

and not everything can be evaluated 

as part of one review. It is appropriate 

to shift emphasis in subsequent 

reviews and to address issues previously 

not covered. 

Appropriate NRC regulations relative 

to LOCAs to control the design 

and operation of the plants’ safety 

systems and develop operator training 

programs and emergency procedures 

were evolving when B&W designed its 

first plants, but they were in place at 

the time of TMI-2’s licensing. The 

problem was that some of the 

regulations were not followed. 

The two omissions – not recognizing 

the safety function of the PORV 

and the block valve, and the failure to 

analyze the stuck-open PORV event – 

were the root causes of the TMI-2 

accident. Correcting the first omission 

would have prevented the accident. 

Correcting the second omission would 

have resulted in prompt and effective 

mitigation of the accident. 

Lessons learned 

Understanding the root causes of 

the TMI accident provides valuable 

guidance for nuclear power plant 

designers, especially for designers 

of new plant types, such as small 

modular reactors. The recognition 

of safety-related components and 

design-specific accidents is more 

complex and more difficult than it 

appears to be. It is the designer’s 

responsibility to identify all safetyrelated 

systems and components 

and to analyze all accident types. 

Many systems and components 

of a plant have both an operational 

function and a safety function. In the 

design of every system, the question 

must be raised as to whether a system 

or component has a safety function. 

Then, if applicable, it must be 

designed for both the operational 

function and the safety function. 

Plant response during accidents 

can be abnormal and never seen 

during normal operation. The plant’s 

safety analysis must be complete, and 

it must describe all potential plant 

responses. 

Designers cannot depend on 

utilities’ reviews and regulatory 

evaluations to correct shortcomings. 

The design must be done right in the 

first place, and the quality assurance 

process should guarantee perfection 

of the design. 

Acknowledgment 

I am grateful to Sheldon Trubatch for 

his valuable suggestions, review of 

this article, and thoughtful comments 

and insight into the era from the legal 

perspective surrounding the accident. 

I have derived great benefit from our 

stimulating discussions. 

References 

1. John G. Kemeny, et al.: Report of the President’s Commission on 

the Accident at Three Mile Island (October 30, 1979). 

2. Mitchell Rogovin, George T. Frampton Jr.: Three Mile Island: A 

Report to the Commissioners and to the Public (January 1980). 

Author 





Das neue Strahlenschutzrecht (IV) – Schutz vor Radon 

Christian Raetzke 

Den vorläufigen Abschluss der kleinen Reihe zum neuen Strahlenschutzrecht sollen heute die neuen Regelungen zum 

Schutz vor Radon bilden. Das ist zwar ein Thema, das in der Kernenergie und Kerntechnik nicht unbedingt im 

Vordergrund steht; dafür dürfte es künftig in Teilen der Wirtschaft und der Bevölkerung um so mehr für Aufmerksamkeit 

sorgen und ist sicherlich von allgemeinem Interesse. 

Der Schutz vor Radon wird im neuen Strahlenschutzrecht 

(im Strahlenschutzgesetz – StrlSchG – vom 27.06.2017 

und in der neuen Strahlenschutzverordnung – StrlSchV – 

vom 29.11.2018), das an vielen Stellen und besonders 

auch hier die Euratom-Grundnorm zum Strahlenschutz 

(Richtlinie 2013/59/Euratom) umsetzt, einschneidend 

neu und viel umfassender als bisher geregelt. In der alten 

StrlSchV gab es bei „Arbeiten“ (so die damalige Terminologie), 

die im Zusammenhang mit natürlich vorkommender 

Radioaktivität standen, bestimmte Regelungen für 

Arbeitsplätze bis hin zu einer Anzeigepflicht (§ 95 StrlSchV 

a.F.); dazu gehörte in Anlage XI Teil A eine kurze Liste 

„Arbeitsfelder mit erhöhten Radon-222-Expositionen“. 

Die Regelungen im neuen Strahlenschutzrecht sind viel 

umfassender aufgestellt. Sie betreffen nicht nur Arbeitsplätze, 

also die berufliche Strahlenexposition, sondern 

auch Aufenthaltsräume, also die Exposition der Bevölkerung. 

Einige Regelungen gelten speziell für eine der beiden 

Expositionskategorien, andere sind allgemeiner, übergreifender 

Art. Am besten findet man sich wohl zurecht, 

wenn man die Regelungen entsprechend in drei Gruppen 

betrachtet: allgemeine, auf Arbeitsplätze bezogene und 

auf Aufenthaltsräume bezogene Bestimmungen. So sind 

sie auch im Gesetz und in der Verordnung eingeteilt. 

Bei den allgemeinen Regelungen ist besonders die 

Pflicht der zuständigen Länderbehörden zu erwähnen, bis 

Ende 2020 sog. Radonvorsorgegebiete auszuweisen. Dieser 

Begriff (der so im Gesetz nicht auftaucht) be zeichnet 

Gebiete gem. § 121 StrlSchG, in denen die Radon-222- 

Aktivitätskonzentration in der Raumluft „in einer beträchtlichen 

Zahl von Gebäuden“ die Referenzwerte für Aufenthaltsräume 

oder Arbeitsplätze überschreitet. Die beiden 

Referenzwerte für Aufenthaltsräume und Arbeitsplätze sind 

im Gesetz formal getrennt festgelegt (§§ 124 bzw. 126 

StrlSchG), lauten aber übereinstimmend 300 Becquerel 

(Bq) je Kubikmeter (m 3 ). Die Bestimmung der Vorsorgegebiete 

erfolgt „auf Grundlage einer wissenschaftlich 

basierten Methode“, die „Vorhersagen“ ermöglicht (so § 153 

StrlSchV); geeig nete Daten hierfür können auf tatsächlichen 

Messwerten, aber auch auf geologischen Daten, also 

letztlich auf Berechnungen beruhen. Eine „beträcht liche 

Zahl von Gebäuden“ soll erreicht sein, wenn auf mindestens 

75 % des auszuweisenden Gebiets der Referenzwert in 

min destens zehn Prozent der Gebäude überschritten wird. 

Sind erst einmal diese Radonvorsorgegebiete ausgewiesen, 

so ergibt sich aus § 123 StrlSchG die Pflicht, bei 

Neubauten in diesen Gebieten „geeignete Maßnahmen zu 

treffen, um den Zutritt von Radon aus dem Baugrund zu 

verhindern oder erheblich zu erschweren“. Das gilt einheitlich 

für Gebäude mit Arbeitsplätzen und/oder mit Aufenthaltsräumen, 

also auch für Wohngebäude oder Bauten 

für Schulen und Kindergärten. § 154 StrlSchV zählt fünf 

Arten von Maßnahmen auf und lässt „mindestens eine“ 

davon genügen; damit soll offenkundig die Anforderung 

handhabbar gemacht und eine Übersteigerung vermieden 

werden. Zu den Maßnahmen zählen etwa die Verringerung 

der Radon-222-Aktivitätskonzentration unter dem 

Gebäude, die Begrenzung der Rissbildung in Wänden und 

Böden mit Erdkontakt und die Absaugung von Radon an 

Randfugen oder unter Abdichtungen. 

Der Bund ist nach § 122 StrlSchG verpflichtet, einen 

Radonmaßnahmenplan zu erstellen; das ist mittlerweile 

geschehen (siehe www.bmu.de/publikation/radonmassnahmenplan). 

Das Dokument nennt, neben den gesetzlich 

geforderten Maßnahmen, auch und vor allem Informationskampagnen, 

Förderung der Allgemeinbildung 

zu Radon und der beruflichen Qualifikation und Weiterbildung, 

Entwicklung einheitlicher Messstrategien, Vergabe 

weiterer Studien und Ähnliches. 

Bei Arbeitsplätzen besteht dann eine Handlungspflicht, 

wenn sie entweder einem der bereits in der alten StrlSchV 

genannten „Arbeitsfelder“ angehören (vor allem Berg werke 

und Anlagen der Wassergewinnung und -ver teilung), und 

zwar unabhängig vom geografischen Gebiet, oder wenn sie 

sich im Erd- oder Kellergeschoss eines Gebäudes in einem 

Radonvorsorgegebiet befinden. §§ 127 bis 131 StrlSchG 

ordnen dann ein abgestuftes Verfahren an. Zeigt eine erste 

Messung, die innerhalb von 18 Monaten nach Einrichten 

des Arbeitsplatzes oder nach Ausweisung des Gebietes vorzunehmen 

ist, eine Überschreitung des Referenzwerts, 

muss der für den Arbeitsplatz Verantwortliche Maßnahmen 

zur Reduzierung der Radonkonzentration ergreifen und 

erneut messen. Liegt der Wert dann immer noch über dem 

Referenzwert, muss er den Arbeitsplatz bei der zuständigen 

Behörde anmelden und die Exposition abschätzen; vom 

Ergebnis dieser Abschätzung hängen dann bestimmte Maßnahmen 

des beruflichen Strahlenschutzes ab. Der Arbeitgeber 

wird damit nicht zum Strahlenschutzverantwortlichen, 

da es sich – in der neuen Terminologie des StrlSchG 

– nicht um eine geplante Expositionssituation, sondern um 

eine bestehende Expositionssituation handelt; er hat aber 

teils vergleich bare Pflichten. 

Für Aufenthaltsräume (also Wohnungen, Schulen, 

Kinderg ärten etc.) enthalten §§ 124 und 125 StrlSchG 

Regelungen, die keinen verpflichtenden Charakter haben, 

sondern auf Unterrichtung der Bevölkerung und auf das 

„Anregen“ von Messungen und ggf. von Schutzmaßnahmen 

hinauslaufen. Wie oben erwähnt, gibt es aber für 

Neubauten in Radonvorsorgegebieten, die Aufenthaltsräume 

enthalten, eine Pflicht, Maßnahmen zum Schutz 

vor Radon zu treffen; insofern sind nur Bestandsbauten 

vorerst von verbindlichen Maßnahmen verschont. 

Wie geht es weiter? Das Thema Radon steht schon seit 

einigen Jahren im Fokus der Strahlenschützer, im allgemeinen 

Bewusstsein dürfte es aber erst dann wirklich ankommen, 

wenn ab 2021 die Radonvorsorgegebiete ausgewiesen 

sind, Informationskampagnen richtig anlaufen und echte 

Pflichten für viele Arbeitgeber und Bauherren entstehen. Es 

steht zu hoffen, dass die neuen Regelungen dann die 

gewünschte Steuerungswirkung entfalten und die findige 

Bevölkerung nicht einfach nur gemäß einem Spruch handelt, 

den der Verfasser von einem geschätzten Kollegen aus 

dem (radonmäßig stark betroffenen) Vogtland gehört hat: 

„Die Zahnarztpraxis unters Dach, die Oma in den Keller“. 

Author 

Rechtsanwalt Dr. Christian Raetzke, 

Beethovenstr. 19, 04107 Leipzig 

525 

SPOTLIGHT ON NUCLEAR LAW 

Spotlight on Nuclear Law 

The New Radiation Protection Law – Protection Against Radon ı Christian Raetzke


526 

RESEARCH AND INNOVATION 

Evaluation of a Double-Ended Guillotine 

LBLOCA Transient in a Generic 

Three-Loops PWR-900 with TRACE Code 

Coupled with DAKOTA Uncertainty 

Analysis 

Andrea Bersano and Fulvio Mascari 

In the present study, the model of a generic three-loops PWR-900 western type reactor has been developed and a 

double-ended guillotine break on the cold leg has been simulated by TRACE code. Through the SNAP graphical 

interface, a DAKOTA uncertainty analysis, based on the probabilistic method to propagate input uncertainty, has been 

performed by selecting uncertain parameters related to the safety injection system and to the initial plant status. In 

particular, six uncertain input parameters have been considered: the accumulators’ initial pressure and temperature, 

the safety injection system temperature and flow rate, the reactor initial power and the containment initial pressure. 

The main figure of merit selected for the application of regression correlation is the hot rod cladding temperature. Both 

Pearson and Spearman’s correlation coefficients have been computed for the cladding temperature of the hot rod to 

characterize its correlation with the input uncertain parameters in the different phases of the transient. In addition, the 

dispersion of the calculated data have been discussed for selected relevant thermal-hydraulic parameters, such as the 

primary pressure, the core mass flow rate and the water collapsed level in the vessel. 

1 Introduction 

Nuclear energy is part of the energy 

mix of several countries considering 

its important role in the reduction of 

air pollution and CO 2 emissions 

caused by energy pro duction [1]. Due 

to the complexity of nuclear plants, 

caused by their complex geometry 

where multicomponent and twophase 

thermal hydraulic phenomena 

take place, their thermal hydraulic behavior 

is characterized with computational 

tools to assess their operating 

conditions and to evaluate their safety 

(both for Design Basis Accidents and 

Beyond Design Basis Accident) [2, 3]. 

The computational tools used in 

the nuclear sector, also called codes, 

undergo a process of Verification and 

Validation (V&V) [4]. It is part of this 

rigorous process the use of an experimental 

“assessment database” [3]; 

results from Separate Effect Test 

Facilities (SETF) and Integral Test 

Facilities (ITF) are used to evaluate 

the qualitative and quantitative code 

accuracy in the prediction of the 

phenomena of interest. The codes 

that use more realistic information 

concerning phenomena and plant 

behavior are often referred to as Best 

Estimate (BE) codes [5]. 

Though the high level of maturity 

reached by BE thermal hydraulic 

system codes in the last decades, in 

their application there are still some 

sources of uncertainty ( uncertainty is 

used as a measure of the error made 

with the code in predicting the plant 

behavior) affecting the calculation 

results [3]. In general the sources of 

uncertainty can be grouped as a) code 

uncertainty (e.g. approximations in 

the onservation equation and in the 

closure models and correlations) 

b) representation uncertainty (nodalization 

effect), c) scaling issue (codes 

validated against scaled-down facilities), 

d) plant uncertainty (e.g. initial 

and boundary conditions), e) user 

effect. For this reason, providing the 

result of a BE calculation alone may be 

not sufficient and the evaluation of 

the uncertainty on the results is 

required. Several methodologies have 

been developed in the past to perform 

Uncertainty Analysis. In particular 

these uncertainty methodologies can 

be grouped in a) methods to propagate 

input uncertainty, divided in 

probabilistic (e.g. CSAU, GRS, IPSN, 

etc.) and deterministic methods (e.g. 

AEAW, EDF-Framatome, etc.); and b) 

method to extrapolate output uncertainty 

(e.g. UMAE) [6]. 

In this framework, the target of the 

present paper is to use the probabi listic 

method to propagate the input 

uncertainty in the calculation of a cold 

leg double- ended guillotine break 

transient of a generic three-loops PWR- 

900 western type reactor with the 

avail ability of active and passive Emergency 

Core Cooling Systems (ECCS). 

The Uncertainty Quanti fication (UQ) 

application of this methodology is 

based on a set of statistical techniques 

for the evaluation of the number of 

needed code runs and for the correlation 

of the output results with the 

uncertain inputs. The method has been 

applied using DAKOTA toolkit [7], 

developed by Sandia National Laboratories, 

and has been used in the SNAP 

(Symbolic Nuclear Analysis Package) 

environment/architecture [8] with the 

related DAKOTA uncertainty plug-in 

[9]. The calculations have been performed 

with TRACE (TRAC/RELAP 

Advanced Computational Engine) v5 

patch 4 BE thermalhydraulic system 

code developed by USNRC [10]. The 

hot rod cladding temperature has been 

selected as a figure of merit and a 

limited number of input un certainty 

parameters, mostly related to plant 

initial conditions, have been chosen for 

the UQ application. 

2 Description of probabilistic 

method to propagate 

input uncertainty, 

Dakota Toolkit, and UQ 

hyphothesis 

2.1 Probabilistic method 

to propagate input 

uncertainty 

The probabilistic method to propagate 

the input uncertainty [11] is, in brief, 

based on a random sampling of the 

input uncertain parameters selected 

by the user; a set of N code runs having 

in the input a combination of the uncertain 

parameters is created and 

solved with the selected code. Then, 

by using regression analysis, the effect 

of the input parameters uncertainty 

on the results is computed, in terms of 

selected Figure of Merits (FOMs). The 


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


main advantages of this method are 

that the number of code runs does not 

depend on the number of input 

uncertain parameters and it is not 

necessary a prior development of 

a Phenomena Identification and 

Ranking Table (PIRT), which is a long 

process. The number of code runs N 

depends on the probability content 

and on the confidence level set by the 

user and on the number of FOMs 

selected for the analysis and it is computed 

using the Wilks formula [12, 

13]. Each uncertain input parameter 

should be defined by its range of 

variation and its Probability Density 

Function (PDF); this means that even 

if the exact value of a parameter is 

uncertain, some values are more likely 

to be close to the real one than others. 

This is one of the main task to be 

tackled because the evaluation of the 

correct PDF for each parameter is not 

a trivial task and a careful study is 

required. The selection of the most 

suitable PDF for each parameter depends 

mainly on the physical quantity 

and on the availability of reducedscale 

or full-scale measured data. If 

data are available, it is possible to 

build a histogram and to derive the 

PDF. A comprehensive overview of 

PDF type and derivation techniques is 

provided in [14]. 

2.2 DAKOTA toolkit in the 

SNAP environment/ 

architecture 

DAKOTA [7] is an open source software 

written in C++ and developed by 

Sandia National Laboratories to perform 

parametric and uncertainty 

analysis in a fast and automatic way. 

The aim of this toolkit is to bridge 

simulation codes and analysis methods 

for parametric evaluation, uncertainty 

quantification and system optimization 

[15]. The DAKOTA toolkit is also 

provided as a plug-in [9] for the 

Symbolic Nuclear Analysis Package 

(SNAP), which is a graphical user 

interface designed to support the use 

of USNRC nuclear codes (e.g. TRACE, 

RELAP, MELCOR, etc.). Using SNAP, it 

is possible to build the input deck in a 

graphical environment and to have a 

direct visualization of the code calculated 

data by using its animation 

capability. Through SNAP it is possible 

to set up the DAKOTA uncertainty 

analysis [16, 17] and to perform automatically 

all the steps qualitatively 

described in the previous section. 

Figure 1 shows a schematic representation 

of DAKOTA uncertainty analysis 

workflow in a SNAP environment/ 

architecture. 

In particular, DAKOTA plugin allows 

to: 

1) Enter the uncertain input parameters 

with their range and PDF; 

2) Select the sampling method 

( Monte Carlo or Latin Hypercube); 

3) Enter the desired FOMs for the 

analysis; 

4) Set the final report that contains 

the results of the uncertainty 

analysis application; the report is 

auto matically generated at the end 

of the uncertainty quantification 

analysis. 

DAKOTA is used at the beginning of 

the analysis to sample the uncertain 

input parameter values and to 

generate the set of code inputs. Then, 

after the solution of the set of code 

inputs and the extraction of the 

desired data, DAKOTA performs the 

uncertainty analysis and apply 

regression techniques to evaluate the 

correlation between input and output 

parameters selected as a FOM. 

The required number of code runs 

can be found solving the Wilks 

formula with respect to N for a probability 

α and a confidence level β [11]: 

(1) 

With α=0.95 and β=0.95, for one 

FOM the required number of code 

runs is 59. 

2.3 UQ application hypothesis 

The target of this analyses, is not to be 

a detail uncertainty study in term of 

input uncertainty parameters as 

presented in [18], but a) to develop a 

full UQ application with TRACE and 

DAKOTA toolkit in a SNAP environment/architecture 

and b) to have 

some insights about the degree of 

corre lation between the input parameters 

selected and the FOM chosen 

for this analysis. Six uncertain parameters 

have been selected for this 

uncertainty application based also on 

BEMUSE program results [18] and 

through SNAP have been implemented 

in the DAKOTA and TRACE 

input: the Safety Injection System 

(SIS) temperature, the SIS characteristic, 

the accumulator initial temperature 

and pressure, the initial core 

power and the initial con tainment 

pressure. The SIS characteristic is a 

value that multiply the default injected 

flow rate curve as function of the 

primary pressure. Table 1 summarizes 

the uncertain input parameters, 

their mean value used for the reference 

calculation, the range of variation 

and the adopted PDF. 

One FOM was selected for the 

analysis, the cladding temperature of 

the hot rod; therefore, with a probability 

of 95% and a confidence level of 

95%, a total of 59 calculations were 

required based on Wilks formula as 

previously described. Latin Hypercube 

sampling [19,20] has been used 

for this analysis. It is a stratified 

sampling method that, with respect to 

a pure Monte Carlo sampling, allows 

to achieve the target statistical 

| Fig. 1. 

DAKOTA uncertainty analysis workflow for TRACE code in a SNAP environment/architecture. 

RESEARCH AND INNOVATION 527 




Parameter Average value Range of variation PDF type 


Safety injection system (SIS) temperature [K] 285 [-10,+10] K Normal 

Safety injection system (SIS) characteristic [-] 1 [0.95,1.05] Normal 

Accumulator initial temperature [K] 325 [-10,+10] K Normal 

Accumulator initial pressure [bar] 40.8 [-2,+2] bar Normal 

Initial core power [MW] 2785 [0.98,1.02] Normal 

Initial containment pressure [bar] 1.013 [0.85,1.15] Uniform 

| Tab. 1. 

Input uncertain parameters selected for the analysis. 

| Fig. 2. 

TRACE nodalization of the primary system and of the containment of the generic three-loops PWR developed with SNAP. 

accuracy with a minor number of 

samples. All 59 runs were correctly 

executed and they reached the end of 

calculation without any failure. To 

analyze the transient progression, the 

primary system pressure, the core 

flow rate, the cladding hot rod temperature 

and the vessel collapsed level 

have been plotted for the complete set 

of 59 runs, while the application of 

regression analysis has been carried 

out only for the cladding temperature. 

3 Generic PWR-900 TRACE 

model and steady state 

operation 

3.1 TRACE nodalization 

description 

TRACE, developed by USNRC, is a 

component-oriented code designed 

to perform best-estimate thermalhydraulic 

analysis for LWR. It is a 

finite volume, two fluid, code with 3D 

capability and it is based on two fluid, 

two-phase field equations. This set of 

equations consists of the conservation 

laws of mass, momentum and energy 

for the liquid and gas phases [10, 21]. 

The code version adopted in this 

analysis is TRACE code v5 patch 4 and 

the input deck has been developed 

with SNAP. 

The nodalization of a generic threeloops 

PWR-900, shown in Figure 2, 

has been developed to perform the 

large break LOCA analysis. In order to 

minimize the computational time (in 

view of input uncertainty propagation 

with probabilistic method) and maintain 

an accurate prediction of target 

phenomena, the nodalization strategy 

used follows the general nodalization 

approach of the TRACE W4loops 

samples input-deck distributed with 

SNAP. Starting from that nodalization 

approach sample and considering the 

level of detail target of this analysis 

(e.g. modeling the three loops separately, 

modeling the interaction containment/primary 

coolant system, 

etc…), more details have been considered 

in the input-deck development. 

In the nodalization of the generic 

three-loop PWR-900 no lumped loops 

have been considered and the three 

loops have been modeled separately, 

one simulates the broken one (loop A) 

and two simulate the intact loops 

(loop B and C). The model is composed 

of 69 Hydraulic Components 

(HC) and 45 Heat Structures (HS). 

Power is provided to the three heat 

structures simulating the core through 

one power component, the total 

thermal power is around 2800 MW; in 

one of the HS that compose the core is 

inserted a supplemental rod that simulates 

the hot rod in the reactor with a 

total peaking factor of 2.278 [22]. The 

pressurizer is connected to the hot leg 

of an intact loop (loop B). The break 

has been modeled with a set of three 

valves; at the break opening one valve 

interrupts the connection between the 

two sections of the guillotine break on 

the cold leg of loop A; simultaneously, 

the other two valves connect the two 

closed sections of the leg to the components 

simulating the containment. 

Particular attention has been paid 

to the containment building modeling 

in order to simulate the interaction 

containment/primary system during 

the LOCA. In particular, the containment 

has been modeled with two connected 

hydraulic regions thermally 

coupled to the heat structure simulating 

the containment solid structure 

and the thermal interaction with 

the environment. Figure 2 shows 

the TRACE nodalization of the 

generic PWR primary system and the 




| Fig. 3. 

TRACE nodalization of 

the RPV of the generic 

three-loops PWR developed 

with SNAP. 

| Fig. 4. 

SNAP animation showing the fluid condition of the primary system during the steady state (at 900 s). 


containment (the containment drawing 

is scaled down with a factor 0.2 in 

comparison with the other reactor 

components for a better visualization 

of the complete plant layout). The 

ECCS are connected on the cold leg of 

each loop; for each loop, the ECCS 

consist of an accumulator and the 

high pressure and low pressure safety 

injection systems (HPIS, LPIS). The 

HPIS and LPIS are modeled by a single 

“fill” component with a table that 

controls the injected flow rate as 

function of the pressure in the primary 

system. The Reactor Pressure Vessel 

(RPV) has been modeled by using the 

3D Vessel component available in 

TRACE. The 3D vessel nodalization 

has been divided into: 2 radial sectors, 

the inner one for the core and the 

outer one for the downcomer, 3 angular 

sectors, one for each primary loop 

and 7 axial sectors, 2 for the vessel 

lower plenum, 3 for the core and 2 

for the vessel upper plenum. The 3D 

vessel nodalization representation 

made with SNAP is shown in Figure 3, 

with the core region highlighted. 

3.2 Steady state 

characterization 

Before the beginning of the LOCA, 

1000 s of steady state operation have 

been simulated to test the nodalization 

in steady state condition. The 

steady state parameters are reported 

in Table 2 and they have been checked 

for consistency with public available 

information [23, 24]. 

Figure 4 shows the fluid con ditions 

in the primary system during the 

steady state; the color legend refers to 

the fluid status, ranging from blue 

(subcooled liquid) to red (superheated 

gas/steam). Figure 5 shows 

the axial pressure profile in the 

primary system in steady state conditions, 

to support the steady state 

qualification of the nodalization. The 

normalized pressure profile is consistent 

with other similar data available 

in the public scientific literature 

[25]. 

4 Cold leg lbloca transient 

and uncertainty analysis 

4.1 LBLOCA transient analysis 

The analyzed transient is a doubleended 

guillotine break (200 %) in the 

cold leg of loop A. After 1000 s of 

steady state simulation the break is 

opened (start of the transient: t = 0 s) 

and 500 s of transient are simulated. 

The reactor SCRAM is supposed after 

0.5 s from the break. From Figure 6 to 

Figure 9 the results for the 59 runs are 

shown for the primary system pressure, 

the core flow rate, the cladding 

hot rod temperature and the vessel 

collapsed level. Figure 6 shows the 

behavior of the primary system pressure; 

after the LOCA initiation, water 

flows from the primary system to the 

containment and the primary system 

pressure drops significantly in few 

tens of milliseconds in agreement 

with the publically available technical 

scientific literature [26]; after this 

first drop of pressure and the phase 

of subcooled depressurization, the 

saturation con dition is reached and 

saturated depressurization starts at a 

reduced rate. The maximum flow rate 

through the break is limited by the 

critical velocity at the break. After the 

transient initiation, the flow rate in 

the core (Figure 7) drops from the 

nominal value and it is reversed since 

the flow is directed to the break 

location; therefore, water flows downwards 

in the core region and then 

upward in the downcomer to reach 

the break in the cold leg. 

The high voiding in the core and 

the subsequent SCRAM stop the chain 

reaction and, due to a sensible 

reduction of heat removal in the 

core, the heat stored in the fuel is 

Thermal power [MW] 2785 

Primary system pressure [bar] 155 

Total core flow rate [kg/s] 13,947 

Core inlet temperature [K] 558.4 

Core outlet temperature [K] 594.1 

Secondary system pressure [bar] 58 

Steam generator feedwater temperature [K] 440 

Steam generator feedwater flow rate [kg/s] 512 

| Tab. 2. 

Steady state parameters of the reference calculation. 

| Tab. 5. 

Normalized axial pressure profile along the primary system in steady state 

condition (at 900 s). 





redistributed, leading to the first 

cladding temperature peak, as shown 

in Figure 8. The cladding temperature 

passes from the steady state value 

(around 617 K) to 955 K. The result 

of the core voiding is a sudden drop 

of the water collapsed level in the 

reactor vessel, as shown in Figure 9, 

under the Bottom of Active Fuel 

(BAF). When the pressure in the 

primary system is lower than the 

accumulator initial pressure, the 

accumulator check valves open and 

water is discharged in the primary 

system; this happens around 11 s after 

the transient initiation. The water 

initially injected in the cold legs by the 

accumulators bypasses the vessel 

lower plenum through the upper 

down comer region and it is directed to 

the break without penetrating the 

core. After the first fast depressurization 

phase, the primary pressure 

continues to reduce at a lower rate, 

equalizing the containment pressure 

(around 0.45 MPa after around 40 s) 

and ending the blowdown phase. 

Water is injected in the primary 

system by the accumulators and the 

low pressure safety injection system 

start to inject water after around 30 s 

from the break initiation. The refill 

phase starts around 40 s, when the 

emergency core coolant water reaches 

the vessel lower plenum and the 

collapsed level in the vessel starts to 

rise. During this phase the core is 

mainly uncovered and heat is not 

removed from the fuel rods, with the 

exception of a small amount of heat 

removed by thermal radiation and 

natural convection of the steam 

present in the core. For this reason 

during the refill period the cladding 

temperature increases (Figure 8) due 

to the quasi adiabatic heating of 

fission product decay. When, around 

50 s from the LOCA initiation, the 

water level reaches the core bottom 

(Figure 9) the refill period ends 

and the reflood phase starts. Water 

collapsed level rises quickly up to 

around 65 s (time of end of accumulators 

injection), and it continues at a 

lower rate due to the LPIS. In this 

phase the net core flow rate is positive, 

even if very small and with many 

oscillations. The water entering the 

core is heated up, starts to boil and 

entrains water droplets that help the 

cooling of the hottest parts of the core. 

With the rising of the water level, the 

cooling is increased and the cladding 

temperature starts to decrease. The 

complete rewetting of the cladding 

surface caused by the rising water 

level produces a strong temperature 

drop (core quenching). This happens 

around 125 s after the beginning of 

the LOCA. 

Analyzing the dispersion of the 

results, the primary pressure (Figure 

6) presents an almost negligible 

dispersion during the blowdown 

phase and the predictions of the 59 

runs are very similar; after the blowdown 

the dispersion band width is 

always lower than 0.1 MPa with a final 

average value of 0.475 MPa. As 

regards the core mass flow rate 

(Figure 7), cladding temperature 

(Figure 8) and vessel water collapsed 

level (Figure 9), the results dispersion 

is very limited during the blowdown 

phase, while it is more noticeable in 

the refill phase, especially for the 

vessel water collapsed level, and it is 

higher during the reflood phase. 

In particular, the refill initial time 

shows a dispersion band width of 12 s 

(33 – 45 s); during the reflood the 

collapsed level dispersion band width 

is around 1 m and the Top of Active 

Fuel (TAF) is reached in a time band 

of 33 s (126 – 159 s). The cladding 

temperature has, at the first peak, a 

low dispersion band width of 14 K 

and the peak has almost the same 

timing for all runs; instead, the dispersion 

band width is higher for the 

second peak (50 K) and with a time 

dispersion band width of 10 s. The hot 

rod cladding quenching time is also 

affected by a dispersion band width 

of 20 s (115 – 135 s). 

| Tab. 6. 

Primary system pressure predicted by TRACE code. 

| Tab. 7. 

Core mass flow rate predicted by TRACE code. 

| Tab. 8. 

Cladding temperature of the hot rod predicted by TRACE code. 

| Tab. 9. 

Reactor vessel water collapsed level predicted by TRACE code. 




| Tab. 10. 

Pearson’s simple regression correlation coefficients. 

4.2 Uncertainty analysis 

response correlations 

4.2.1 Description of regression 

correlations in DAKOTA 

As a result of the uncertainty analysis, 

DAKOTA [7, 8, 20] computes four 

response correlation coefficients: 

simple, partial, simple rank and 

partial rank. The simple coefficient is 

related to the actual input and output 

data. The simple coefficient r between 

an input variable x and an output 

variable y, in n samples, is computed 

using the Pearson’s correlation. It is a 

measure of the degree of linear correlation 

between the two variables and 

its value is comprised between -1 and 

1. If r0 the 

correlation is positive (an increment 

of x leads to an increment of y). The 

simple correlation coefficient between 

x and y is obtained by dividing the 

covariance of the two variables by the 

product of their standard deviations 

[27]: 

(2) 

The partial correlation coefficient is 

computed similarly to simple one but 

taking into accounts the effects of 

the other variables. This is useful, for 

example, if there is a strong correlation 

between two inputs; in this way 

the correlation of the second input on 

the output may be adjusted after 

having considered the correlation between 

the first input and the output 

[7]. Rank correlation coefficients use 

the ranked data instead of the actual 

ones. Ranks are obtained by ordering 

the data in ascending order, and are 

more convenient to be used when 

inputs and outputs are characterized 

by sensible difference in magnitude; it 

is possible to understand if the input 

sample with the lower rank is associated 

to the output with the lower 

rank and so on [7, 20]. To compute the 

rank correlation, DAKOTA uses the 

Spearman’s rank correlation that is 

similar to Pearson’s one but with the 

ranked data instead of the actual 

values. If two variables are monotonically 

related, without repetitions, 

the Spearman coefficient is -1 or +1 

(depending if the function is monotonically 

decreasing or increasing), 

since the ranked values are used. 

Moreover, Spearman’s correlation is 

less sensitive to possible outlier values 

of the variables than Pearson’s one. 

4.2.2 Results of response correlation 

coefficients for the 

cold leg LBLOCA transient 

The time dependent computation of 

response correlation coefficients has 

been performed extracting the FOM 

value at different selected instant of 

the transient evolution. Figure 10 

shows the Pearson’s simple correlation 

coefficient for the six input 

uncertain parameters from the LOCA 

beginning to the complete core 

quenching (after 130 s). Figure 11 

shows the Spearman’s rank correlation 

coefficient for the same parameters. 

On both graphs the values 0.2 

and 0.5 (and -0.2 and -0.5) have been 

highlighted as measure of the correlation 

between the input parameter 

and the FOM. As indicated in the [20], 

for the Spearman coefficient, if the 

coefficient is higher than 0.5 (or lower 

than -0.5) the correlation is significant, 

if it is between 0.2 and 0.5 (or 

-0.2 and -0.5) the correlation is 

moderate, otherwise it is low [20]. In 

| Tab. 11. 

Spearman’s simple rank regression correlation coefficients. 

the following analysis the same 

threshold values have been adopted 

also for the Pearson coefficient. In this 

application, the two response correlations 

show similar trends; the main 

advantage of this time dependent 

analysis is the possibility to have a 

measure and characterize the correlation 

of the different input parameters 

on the uncertainty of the 

selected FOM in all the phases of the 

transient. 

In the blowdown phase (0-40 s) the 

effect of the initial core power on the 

hot rod cladding temperature is 

positive due to the heat stored in the 

solid structures. Both Pearson’s and 

Spearman’s coefficients are higher 

than 0.8 for this parameter so the 

correlation is significant and almost 

monotonically linear. In the remaining 

part of the transient both coefficients 

are close to 0.2 so the corre lation is 

much weaker than in the initial phase. 

After around 20 s from the LOCA 

initiation the accumulators initial 

pressure has a negative Pearson’s and 

Spearman’s coefficients around -0.3; 

therefore the lower is the accumulator 

pressure the higher is the cladding 

temperature since the accumulator 

starts to inject water later. In the refill 

phase (40-50 s), also the SIS characteristic 

has both response correlation 

coefficients around -0.3; in fact, with a 

lower injection flow rate the hot rod 

cladding temperature is higher. In the 

reflood phase (after 50 s) also the 

initial containment pressure has negative 

Pearson’s and Spearman’s coefficients 

around -0.35. This is due to the 

fact that with a lower pressure in the 

containment a greater amount of 

coolant water is expelled through 

the break and the cladding temperature 

is higher. The uncertainty in the 

accumulators and SIS temperature 

have a low correlation with the FOM. 






5 Conclusions 

Uncertainty analysis is used in the 

nuclear sector to evaluate uncertainties 

still present in the results of code’s 

calculations. Among the several 

methodologies developed in the past 

to perform Uncertainty Analysis, the 

probabi listic method to propagate the 

input uncertainties has been selected 

for this analysis considering its 

suitability to be coupled with simulation 

codes; moreover, several 

toolkits, some integrated in computational 

platforms, have been developed 

for this purpose. In the present 

activity, a double-ended cold leg 

LBLOCA transient has been simulated 

with the BE thermal hydraulic system 

code TRACE for a generic three-loops 

PWR-900 reactor. Using DAKOTA 

toolkit, in a SNAP environment/architecture, 

an uncertainty analysis has 

been carried out by selecting six 

uncertain input parameters and the 

hot rod cladding temperature as the 

main figure of merit. In addition, the 

primary pressure, the core flow rate 

and the pressure vessel collapsed level 

have been analyzed to evaluate the 

transient progression and the results 

dispersion. The aim of this analysis is 

not to be a detail and exhaustive 

uncertainty study in term of input 

uncertainty parameters but to develop 

a complete uncertainty quantification 

application with DAKOTA in a SNAP 

environment/architecture and to have 

some insights characterizing the 

correlation between the input uncertainty 

parameters and the selected 

FOM. Pearson’s and Spearman’s 

response correlation coefficients have 

been computed between the LOCA 

initiation and the complete core 

quenching. In the blowdown phase, 

the hot rod cladding temperature has 

a significant correlation with the 

initial core power; the accumulators’ 

initial pressure has a moderate correlation 

with the FOM only in the period 

of water injection from the accumulators. 

The time dependent response 

analysis, adopted in this application, 

is very useful since it could be used to 

characterize the effect of the uncertain 

input parameters on the output 

global uncertainty in the different 

phases of a transient. In a future 

follow-up, additional uncertain input 

parameters and FOMs can be introduced 

in the analysis in order to have 

a more complete evaluation of the 

results uncertainty. 

Acknowledgement 

The authors are grateful to Ms Cristina 

Bertani for the review of the manuscript. 

Nomenclature 

N 

n 

p 

r 

x 

y 

α 

β 

Subscripts 

m 

Number of code runs 

Number of samples of a certain variable 

Number of Figure of Merit 

Pearson’s correlation coefficient 

Input variable in the computation of response correlations 

Output variable in the computation of response correlations 

Probability 

Confidence level 

mean value of the related parameter 

References 

1. F. Mascari, G. Vella, B.G. Woods, K. Welter, F. D’Auria, “Analysis 

of Primary/Containment Coupling Phenomena Characterizing 

the MASLWR Design During a SBLOCA Scenario”, Nuclear 

Power Plant, Intech (2012) 

2. F. Mascari, H. Nakamura, F. De Rosa, F. D’ Auria, “Scaling 

Rationale Design and Extrapolation Problem for ITF and SETF”, 

Book of Abstract of International Workshop on Nuclear Safety 

and Severe Accident (NUSSA-2014), Kashiwa, Chiba, Japan, 

September 3-5 (2014) 

3. F. Mascari, H. Nakamura, K. Umminger, F. De Rosa, F. D’Auria, 

“Scaling Issues For The Experimental Characterization Of 

Reactor Coolant System In Integral Test Facilities And Role Of 

System Code As Extrapolation Tool”, Proceedings Of 

International Topical Meeting on Nuclear Reactor Thermal 

Hydraulics 2015, NURETH 2015, Volume 6, pp. 4921-4934 

(2015) 

4. W.L. Oberkampf, T.G. Trucano, “Verification and validation 

benchmarks”, Nuclear Engineering and Design 238, 

pp. 716-743 (2008) 

5. NEA/CSNI “A state-of-the-art report on scaling in system 

thermal hydraulics applications to nuclear reactor safety and 

design”, NEA/CSNI/R(2016)14 (2017) 

6. IAEA International Atomic Energy Agency “Best Estimate 

Safety Analysis for Nuclear Power Plants: Uncertainty 

Evaluation”, Safety Reports Series (2008) 

7. B.M. Adams, M.S. Ebeida, M.S. Eldred, G. Geraci, J.D. Jakeman, 

K.A. Maupin, J.A. Monschke, J.A. Stephens, L.P. Swiler, 

D.M. Vigil, T.M. Wildey, W.J. Bohnhoff, K.R. Dalbey, J.P. Eddy, 

J.R. Frye, R.W. Hooper, K.T. Hu, P.D. Hough, M. Khalil, 

E.M. Ridgway, A. Rushdi, Dakota, A Multilevel Parallel 

Object-Oriented Framework for Design Optimization, 

Parameter Estimation, Uncertainty Quanti fication, and 

Sensitivity Analysis: Version 6.7 User’s Manual, 

SAND2014-4633 (2017) 

8. Applied Programming Technology, Inc. Symbolic Nuclear 

Analysis Package (SNAP) User’s Manual (2012) 

9. Applied Programming Technology, Inc. Uncertainty analyses 

User manual, Symbolic Nuclear Analyses Package (SNAP), 

(2012) 

10. U.S. Nuclear Regulatory Commission TRACE v5.840 Theory 

Manual (2013) 

11. H. Glaeser, “GRS Method for Uncertainty and Sensitivity 

Evaluation of Code Results and Applications”, Science and 

Technology of Nuclear Installations (2008) 

12. S.S. Wilks, “Determination of sample sizes for setting tolerance 

limits”, The Annals of Mathematical Statistics 12(1), pp. 91-96 

(1941) 

13. S.S. Wilks, “Statistical prediction with special reference to the 

problem of tolerance limits”, The Annals of Mathematical 

Statistics 13(4), pp. 400-409 (1942) 

14. M.E. Stephens, B.W. Goodwin, T.H. Andres, “Deriving 

parameter probability density functions”, Reliability 

Engineering & System Safety 42, pp. 271-291 (1993) 

15. https://dakota.sandia.gov/ 

16. J.R. Wang, C.W. Tsai, H.T. Lin, C. Shih, “Performing Uncertainty 

Analysis of IIST facility SBLOCA by TRACE and DAKOTA”, 

NUREG/IA-0428 (2013) 

17. J.R. Wang, J.H. Yang, H.T. Lin, C. Shih, “Uncertainty Analysis for 

Maanshan LBLOCA by TRACE and DAKOTA”, NUREG/IA-0448 

(2014) 

18. M. Perez, F. Reventos, L. Batet, A. Guba, I. Tóth, T. Mieusset, 

P. Bazin, A. de Crécy, S.Borisov, T. Skorek, H. Glaeser, J. Joucla, 

P. Probst, A. Ui, B.D. Chung, D.Y. Oh, R. Pernica, M. Kyncl, 

J. Macek, A. Manera, J. Freixa, A. Petruzzi, F. D’Auria, 

A. Del Nevo, “Uncertainty and sensitivity analysis of a LBLOCA 

in a PWR Nuclear Power Plant: Results of the Phase V of the 

BEMUSE programme”, Nuclear Engineering and Design 241, 

pp. 4206-4222 (2011) 

19. J.C. Helton, F.J. Davis “Latin hypercube sampling and the 

propagation of uncertainty in analyses of complex systems”, 

Reliability Engineering & System Safety 81, pp. 23-69 (2003) 

20. K.A. Gamble, L.P. Swiler “Uncertainty Quantification and 

Sensitivity Analysis Applications to Fuel Performance 

Modeling”, SAND2016-4597C 

21. F. Mascari, F. De Rosa, B.G. Woods, K. Welter, G. Vella, 

F. D’Auria, “Analysis of the OSU-MASLWR 001 and 002 Tests by 

Using the TRACE Code”, NUREG/IA-0466 (2015) 

22. T. Iguchi, T. Okubo, Y. Murao, “Effect of Loop Seal on Reflood 

Phenomena in PWR”, Journal of Nuclear Science and 

Technology 25(6), pp. 520-527 (1988) 

23. P. Coppolani, N. Hassenboehler, J. Joseph, J.F. Petetrot, J.P. Py, 

J.S. Zampa, La chaudière des réacteurs à eau sous pression, 

EDP Sciences, Les Ulis, France (2004) 

24. F. Mascari, J.C. De La Rosa Blul, M. Sangiorgi, G. Bandini, 

“Analyses of an unmitigated station blackout transient with 

ASTEC, MAAP and MELCOR code”, 9th Meeting of the 

“ European MELCOR User Group” (2017) 

25. F. D’Auria, M. Frogheri, W. Giannotti “RELAP/MOD3.2 Post Test 

Analysis and Accuracy Quantification of SPES Test SP-SB-04”, 

NUREG/IA-0155 (1999) 

26. O.C. Jones, Nuclear Reactor Safety Heat Transfer, Hemisphere 

Publishing Corporation, Washington, USA (1981) 

27. J.L. Rodgers, W.A. Nicewander, “Thirteen Ways to Look at the 

Correlation Coefficient”, The American Statistician 42(1), 

pp. 59-66 

Authors 

Andrea Bersano 

Energy Department (DENERG) 

Politecnico di Torino 

Corso Duca degli Abruzzi 24 

10129, Turin 

Italy 

Fulvio Mascari 

Nuclear Safety 

Security and Sustainability Division 

(FSN-SICNUC), ENEA 

Via Martiri di Monte Sole 4 

40129, Bologna 

Italy 




Experiment Research on the Insurge 

Transient Behavior of Gas-steam 

Pressurizer under Various Pressure 

Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun 

For small and medium sized reactors with gas-steam pressurizer, the transient behavior of gas-steam pressurizer plays 

a vital role on the safety of the nuclear reactor operation. This paper focuses on the transient behavior of gas-steam 

pressurizer. The pressure response was investigated under various insurge experimental conditions, and the influence 

factors of pressure changes were analyzed. 

Research shows initial pressure of the pressurizer and the presence of a non-condensable gas with varying mass 

fraction will all have some kind of effect on the transient behavior of gas-steam pressurizer. Initial pressure decides initial 

magnitude of liquid temperature of the pressurizer, and the presence of a non-condensable gas with varying mass 

fraction can greatly affect the heat and mass transfer process both wall and interface, and further affect the system 

pressure variations. This paper focus on the effects on pressurizer insurge transients under various initial pressure of the 

pressurizer and the presence of a non-condensable gas with varying mass fraction. 

Introduction 

A transient can be caused from a 

simple loss of secondary steam flow to 

more complicated accidents. After 

the Three Mile Island accident [1], 

Chernobyl accident [2] and Fukushima 

nuclear accident [3]. Many people 

realized the importance of small break 

LOCA’s and the necessity for having 

reliable physical models for all of the 

components in the loop. 

In view of increasing pressure on 

energy restructuring and the serious 

environmental pollution concerns, 

the integrated natural circulation 

reactor is receiving a great attention 

for its ability to provide energy that is 

clean, safe and economic. And the integrated 

reactor is going to be one of 

the best option for small and medium 

sized reactors (SMRs) in lots of countries 

with the high level of great safety 

and reliability. SMRs have been developed 

in many countries for small-scale 

power generation, district heating, 

and seawater desalination. Argentine 

CAREM reactor, Russian VVER-300/ 

VK-300 reactor, Korean REX-10/ 

SMART reactor, Japanese IMR reactor, 

American NuScale reactor, Chinese 

NHR reactor and so on, these are the 

typical representatives of the SMRs 

[4]. In order to simplify the structure 

and design and enhance safety, the 

steam-gas pressurizer is generally 

utilized in the integrated reactor. The 

non-condensable gas is used to keep 

the pressure stable in the steam-gas 

pressurizer and the transient behavior 

of gas-steam pressurizer plays a vital 

role on the safety of the nuclear reactor 

operation. As so far, early pressurizer 

transient models were generally 

developed under certain conditions, 

slow insurge velocity or low pressure 

for example. And these models may 

not be applicable to various velocity or 

create a greater risk of inaccurate 

results under the high pressure. 

Westinghouse [5] developed the 

TOPS pressurizer model included 

the effects of wall condensation by applying 

Nusselt Laminar film theory 

to estimate a wall heat transfer 

coefficient. Saedi [6] investigated the 

relative magnitudes of the physical associated 

with insurge transients, and 

initiated a data base (at low pressure) 

for a model, developed 

by Kim [7], at MIT. In some other 

pressurizer analysis, Mark [8] came 

up with the effects of the presence of a 

non-condensable gas on insurge 

transient. Leonard [9] performed the 

experiment on the pressure behavior 

of steam-gas pressurizer during the 

insurge and he focused on the various 

non-condensable gas, various gas 

content and stratification in his study. 

Kim [10] found the condensation heat 

transfer at wall is an important physical 

phenomenon during the insurge 

transient. Paulsen [11] improved and 

developed the theoretical modeling 

for RELAP5. Wu lei [12, 13] and Ma 

Xizhen [14, 15] established the nonequilibrium 

gas-steam pressurizer 

model and improved the steam condensation 

heat transfer model in 

presence of non-condensable gas. 

In order to understand and model 

an accident, one should recognize the 

processes that take place during a 

transient. In general, these processes 

include insurges, outsurges, and combined 

insurges and outsurges. Early 

pressurizer transient models were 

generally developed under certain 

conditions, slow insurge flow rate or 

low pressure for example, and these 

models may not be applicable under 

various pressure. At present, the 

transient behavior of steam-gas pressurizer 

under various pressure need to 

be further studied in the integrated 

natural circulation reactor with 

steam-gas pressurizer. Through the 

design and establish of the experiment 

system, the combination of 

theoretical and experimental which 

can provide the data support for the 

design and operation of the integrated 

natural circulation reactor. 

1 Description of 

Experimental System 

To find out the experimental physical 

processes occurs in a pressurizer, an 

experimental apparatus has been 

built (Figure 1). 

The primary tank (2420 mm high 

and 450 mm ID) which models the 

pressurizer volume equipped with a 

magnetic level gauge to accurately 

measure the initial liquid level and 

the change of liquid level during the 

insurge. The storage tank (2490 mm 

high and 400 mm ID) is filled with 

cold water to a level also measured 

with a magnetic level gauge and 

pressurized with nitrogen. This tank 

| Fig. 1. 

Schematic diagram of apparatus. 

Nomenclature 

p t system pressure 

T w temperature at each 

measured point of gasphase 

space 

p N2 

nitrogen partial pressure 

p s steam partial pressure 

n N2 

nitrogen amount of 

substance 

n s steam amount of substanc 

V g gas-phase space volume 

Z N2 

nitrogen compressibility 

factor 

Z s 

steam compressibility 

factor 

x N2 

nitrogen mole fraction 

T g average temperature in 

gas-phase space 



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



| Fig. 2. 

Heating rod installation method and voltage regulator. 

| Fig. 3. 

Data acquisition system and computer data network. 

serves as a reservoir for “cold” injection 

water. 

Two tanks are both equipped with 

nine heating rods totaling 36 kW and 

the electric heating system utilizes the 

TSGC2J-40 type voltage regulator to 

adjust the load voltage of the heating 

rod, and then control the heating rod 

power to meet the requirements 

of each experimental condition 

(Figure 2). 

Besides, the experimental system 

also consists of electric heating system 

and data measurement and acquisition 

system (Figure 3). 

2 Experimental conditions 

and parameter control 

Several precautionary procedures were 

listed as followed prior to initiating 

each of the transient experiments. The 

primary and storage tanks were filled 

to the required levels with deionized 

water. Steam was bubbled through the 

primary tank to assist the heating rods 

in bringing the tank up to saturation 

conditions. The relief valve was opened 

and the water allowed to boil for a 

while when the primary tank reached 

saturation. This process was intended 

to rid the system of dissolved gases in 

the water. For the experiments with an 

initial non-condensable gas fraction in 

the vapor, the gas was injected into the 

vapor space after the tank had been 

purged of dissolved gases, and before 

the system had reached the desired 

operating pressure. 

The partial pressure of the steam 

and the temperature of gas-phase 

space, which are directly determined 

by nitrogen mole fraction in the 

pressurizer. Therefore, the experiment 

should put strict controls on its 

initial value. Based on the real gas 

equations, the nitrogen mole fraction 

can be obtained by measuring the 

temperature of gas-phase space and 

system pressure. It should be noted 

that the calculation leaves out of 

account stratification of hot gas and 

steam superheating phenomenon. 

The concrete calculating methods are 

as follows: measurement parameters: 

system pressure p t ,the temperature at 

each measured point of gas-phase 

space in the pressurizer T w1 , T w2 … T wn . 

Based on the real gas equations, 

the nitrogen mole fraction can be 

obtained by measuring the temperature 

of gas-phase space and system 

pressure. It should be noted that the 

calculation leaves out of account 

stratification of hot gas and steam 

superheating phenomenon. The 

concrete calculating methods are as 

follows: 

Measurement parameters: system 

pressure, the temperature at each 

measured point of gas-phase space in 

the pressurizer T w1 , T w2 … T wn . 

Middle parameters: nitrogen 

partial pressure p N2 

, steam partial 

pressure p s , nitrogen amount of 

substance n N2 

, steam amount of substance 

n s , gas-phase space volume V g , 

nitrogen compressibility factor Z N2 

, 

steam compressibility factor Z s . 

The average temperature T g in gasphase 

space can be obtained by 

measuring the temperature at each 

measured point of gas-phase space 

in the pressurizer. 

The real gas equations of steam and 

nitrogen are as follows where R indicates 

gas constant: 

Then, it can come to the nitrogen mole 

fraction X n2 

. 

= 

| Fig. 4. 

Computing framework. 

The computing framework is demonstrated 

by Figure 4. 




Parameter Value Unit 

Non-condensable gas species Nitrogen / 

Non-condensable gas mass fraction 0, 10%, 20% / 

Regulator space initial pressure 0.5, 1.0, 1.5 MPa 

Initial height of liquid level 800 mm 

Final height of liquid level 1200 mm 

Insurge time 20 s 

| Tab. 1. 

Specific experimental conditions. 

The data acquisition system was 

initiated prior to beginning the 

insurge to allow some steady data to 

be taken, the specific experimental 

conditions are shown in Table 1. 

3 Experimental results and 

analysis 

There are two main reasons affected 

the pressure in a transient experiment, 

the first part is the increase of 

liquid level plays a role that compresses 

gas-phase space and the 

pressure is going up, the second part is 

a higher initial pressure leading to a 

higher initial liquid temperature of 

the pressurizer and affect the heat and 

mass transfer process both wall and 

interface, and further affect the system 

pressure variations. The pressure 

variation is a combination of the two 

aspects. 

The pressure and temperature 

histories for the experiments with 

20 % mass fraction nitrogen are 

shown in Figure 5 and Figure 6. To 

make conclusions of trials more 

comparable, this analysis applies 

dimensionless method to the system 

pressure, and pressure variation 

was represented by the ratio of 

system pressure to initial pressure. 

Figure 7 shows the dynamic change 

process of dimensionless pressure 

corresponding to different initial 

pressures. 

The system pressure growth rate 

and the initial pressure of the pressurizer 

have positive correlation. The 

higher the initial pressure, the faster 

the system pressure growth rate in the 

transient. 

The presence of a non-condensable 

gas with varying mass fraction will all 

have some kind of effect on the 

transient behavior of gas-steam 

pressurizer. At the same system pressure, 

the main effects of the presence 

of a non-condensable gas are given as 

follows: 

I 

Affect the nitrogen partial pressure, 

the steam partial pressure 

and the vapor temperature of the 

pressurizer. 

II Affect the compressibility of the 

vapor space in the pressurizer. 

III Affect the heat and mass transfer 

process of the vapor space in the 

pressurizer. 

The pressure histories for the experiments 

with various mass fraction 

nitrogen are shown in Figure 8, 

Figure 9 and Figure 10. The initial 

primary tank non-condensable gas 

mass fraction for the transient is 

labeled in the figure. The most obvious 

difference in the results is the large 

variation in peak pressure, the higher 

initial non-condensable gas mass 

fraction in the primary tank, the 

higher the peak pressure in the 

transient. And the specific values are 

shown in Table 2. 


| Fig. 5. 

The system pressure histories with 20 % mass fraction nitrogen. 

| Fig. 6. 

The temperature histories with 20 % mass fraction nitrogen. 

| Fig. 7. 

The dimensionless pressure histories with 20 % mass fraction nitrogen. 

| Fig. 8. 

The pressure histories with various mass fraction nitrogen under 0.5MPa pressure. 





Initial 

system pressure 

| Tab. 2. 

Specific experimental values. 

| Fig. 9. 

The pressure histories with various mass fraction nitrogen under 1.0 MPa 

pressure. 

Non-condensable gas 

mass fraction 

Peak 

pressure 

0.5 0 0.55 

0.5 10 % 0.59 

0.5 20 % 0.61 

1.0 0 1.08 

1.0 10 % 1.21 

1.0 20 % 1.25 

1.5 0 1.69 

1.5 10 % 1.87 

1.5 20 % 1.92 

In particular, the research demonstrated 

that when the pressurizer is 

free from non-condensable gas, 

may present the twice peak pressure 

phenomenon during the transient, 

and the first peak pressure will present 

about the same time. While only 

one peak pressure will present during 

the transient containing the noncondensable 

gas. 

There are two main reasons 

affected the pressure in a transient 

experiment, the first part is the increase 

of liquid level plays a role that 

compresses gas-phase space and the 

pressure is going up, the second part is 

steam condensation leading to a decrease 

pressure at some degrees. The 

peak pressure is combination of the 

two aspects. In the early stage of the 

transient, the compression of the 

vapor space plays leading roles and 

the pressure is going up. At this stage, 

there was no distinct effect on the temperature 

of the interface by incoming 

the cold water during the transient. In 

the second stage, steam condensation 

leads to a decrease pressure at some 

degrees due to the temperature reduction 

of the interface. And in next stage, 

the compression of the vapor space 

also plays leading roles comparing 

with the steam condensation and the 

pressure is going up. 

Conclusion 

Based on the proceeding analysis and 

experiments, an experimental study 

of the pressure response during an 

insurge transient under various pressure 

has been performed. The following 

conclusions are drawn based upon 

the experimental and analytical 

results: 

I 

The system pressure growth rate 

and the initial pressure of the pressurizer 

have positive correlation. 

The higher the initial pressure, the 

faster the system pressure growth 

rate in the transient. 

II The changes in system temperature 

coincided well with those of 

system pressure. 

III The higher initial non-condensable 

gas mass fraction in the primary 

tank, the higher the peak pressure 

in the transient. 

IV In particular, when the pressurizer 

is free from non-condensable gas, 

may present the twice peak pressure 

phenomenon during the transient, 

and the first peak pressure 

will present about the same time. 

While only one peak pressure will 

present during the transient containing 

the non-condensable gas. 

Acknowledgments 

Special thanks should go to Mr Ma 

Xizhen who have put considerable 

time and effort into the designing and 

constructing work of the experiment 

plant. 

References 

[1]. Bot P L. Human reliability data, human error and accident 

models—illustration through the Three Mile Island accident 

analysis[J]. Reliability Engineering & System Safety, 2004, 

83(2):153-167. 

[2]. Jr H T P. Summary report on the post-accident review meeting 

on the chernobyl accident : International Nuclear Safety Advisory 

Group. International Atomic Energy Agency (IAEA) Safety Series 

No. 75-INSAG-1 (STI/PUB/ 740), Vienna, IAEA, 1986. 260 Austrian 

Schillings[J]. Journal of Environmental Radioactivity, 1987, 

5(5):403-404. 

| Fig. 10. 

The pressure histories with various mass fraction nitrogen under 1.5 MPa 

pressure. 

[3]. Kinoshita N, Sueki K, Sasa K, et al. Assessment of individual 

radionuclide distributions from the Fukushima nuclear accident 

covering central-east Japan[J]. Proceedings of the National 

Academy of Sciences of the United States of America, 2011, 

108(49):19526-19529. 

[4]. International Atomic Energy Agency: Advances in small modular 

reactor technology developments. Vienna: the IAEA in Austria,2014. 

[5]. Redfield, J. A., Prescop, V., & Margolis, S. G. (1968). Pressurizer 

performance during loss-of-load tests at shippingport: analysis 

and test. , 4(3), 173-181. 

[6]. Saedi, H. R. (1982). Insurge pressure response and heat 

transfer for PWR pressurizer. Massachusetts Institute of Technology. 

[7]. Kim, S. N. (1984). An experimental and analytical model of a 

pwr pressurizer during transients. British Journal of Surgery, 

87(12), 1615–1616. 

[8]. Leonard, M. T., & Griffith, P. (1983). The effects of a 

noncondensable gas on pressurizer insurge transients. Trans. 

Am. Nucl. Soc.; (United States), 46(6), 844-845. NOMURA Katsuya, 

K.S.O.Y., Numerical analysis of droplet breakup behavior using 

particle method. 1999. 38(12): p. 1057-1064. 

[9]. Leonard M T, Griffith P. The effects of a non-condensable gas 

on pressurizer insurge transients[J]. Trans. Am. Nucl. Soc.; (United 

States), 1984, 46(6):844-845. 

[10]. Kim, S.N., Griffith, P., 1987. PWR pressurizer modeling. Nucl. 

Eng. Des.102, 199–209. 

[11]. Paulsen, M.P., et al., 1996. RETRAN-3D—a program for transient 

thermal–hydraulic analysis of complex fluid flow systems. 

Electric Power Research Institute, NP-7450. 

[12]. WU Lei JIA Hai-jun LIU Yang MA Xi-zhen. Transient Characteristics 

of Integrated Non-condensable Gas-steam Pressurizer [J], 

Atomic Energy Science and Technology, 2014, 48(s1):200-207. 

[13]. WU Lei LIU Yang JIA Hai-jun YANG Xing-tuan. Research on 

Steam Condensation Heat Transfer Model in Presence of Noncondensable 

Gas at High Pressure [J]. Atomic Energy Science and 

Technology, 2016, 50(2):261-266. 

[14]. MA Xi-zhen JIA Hai-jun LIU Yang WU Lei. Numerical 

Simulation Study on Steam Condensation in Presence of Noncondensable 

Gas [J]. Atomic Energy Science and Technology, 

2015, 49(s1):265-269. 

[15]. MA Xi-zhen JIA Hai-jun LIU Yang WU Lei. Effect of Noncondensable 

Gas on Steam-gas Pressurizer [J]. Atomic Energy 

Science and Technology, 2016, 50(9):1586-1591. 

Authors 

Wang Bolong 

Li Weihua 

Jia Haijun 

Li Jun 

Zhang Yajun 

Institute of Nuclear and New Energy 

Technology 

Tsinghua University 

Beijing 100084, China 

Collaborative Innovation Center of 

Advanced Nuclear Energy Technology 



Key Laboratory of Advanced Reactor 

Engineering and Safety of Ministry of 

Education 






Stilllegung und Rückbau des 

Rossendorfer Forschungsreaktors RFR 

Teil 1: Objektbeschreibung, Genehmigungsverfahren, Ausgangssituation, 

Planungskonzept und Meilensteine 

Reinhard Knappik, Klaus Geyer, Sven Jansen und Cornelia Graetz 

Mit der Entlassung des Rossendorfer Forschungsreaktors im September 2019 aus dem Geltungsbereich des Atomgesetzes 

(AtG) sind die Stilllegung und der Rückbau der alten kerntechnischen Anlagen des ehemaligen Zentralinstituts für Kernforschung 

(ZfK) der Akademie der Wissenschaften der DDR bis auf ein ca. 50 m langes Reststück einer Rohrleitung der 

Speziellen Kanalisation abgeschlossen. 

In der zweiteiligen Veröffentlichung werden die Arbeiten zur Stilllegung und zum Rückbau des Rossendorfer 

Forschungsreaktors (RFR) übersichtsartig vorgestellt, wobei im Teil 1 (atw 11-12 2019) anknüpfend an die Objektbeschreibung 

die Genehmigungsverfahren, die Ausgangssituation radiologisch und konventionell, das realisierte Planungskonzept 

sowie die Meilensteine vor gestellt werden. Im Teil 2 (atw 1/2020) wird auf ausgewählte Aspekte der Stilllegung- und 

Rückbaudurchführung, zum Strahlenschutz, zur Freigabe sowie zum Reststoff- und Abfallmanagement eingegangen. 

1 Einleitung Die Geschichte des Rossendorfer Forschungsreaktors und damit des Forschungsstandortes 

Rossendorf bei Dresden (jetzt Dresden-Rossendorf) begann mit dem „Abkommen über die Hilfeleistung der Union der 

Sozialistischen Sowjetrepubliken an die Deutsche Demokratische Republik bei der Entwicklung der Forschung auf dem 

Gebiet der Physik des Atomkerns und der Nutzung der Kernenergie für die Bedürfnisse der Volkswirtschaft“, das am 

28. April 1955 abgeschlossen wurde. Auf dieser Grundlage erfolgte die Lieferung zweier Großgeräte, ein 2 MW 

Forschungsreaktor vom Typ WWR-S (Abb. 1) und ein 25 MeV-Zyklotron, an das am 1. Januar 1956 gegründete 

Zentral institut für Kernphysik (später Zentralinstitut für Kernforschung). 

Am 14. Dezember 1957 wurde der 

erste Forschungsreaktor der DDR 

nach einer Bauzeit von nur 21 

Monaten erstmals kritisch und am 

16. Dezember 1957 offiziell im Rahmen 

eines Staatsaktes in Betrieb 

genommen. Der RFR war ein leichtwassermoderierter 

und -gekühlter 

Tankreaktor. Über einen Zwischenschritt 

(von 2 MW auf 4 MW und 

5 MW im Jahre 1965) erfolgte zuletzt 

eine Leistungs erhöhung auf 10 MW 

(1967), die ab 1981 im Dauerbetrieb 

realisiert wurde. Die Leistungserhöhung 

auf 10 MW konnte nur durch 

Einsatz neuer Brennstäbe mit einer 

Anreicherung von 36 % U-235 gegenüber 

vorher von 10 % U-235 erreicht 

werden. Die Energieabgabe des RFR 

betrug insgesamt rund 28.000 MWd 

bei einem Leistungseinsatz von rund 

105.000 Stunden. Es gab während der 

Betriebszeit des RFR kein Ereignis, 

welches strahlenschutzrelevant war. 

Interessante Details über die RFR- 

Betriebszeit enthält das Buch „Beiträge 

zur Geschichte des Rossendorfer 

Forschungsreaktors RFR“ [1]. 

Im Zuge der weiteren Entwicklung 

des Forschungsstandortes entstanden 

eine Vielzahl von weiteren Anlagen 

und Einrichtungen in denen eine kerntechnische 

Nutzung bis 1991 erfolgte, 

wobei hier der RFR nur bis zum 

27. Juni 1991 betrieben wurde. Mit 

der Neuordnung des Forschungsstandortes 

nach der Wiedervereinigung und 

der Auflösung der Akademie der 

Wissenschaften zum 31. Dezember 

1991 wurden dann auf der Grundlage 

mehrerer in den Jahren 1993 und 

1996 vollzogener Kabinettsbeschlüsse 

mit der Stilllegung und dem Rückbau 

der kerntechnischen Anlagen am 

Forschungsstandort begonnen. Der 

Freistaat Sachsen beauftragte den 

Verein für Kern verfahrenstechnik und 

Analytik Rossendorf e. V. (VKTA, Umbenennung 

in „VKTA – Strahlenschutz, 

Analytik & Entsorgung Rossendorf 

e. V.” im Dezember 2014) mit der vollständigen 

Beseitigung der nuklearen 

Altlasten des Forschungsstandortes 

Rossendorf. Im Folgenden werden die 

wichtigsten Rückbauobjekte mit 

Termin der Ent lassung (in Klammern) 

aus dem Geltungsbereich des Atomgesetzes 

aufgeführt: 

p die Rossendorfer Anordnung für 

kritische Experimente (RAKE, 

1998) 

p das Urantechnikum (2000) 

p der Rossendorfer Ringzonenreaktor 

(RRR, 2000) 

p die Spezielle Kanalisation am 

Forschungsstandort (AtG-Teilentlassungen 

2010, 2013 und 2018) 

p die Objekte und Anlagen der Isotopenproduktion 

(von 2009 bis 

2014, [2]) 

p die Anlagen und das Gelände der 

Entsorgungs wirtschaft des ehemaligen 

ZfK (2005, 2011 und 2018) 

p der RFR (2009, 2010, 2019) 

| Abb. 1. 

Ansicht des Forschungsreaktors während der Fertigstellungsphase. 

Die Kosten für Stilllegung, Rückbau 

und Entsorgung werden vollständig 

vom Freistaat Sachsen getragen, 

sodass die Mittelbereitstellung auch 

Einfluss auf die terminlichen Arbeitsabläufe 

hatte. So mussten in den 

Jahren 2005 und 2006 die Rückbauarbeiten 

aufgrund der zu knapp bemessenen 

Finanzmittel durch den 

Freistaat Sachsen eingestellt werden, 

was zu entsprechend längeren Rückbauzeiten 

führte, da z. B. Mitarbeiter 

nicht mehr zur Verfügung standen. 

2 Objektbeschreibung 

Zu den RFR-Anlagen (Abb. 2) gehörten 

der „Rückbau komplex RFR“ 

mit einer Fläche von ca. 9.000 m 2 und 

den Objekten: 

p Labortrakt (1) mit Reaktorwarte 

(2) 

537 

DECOMMISSIONING AND WASTE MANAGEMENT 


Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz


DECOMMISSIONING AND WASTE MANAGEMENT 538 

| Abb. 2. 

Territoriale Lage der RFR-Anlagen. 

p Reaktorhalle mit Vorhalle und 

Anbau (3) 

p Pavillon und einen weiteren 

Anbau (4) 

p Ventilations- und Filtergebäude 

mit Fortluftschornstein (5) 

p Schauer (6) 

sowie die außerhalb des Komplexes 

liegenden Objekte: 

p Trafostation (7) 

p Notstromgebäude (8) 

p altes Armaturenhaus (9) 

p Armaturenhaus (10) 

p Trockenkühlturm 1 und 2 (11) 

p Rohrleitungen des 2. Kühlkreislaufes 

(12) 

Weiterhin gab es außerhalb noch die 

Sammelbehälter anlage mit der Pumpbedienstation, 

in der die radioaktiven 

Abwässer des RFR aufgefangen 

wurden. Diese Anlagen befanden sich 

im Kontrollbereich der Entsorgungswirtschaft 

und wurden im Rahmen 

dieses Rückbaukomplexes aus der 

atomrechtlichen Aufsicht entlassen. 

Die territoriale Lage der nummerierten 

RFR-Anlagen ist schematisch in der 

Abb. 2 dargestellt. Im Betriebshof befanden 

sich befestigte Verkehrsflächen, 

Rohre der Regen- und Schmutzwasserkanalisation 

sowie Leitungen für 

konta minationsverdächtige Abwässer 

bis zu einer Tiefe von 5 m, unterirdische 

Abluftleitungen, Heizungskanäle, 

Schächte sowie Medienleitungen. 

Im Folgenden werden wichtige 

Objekte kurz charakterisiert: 

Labortrakt mit Reaktorwarte 

(1 und 2) 

Der Labortrakt war ein viergeschossiger 

Ziegelbau mit Stahlbetondecken, 

dessen Grundfläche ca. 960 m 2 

betrug. In diesem Gebäude waren 

ursprünglich Laboratorien, die Reaktorwarte 

und Büroräume untergebracht. 

Im Zuge der Erneuerung 

des RFR wurde eine neue Warte in 

Stahlbeton- Skelettbauweise mit einer 

Fläche von ca. 260 m 2 an der NO-Seite 

des Labortraktes angebaut. 

Reaktorhalle (3) 

Die Reaktorhalle war ein unterkellerter 

Ziegel-Stahl- Skelettbau mit 

einer Gesamthöhe von 24,5 m und 

einer Grundfläche von ca. 700 m 2 . 

Darin befanden sich neben dem 

Reaktor unter anderem Brennelemente-Lagerbecken 

(AB 1 und AB 2) 

sowie im Keller geschoss vier Heiße 

Kammern. 

Pavillon (4) 

Der Gebäudebereich „Pavillon“ mit 

einer Grundfläche von ca. 170 m 2 

bildete einen stark gegliederten einbzw. 

zweigeschossigen Anbau, unter 

dem Abluftkanäle zum Filter- und 

Ventilationsgebäude sowie Rohrleitungen 

für kontaminierte Abwässer 

verliefen. Im Gebäudebereich waren 

ein radiochemisches Labor und eine 

Gamma- Bestrahlungsanlage untergebracht. 

Ventilations- und Filtergebäude (5) 

Das Filter- und Ventilationsgebäude 

mit einer Grundfläche von ca. 350 m 2 

war ein anderthalbgeschossiger 

Ziegel bau mit einem Flachdach, auf 

dem sich der ca. 33 m hohe Fortluftschornstein 

befand. Im Inneren 

waren die Venti lationskammern mit 

den dazugehörigen Filteranlagen. 

Altes Armaturenhaus (9) 

Das Gebäude mit einer Grundfläche 

von ca. 22,7 m x 7,2 m und einer Höhe 

von ca. 4,5 m hatte über zwei Drittel 

der Grundfläche eine ca. 3,5 m tiefe 

Grube mit dem Pumpenfundament. 

Das Gebäude bestand aus Mauerwerk; 

die Fundamente und Teile der 

aufgehenden Wände aus Stampfbeton. 

Armaturenhaus (10) 

Das Armaturenhaus mit einer 

Abmessung von ca. 18,4 m x 12,4 m x 

6,0 m wurde im Rahmen der Gesamtrekonstruktion 

RFR im Jahr 1985 neu 

errichtet und beinhaltete Pumpen, 

Armaturen und eine Reinigungs anlage 

für das Sekundärkühlsystem. Es 

handelte sich um eine Stahl betonskelettkonstruktion 

aus Fertigteilen, 

| Abb. 3. 

Schnittdarstellung RFR-Gebäudekomplex. 




wobei die Außenwände aus Gasbetonfertigteilen 

bestanden. 

Trockenkühltürme (11) 

Die Trockentürme hatten eine Abmessung 

von ca. 23,3 m x 13,2 m x 8,0 m. 

Sie bestanden aus einem Traggerüst 

aus Stahl, das mit Leichtmetallelementen 

verkleidet war. 

Die umbaute Fläche des RFR ohne 

Filter- und Venti lationsgebäude betrug 

ca. 2.100 m 2 , wobei der Kontrollbereich 

die Reaktorhalle mit Kellergeschoss, 

Teile des Kellergeschosses des 

Anbaus der Reaktorhalle und Teile des 

Kellergeschosses des Labortraktes 

umfasste. Die Schnittdarstellung in 

der Abb. 3 vermittelt einen Eindruck 

von der baulichen Anordnung. 

Neben dem Reaktor (Detaildarstellung 

nach Re konstruktion in Abb. 4) 

waren die vier Heißen Zellen, das 

Lagerbecken AB 2 und der Pumpenraum 

(vgl. Abb. 5) hinsichtlich des 

Rückbaus schwierige Teilobjekte. 

Aus der Betriebshistorie waren für 

die ersten Schritte des Rückbaus und 

der Entsorgung vor allem zu berücksichtigen: 

p betriebliche und strahlenschutzrelevante 

Ereignisse 

p der Einsatz von Brennelementen 

mit einem hohen Anreicherungsgrad 

p die nahezu vollständige Erneuerung 

des RFR in den Jahren 1986 

bis 1989 mit der Wiederinbetriebnahme 

am 27. Januar 1990 sowie 

die endgültige Einstellung des 

nuklearen Betriebes des RFR am 

27. Juni 1991 bedingt durch eine 

befristete Genehmigung zum 

Versuchsbetrieb. 

3 Genehmigungsverfahren 

Im Jahre 1991 wurden Aufsichtliche 

Anordnungen gemäß § 19 Absatz 

3 AtG durch das Sächsische 

Staatsministerium für Umwelt und 

Landesentwicklung (SMU) erlassen, 

um bedingt durch die Auflösung der 

Institute der Akademie der Wissenschaften 

der DDR zum 31. Dezember 

1991 letztendlich einen ungeregelten 

Zustand für den RFR zu vermeiden 

sowie am 19. Dezember 1991 den 

Betreiberwechsel zum VKTA vorzunehmen. 

1993 fasste die Sächsische 

Staatsregierung den Kabinettsbeschluss, 

den Forschungsreaktor endgültig 

stillzulegen und zurück zubauen. 

Um das Genehmigungsverfahren 

nach § 7 Absatz 3 AtG einzuleiten, 

stellte der VKTA bereits im Dezember 

1994 beim Sächsischen Staatsministerium 

für Umwelt und Landesentwicklung 

einen Antrag auf Genehmigung 

zur Stilllegung und zum Abbau 

| Abb. 4. 

Schnittdarstellung RFR. 

des RFR. Von 1993 bis 1998 wurden 

technische, sicherheitstech nische und 

strahlenschutztechnische Maßnahmen 

zur Anpassung an den bundesdeutschen 

Standard durch geführt. 

Hervorzuheben ist die Erstellung 

eines brandschutztechnischen Gutachtens 

durch eine Fremdfirma, auf 

deren Grundlage bautechnische Maßnahmen, 

die Reduzierung der Brandlast 

sowie die Inbetriebnahme einer 

neuen Brandschutzanlage erfolgten. 

Weiterhin mussten am Forschungsstandort 

Rossendorf erst die Voraussetzungen 

geschaffen werden, um die 

Kernbrennstoffe bzw. Kernmaterialien 

verwahren, die radioaktiven Abfälle 

und Reststoffe zwischenlagern, 

behandeln, analysieren und freigeben 

zu können. Dies erforderte zahlreiche 

atom- und strahlenschutzrechtliche 

Genehmigungen sowie deren bauliche 

Umsetzung, die in den Jahren 

1992 bis 1999 ebenfalls stattfanden. 

Hervorzuheben sind hierbei die Einrichtung 

zur Entsorgung von Kernmaterial, 

die Transportbereitstellungshalle 

für die CASTOREN, das 

Zwischenlager, die Reststoffbehandlungsanlage, 

das Freimesszentrum, 

das Analytiklabor sowie die notwendigen 

Einrichtungen für den 

Strahlenschutz, wie die Inkorporationsmessstellen 

sowie für die 

| Abb. 5. 

Schnittdarstellung RFR. 

Umgebungsüberwachung. Nach Erhalt 

der ersten beiden Genehmigungen 

zur RFR-Stilllegung [3, 4] wurden 

von 1998 bis 2001 die Anlagen kernbrennstoff- 

und kernmaterialfrei [5, 

6, 7] gefahren und für den VKTA mit 

dem Rückbau des 2. Kühlwasserkreislaufes 

der Abschluss der Stilllegungsphase 

erreicht. Von 2001 bis 2018 

erfolgten dann der Rückbau sowie die 

Geländesanierung mit abschließender 

Flächenprofilierung im Rahmen 






der Dritten und Vierten RFR- 

Stilllegungsgenehmigungen. 

Die Grundlage für die Stilllegung 

und den Rückbau des RFR waren insgesamt 

vier Genehmigungen mit ihren 

Änderungen. Die Erste Genehmigung 

45-4653.18 VKTA 01 zur Stilllegung 

des Rossendorfer Forschungsreaktors 

RFR [3] wurde am 30. Januar 1998 

durch das SMU erteilt. Deren 1. Änderungsgenehmigung 

45-1653.18 VKTA 

01/1, erteilt vom Sächsischen Staatsministerium 

für Umwelt und Landwirtschaft 

(SMUL) am 06. November 

2000, bein hal tete vor allem das Innehaben 

der endgültig abgeschalteten 

Anlage, die sichere Betriebsführung 

der abgeschalteten Anlage zum Zwecke 

der Stilllegung, die Überführung 

von Brennelementen aus der Spaltzone 

in das Brennelement lagerbecken 

AB 2 und den innerbetrieblichen 

Transport. 

Mit der Zweiten Genehmigung 

45-4653.18 VKTA 02 des SMUL vom 

30. Oktober 1998 zur Stilllegung des 

Rossendorfer Forschungsreaktors 

RFR [4] (eingereicht im Oktober 

1997) konnte der Rückbau des 2. 

Kühlkreislaufes realisiert werden, 

deren 1. Änderung vom 11. Februar 

beinhaltete die Erweiterung des 

Genehmigungsumfanges bzgl. eines 

Raumes im RFR-Labortrakt. 

| Abb. 6. 

Überblick zum Aktivitätsinventar. 

Die Dritte Genehmigung 4653.18 

VKTA 03 zur Still legung und zum 

Abbau des Rossendorfer Forschungsreaktors 

RFR des SMUL vom 3. April 

2001 [8] ermög lichte die Entsorgung 

der Betriebsmedien sowie die Außerbetriebnahme 

und den Rückbau der 

nicht mehr benötigten Systeme und 

Komponenten des RFR. Dazu zählte 

z. B. der Rückbau des 1. Kühlkreislaufes, 

des Speisewassersystems und 

des Reaktorbehälters. Insgesamt gab 

es 14 Vorhaben, wobei mit Erteilung 

der Genehmigung bereits für vier 

Vorhaben die Zustimmungen des 

SMUL vorlagen. Die restlichen Vorhaben 

wurden abbaubegleitend mit 

dem SMUL abgestimmt. 

Mit der Vierten Genehmigung 

4653.18 VKTA 04 des SMUL vom 

1. Februar 2005 [9] konnte schließlich 

der Abbau der Restanlage des RFR vorgenommen 

werden. Der Änderungsbescheid 

4653.18 VKTA 04/1 vom 

9. November 2010 beinhaltete die Freigabe 

und Ent lassung des Raumes 01 im 

Gebäude 103 (Notstrom gebäude) und 

mit der 2. Änderungsgenehmigung 

4653.18 VKTA 04/2 des SMUL vom 

9. Januar 2014 wurden die Änderungen 

des räumlichen Geltungsbereiches 

sowie des Genehmigungsumfanges 

beschieden. Hintergrund war die Entscheidung 

des VKTA aufgrund von 

radiologischen Voruntersuchungen sowie 

durch technologisch bedingte 

Änderungen der Planungen des Abbaus, 

einen Total abbruch der RFR- 

Restanlage unter Strahlenschutzbedingungen 

vorzunehmen, wobei 

hierfür die notwen digen Erläuterungsberichte 

benötigt wurden wie ebenfalls 

für die Baufreiheit die Erweiterung des 

räumlichen Geltungsbereiches. 

Zur jeweiligen Genehmigungsplanung 

erfolgte die grundlegende 

Beschreibung des Gesamtvorhabens 

durch Erläuterungsberichte. Im 

Rahmen der Vierten Genehmigung 

gab es beispielsweise 18 Vorhaben, 

wobei das Vor haben 11 und das Vorhaben 

15 jeweils in drei Teile 

gegliedert war. Es wurden folglich 

22 Erläuterungs berichte erstellt, von 

denen sieben bereits bei Erteilung der 

Genehmigung von der zuständigen 

Behörde bestätigt wurden. Die weiteren 

Erläuterungsberichte erhielten im 

Zuge des Aufsichtsverfahrens die 

Zustimmung. Nach Abschluss eines 

Vorhabens wurde ein Abschlussbericht 

erstellt und an die zuständige 

Behörde übergeben. 

4 Ausgangssituation 

(radiologisch, 

konventionell) 

Die radiologische Ausgangssituation 

war geprägt durch das Vorhandensein 

von be- und unbestrahltem Kernbrennstoff, 

Kernmaterial sowie von 

Aktivierung und Kontamination 

unterschiedlichster Stoffe, die lokal 

und aktivitäts mäßig sehr differenziert 

waren. Einen Grobüberblick zum 

Aktivitätsinventar gibt die Abb. 6. 

Die Herstellung der Kernbrennstoff- 

Freiheit der RFR- Anlage war ein 

wichtiger Schritt für den Start des 

Rückbaus. Dazu mussten sowohl die 

951 bestrahlten Brenn elemente mit 

einer Gesamtaktivität von 8,91 E+15 

Bq als auch einige Posten kernbrennstoffhaltige 

Abfälle aus der RFR- 

Anlage entfernt werden. Diese Vorhaben 

wurden im Februar 2001 abgeschlossen. 

Weitere Einzelheiten sind 

im Abschnitt 7.2 zu entnehmen (s. atw 

1/2020). 

Hinsichtlich der Aktivierung waren 

alle reaktornahen Baustrukturen 

und Anlagenteile, wie z. B. die Inneneinbauten 

des Reaktors, die Bestrahlungskanäle, 

das Biologische Schild, 

die Thermische Säule zu beachten. 

Wichtige Aktivierungsnuklide waren 

beispielsweise Eu-152, Ba-133, H-3, 

Co-60 (Beton/Bauschutt), Fe-55, 

Co-60, Ni-63, H-3 (Eisenteile), Fe-55, 

Co-60, Ni-63, H-3 (Stahlguss-Reaktordeckel), 

H-3, Ni-63, Co-60, Be-10 




(Be-Reflektorelemente), Ni-63, Ni-59, 

H-3, C-14, Co-60 (Borcarbid- 

Absorberstäbe), C-14, H-3, Eu-152, 

Co-60 (Graphit). Unerwartet wurde 

bei der radiologischen Voruntersuchung 

festgestellt, dass anstatt 

des in den auunterlagen ausgewiesenen 

Barytbetons das Biologische 

Schild aus Beton mit Eisenteilen 

bestand. 

Kontaminationen resultierten vor 

allem aus flüssigkeitsgetragenem 

Transport nach Freisetzung sowie aus 

einem Am-241-Ereignis (Freisetzung 

aus einem Target, 1969). Die Hauptnuklide 

in der RFR-Anlage waren 

H-3, Co-60, Sr-90+, Cs-137, Eu-152, 

U- und Pu-Nuklide und Am-241. Zur 

Erfassung der jeweiligen Ausgangssituation 

wurde nach der historischen 

Erkundung die radiologische Voruntersuchung 

nach Möglichkeit mit 

konventionellem Schadstoffuntersuchungen 

gekoppelt. Dies reduzierte 

später den zeitlichen und finanziellen 

Aufwand und so konnten z. T. potentielle 

Verdachtsflächen besser beurteilt 

und die Entsorgung schneller 

voran gebracht werden. 

Das Auftreten von chemotoxischen 

(konventionellen) Schadstoffen resultiert 

sowohl aus Betriebsabläufen 

als auch im Wesentlichen aus schädlichen 

Bestandteilen der Baustoffe. 

Zu den Gebäudeschadstoffen zählen 

beispielsweise Polyzyklische Aromatische 

Kohlenwasserstoffe (PAK, 

insbesondere in Teeranstrichen), 

Polychlorierte Biphenyle (PCB, Oberflächenbeschichtungen), 

künstliche 

Mineralfasern (KMF) und Asbest (in 

Dämmstoffen). In Ausrüstungen 

und Anlagen fand man z. B. PCB 

(Transformatoren, Kondensatoren), 

Mineral kohlenwasserstoffe (MKW, 

Pumpen) oder Asbest (Dichtungen). 

Aus Betriebsabläufen stammen u. a. 

Kontaminationen von Schwermetallen 

und MKW. Das frühzeitige 

Erkennen der Schadstoffsituation 

ermöglichte die planerische Einarbeitung 

in die Rückbauprozesse 

sowie vor allem einen zeitlichen Vorlauf 

für eine sachgerechte Entsorgung 

des betreffenden Materials zu erhalten. 

Zu der Schadstoffproblematik 

veröffentlichte der VKTA einige 

Beiträge [10–12]. Bezüglich des RFR- 

Rückbaus sind insbesondere festgestellte 

PAK-Kontaminationen im Beton 

und Erdreich, das Vorfinden von 

Asbest hinter den Stahlbecken im 

Lager becken AB 2 (nicht in Bauzeichnung 

erwähnt) sowie lokale Kontaminationen 

mit MKW, Quecksilber 

und Schwermetallen im Beton hervorzuheben. 

5 Planungskonzept 

Grundsatz für die Planung war zum 

einen das Ziel, bei Stilllegung und 

Rückbau die Entsorgung von freigegebenen 

Stoffen parallel vorzunehmen 

und die radio aktiven Abfälle im 

1999 errichteten Zwischenlager 

Rossendorf, welches 2000 erweitert 

wurde, ordnungs gemäß für die spätere 

Endlagerung zu lagern. Zum 

anderen sollte das Betriebspersonal 

des RFR so weit wie möglich eingebunden 

werden und der VKTA nicht 

nur die Planungshoheit innehaben, 

sondern auch möglichst viele kostengünstige 

Beiträge zur Aufgabenerfüllung 

leisten. Da der VKTA planungstechnisch 

nicht alles abdecken konnte, 

wurde ab 2004 mit der heutigen 

Siempelkamp NIS Ingenieurgesellschaft 

mbH (NIS) eine Fachfirma eingebunden. 

NIS war, unterstützt durch 

Fachplaner in den Gebieten Lüftung, 

Tragwerksplanung und Schadstofferkundung, 

mit der Planung, der 

Erstellung und Bewertung von Ausschreibungsunterlagen 

sowie der 

Bau überwachung beauftragt. Ebenso 

wurden für spezielle Aufgaben weitere 

Fachfirmen gebunden. Der VKTA 

selbst, zeitweilig unterstützt durch 

einen Projektsteuerer, stellte die 

Rückbauleitung, bestehend aus: 

p Rückbauleiter 

p Gebäudeverantwortlicher, 

Strahlen schutzbeauftragter 

p Strahlenschutzingenieur/in 

p Strahlenschutzfachkraft 

p technischem Personal für strahlenschutztechnische 

Messungen, 

Transporte und kleineren technischen 

Aufgaben 

Die personelle Absicherung der Rückbauleitung 

in der Rückbau-Etappe bis 

2007 wurde dabei ausschließlich 

durch das ehemalige Betriebspersonal 

des RFR sicher gestellt. Die ab 2006 

auszuführenden Arbeiten wurden 

strukturiert, geplant, ausgeschrieben 

und an externe Dienstleister im 

Rahmen von Verträgen nach VOL 

oder VOB vergeben. Hierzu wurde 

eine Struktur verwendet, die sich 

einheitlich in Arbeits-, Termin- und 

Finanzpläne gliederte mit den 

Baulosgruppen: 

p Vorbereitende Maßnahmen 

p Bereitstellung von Ausrüstungen 

p Dienstleistungen Abbau 

p Abbaubegleitende lufttechnische 

Maßnahmen 

p Dienstleistungen Dekontamination 

p Arbeitsbegleitender Strahlenschutz 

p Mess- und Freimessprogramme in 

Planung und Durchführung 

p Abbaubegleitende sonstige 

Maßnahmen 

p Abbrucharbeiten bis zur Geländeprofilierung 

Die Leistungen des arbeitsbegleitenden 

Strahlenschutzes, der Planung 

und Durchführung von radiolo gischen 

und konventionellen Messungen/Analysen 

und von (Frei)Mess programmen 

übernahm der VKTA weitestgehend 

selbst. Dazu nutzte der VKTA u. a. sein 

nach DIN EN ISO/IEC 17025 akkreditiertes 

Labor für Umwelt- und Radionuklidanalytik. 

Andere auszuführende 

Arbeiten wurden in rund 200 Losen 

i. d. R. ausgeschrieben und den Losgruppen 

zugeordnet. Die bewusst 

kleinteilige Vergabe von Losen erlaubte 

es dem VKTA, mit seinen Planern 

auf besondere und erst im Verlauf des 

Abbaus erkennbare neue Situationen 

zügig zu reagieren, ortsnahe Firmen 

einzubinden und letztlich Kosten zu 

sparen. 

Der Arbeitsumfang des Gesamtprojektes 

entsprechend der jeweiligen 

Genehmigung wurde, wie bereits 

erwähnt, in Einzelvorhaben gegliedert, 

die in Erläuterungsberichten 

der Genehmigungsbehörde dargelegt 

wurden. Die zeit liche Abarbeitung der 

einzelnen Vorhaben erfolgte entsprechend 

eines Rahmenplanes, der 

mehrmals den objektiven Umständen 

angepasst werden musste. Für jedes 

Einzelvorhaben lag ein Detailplan vor. 

Die Durch führung der Arbeiten erfolgte 

auf der Grundlage von Arbeitsanweisungen 

und Ablaufplänen, 

wobei u. a. vor jedem Vorhaben ein 

Rückbauerlaubnis- sowie ein Arbeitserlaubnisverfahren 

gemäß der entsprechenden 

VKTA- Regelung durchzuführen 

war. 

6 Meilensteine Stilllegung 

und Rückbau 

Nach dem Erhalt der jeweiligen 

Genehmigung konnten die geplanten 

Arbeiten durchgeführt werden, wobei 

nach folgend wichtige Meilensteine 

des Rückbaufortschrittes beim RFR 

chronologisch (Durchführungszeitraum 

in Klammern) aufgeführt 

werden: 

p Kabinettsbeschluss des Freistaates 

Sachsen zur „Endgültige Stilllegung“ 

(13. Juli 1993) 

p Antrag auf Stilllegung 

(21. Dezember 1994), Erhalt der 

Ersten Genehmigung (30. Januar 

1998, vgl. Abschnitt 3) und somit 

ab 2. Februar 1998 RFR-Betriebsführung 

nicht mehr auf der 

Grund lage der Aufsichtlichen 

Anord nungen des SMU, u. a. 

vom 28. Juni 1991 

p Herstellen der Kernbrennstofffreiheit 

des Reaktor behälters 






durch Umlagerung der Brennelemente 

aus dem Reaktorkern ins 

Lagerbecken AB 2 und dem Ausbau 

der kernbrennstoffhaltigen 

Neutro nen detektoren (6. April 

1998) 

p Antrag auf Zweite Stilllegungsgenehmigung 

(10. Oktober 1997); 

Erhalt der Zweiten Genehmigung 

(30. Oktober 1998) mit 

1. Änderung (11. Februar 1999) 

p Rückbau des 2. Kühlkreislaufes 

mit Entkernung und Abriss des 

Armaturenhauses, der Trockenkühltürme, 

des Pumpenhauses 

sowie der Verbindungsleitungen 

zwischen den Gebäuden und dem 

Reaktorgebäude; mit SMUL- 

Schreiben vom 16. August 1999 

waren diese Anlagen nicht mehr 

Bestandteil der RFR-Anlage 

p Überführung der Brennelemente 

aus dem Lagerbecken AB 2 in 

CASTOR® MTR 2-Behälter, Endabfertigung 

mit anschließendem 

Transport in die Transportbereitstellunghalle 

des VKTA 

(24. November 2000) 

p RFR-Anlage kernbrennstofffrei 

nach Abgabe/Konditionierung der 

restlichen Posten Kernbrennstoffabfälle 

(26. Februar 2001) 

p Antrag auf Dritte Stilllegungsgenehmigung 

(29. Dezember 

1998); Erhalt der Dritten 

Genehmigung (3. April 2001) 

p Ausbau des Reaktorbehälters, 

Arbeiten zur Transportbereitstellung 

und Transport zum 

Konditionierer (2001/2002) 

p Antrag auf Vierte Stilllegungsgenehmigung 

(31. Juli 2003); 

Erhalt der Vierten Genehmigung 

(1. Februar 2005) mit 2. Änderungsgenehmigung 

(9. Januar 

2014) 

p Transport der RFR-Brennelemente 

mittels CASTOREN ins Zwischenlager 

Ahaus (3 Konvoitransporte 

vom 30. Mai 2005 bis 13. Juni 

2005) 

p Abbau des RFR-Baukörpers einschließlich 

der Aus kleidung des 

Lagerbeckens AB 1 (2008/2009) 

p Ausräumen, Dekontamination und 

Abbruch der Heißen Zellen 

(2009/2010) 

p Ausbau des Deaerators 

(2010/2011) 

p Ausräumen, Dekontamination und 

Entkernung des Kellergeschosses 

(2011/2012) 

p Abbau, Dekontamination und 

Entsorgung der Teile des 

frei gegebenen Fortluftschornsteines 

(2013/2014) 

p Dekontamination, Entkernung 

und Abriss des Filter- und 

Ven tilationshauses (2014/2015) 

p Dekontamination, Entkernung 

und Abriss des Labortraktes mit 

Reaktorhalle und -warte (2016) 

p Abschluss des Ausbaus aller 

tiefliegender Bau strukturen und 

Bodensanierung (2016) 

p Abschluss der Baugrubenverfüllung 

sowie Beendigung der 

restlichen bautechnischen 

Maß nahmen (Profi lierung, 

Oberflächengestaltung 

(2017/2018) 

Autoren 

Reinhard Knappik 

Klaus Geyer 

Sven Jansen 

Cornelia Graetz 

VKTA Rossendorf 

Bautzner Landstraße 400 

01328 Dresden 

Deutschland 

Imprint 

| Editorial Advisory Board 

Frank Apel 

Erik Baumann 

Dr. Erwin Fischer 

Carsten George 

Eckehard Göring 

Dr. Florian Gremme 

Dr. Ralf Güldner 

Carsten Haferkamp 

Christian Jurianz 

Dr. Anton Kastenmüller 

Prof. Dr. Marco K. Koch 

Ulf Kutscher 

Herbert Lenz 

Jan-Christan Lewitz 

Andreas Loeb 

Dr. Thomas Mull 

Dr. Joachim Ohnemus 

Olaf Oldiges 

Dr. Tatiana Salnikova 

Dr. Andreas Schaffrath 

Dr. Jens Schröder 

Norbert Schröder 

Prof. Dr. Jörg Starflinger 

Dr. Brigitte Trolldenier 

Dr. Walter Tromm 

Dr. Hans-Georg Willschütz 

Dr. Hannes Wimmer 

| Editorial Office 

Christopher Weßelmann (Editor in Chief) 

Im Tal 121, 45529 Hattingen, Germany 

Phone: +49 2324 4397723 

Fax: +49 2324 4397724 

E-mail: editorial@nucmag.com 

Nicole Koch (Editor) 

c/o INFORUM, Berlin, Germany 

Phone: +49 176 84184604 

E-mail: nicole.koch@nucmag.com 

| Official Journal of Kerntechnische Gesellschaft e. V. (KTG) 

| Publisher 

INFORUM Verlags- und Verwaltungsgesellschaft mbH 

Robert-Koch-Platz 4, 10115 Berlin, Germany 

Phone: +49 30 498555-30 

Fax: +49 30 498555-18 

www.nucmag.com 

| General Manager 

Dr. Thomas Behringer 

| CvD 

Marika Didonaki 

| Advertising and Subscription 

Petra Dinter-Tumtzak 

Phone: +49 30 498555-30 

Fax: +49 30 498555-18 

E-mail: petra.dinter@nucmag.com 

| Layout 

zi.zero Kommunikation 

Antje Zimmermann 

Berlin, Germany 

| Printing 

inpuncto:asmuth 

druck + medien gmbh 

Baunscheidtstraße 11, 53113 Bonn, Germany 

| Price List for Advertisement 

Valid as of 1 January 2019 

Published monthly, 9 issues per year 

Germany: 

Per issue/copy (incl. VAT, excl. postage) 24.- € 

Annual subscription (incl. VAT and postage) 187.- € 

All EU member states without VAT number: 

Per issue/copy (incl. VAT, excl. postage) 24.- € 

Annual subscription (incl. VAT, excl. postage) 187.- € 

EU member states with VAT number 

and all other countries: 

Per issues/copy (no VAT, excl. postage) 22.43 € 

Annual subscription (no VAT, excl. postage) 174.77 € 

| Copyright 

The journal and all papers and photos contained in it are protected by 

copyright. Any use made thereof outside the Copyright Act without the 

consent of the publisher, INFORUM Verlags- und Verwaltungsgesellschaft 

mbH, is prohibited. This applies to reproductions, translations, 

micro filming and the input and incorporation into electronic systems. 

The individual author is held responsible for the contents of the 

respective paper. Please address letters and manuscripts only to the 

Editorial Staff and not to individual persons of the association´s staff. 

We do not assume any responsibility for unrequested contributions. 

Signed articles do not necessarily represent the views of the editorial. 

ISSN 1431-5254 




First On-site Demonstration of Laser- based 

Decontamination Technology in Germany 

Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann 

and Antonio Hurtado 

1 Introduction Several ongoing dismantling and decommissioning (D&D) projects of nuclear facilities in 

Germany produce wide experiences in terms of the regulation framework and the practical implementation. The D&D 

of commercial and research reactors require a comprehensive adaption of the available technologies to meet the 

demands of the associated health and safety requirements. This causes a rising monetary and human effort. 

To face this challenge the German 

federal government focuses the 

research and development activities 

to: 

a- protect human beings and the 

environment during nuclear 

decommissioning, 

b- educate staff and young academics 

and 

c- develop and optimize decommission 

methods. 

The dismantling and decommissioning 

of the various nuclear facilities 

located at Karlsruhe Institute of 

Technology (KIT) Campus North, the 

Kerntechnische Entsorgung Karlsruhe 

GmbH (KTE) faces several technical 

challenges. 

On the one hand, metallic surfaces, 

e. g. in the Hot Cells Facility (HZ), 

must be decontaminated radiologically, 

on the other hand, paint layers 

as well as radiological contamination 

must be removed from metallic components 

during post treatment, e. g. 

in the Waste Treatment Department 

(EB). 

Furthermore, large surfaces of 

concrete structures, e. g. in the 

Reprocessing Building (PG) of the 

Karlsruhe Reprocessing Plant (WAK), 

in the Multi-Purpose Research Reactor 

(MZFR) and in the Research Reactor 

(FR-2) must be cleared from paint 

layers, containing sometimes high rate 

of polychlorinated biphenyls (PCB). 

The overriding task in these 

processes is to enclose the radioactive 

substances and limit the radiation 

exposure of personnel and environment. 

Furthermore, the volume of 

waste should be minimized. The work 

situation for the operating personnel 

must be constantly monitored and 

the workload should be reduced 

wherever possible. 

Sometimes, the currently available 

chemical and mechanical processes, 

reach their limit to comply with the 

objects mentioned before. 

Abrasive blasting of metallic structures 

results in the production of 

secondary waste, which has to be 

further processed in case of PCB. In 

chemical decontamination processes, 

pickling acids also must be specially 

treated or stored. 

Furthermore, all mechanical processes, 

such as shaving or sandblasting, 

also cause high workloads 

for the operating personnel due to 

vibrations, restoring forces and noise. 

Chemical processes, however, can 

cause damage through skin contact 

and inhalation. 

Decontamination by laser beam 

represents an interesting alternative. 

This technology can expand the 

repertory of currently available decontamination 

tech nologies. Being a 

contact-free procedure it comes along 

with the advantage of waste minimization 

and reduces restoring force for 

hand-held as well as semi-automatic 

application, following NEA recommendations 

[1]. 

At TU Dresden laser-based decontamination 

tech nology has been developed 

during the last years. The 

research project LaPLUS has been 

aimed at the optimi zation of the 

chemical-toxic decontamination of 

concrete surfaces and the technology 

transfer for the radiologic decontamination 

of metal surfaces. For that 

purpose, special hand-held laser tools 

for the use in nuclear sites were 

designed and tested on a laboratory 

scale. An essential part of the project 

was the technology transfer from 

laboratory scale to prototype status. 

For that purpose, the laser-based zi. 

decontamination was demonstrated 

in a realistic environment at the 

Multi-Purpose Research Reactor 

( MZFR). 

Under project leadership of TU 

Dresden, TU Berg akademie Freiberg 

(Development of process analysis 

tools) and IABG mbH (Design of laser 

tools) participated in the project. The 

KTE accompanied the project as an 

associated partner and supported the 

technology transfer to the MZFR. 

This paper presents the major 

results achieved inside the project. It 

covers the laser-based decontamination 

of PCB containing coatings on 

concrete and that of radiologic contaminated 

metal surfaces, the design 

and the test of special laser tools 

developed for the application in 

nuclear facilities. 

2 Application of laser 

systems for decontamination 

When a laser beam hits a surface, 

absorption of the energy leads to rapid 

heating of the substrate. The resulting 

spatial and temporal temperature 

distribution depends on optical and 

thermal characteristics of the substrate 

as well as on the parameters of 

the laser process [2]. Heating of the 

surfaces can result in melting, evaporation 

or sublimation. All of those 

mechanisms can be used for the 

removal of unwanted species, as 

necessary for surface decontamination. 

Additionally different 

practical requirements defined by the 

decontamination task require a 

careful selection of a suitable laser 

system. 

Rapid heating of the treated 

substrate surface above evaporation 

temperature can be achieved using 

pulsed lasers. The pulse duration is directly 

proportional to the volume of 

the heat affected zone [3], which 

means that shorter pulses lead to 

smaller heat affected volume. The 

possibilities and requirements of 

radiologic decontami nation using 

short pulsed lasers is treated in 

section 3. 

A different result can be achieved 

applying a con tinuous working (cw) 

laser beam, which continuously heats 

the surface. In case of interaction with 

paint, e.g. epoxy based coatings, the 

volatile components of that coating 

will evaporate and incinerate after 

reaching the ignition temperature. A 

subsequent combustion process of the 



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



coating is established. Chemical-toxic 

decontamination of PCB-containing 

paint using a continuous wave laser is 

treated in section 4. Application of 

high power diode lasers allows rapid 

treatment of surfaces with economically 

desirable laser equipment. 

Apart from beam parameters, 

further practical requirements have to 

be met for successful application of 

lasers in nuclear facilities. 

High mobility and toughness is 

required from the beam shaping optics, 

as opposed to regular industrial 

application. This demand is coped 

with using simplified 1D scanning 

optics. 

For the cost-efficient use of lasers, 

the application of optical fibres for 

beam transport is mandatory. This 

allows to arrange the laser-system 

outside the contaminated area, thus 

eliminating the risk of contamination 

for the laser- system. Only part of the 

fibre and the beam-shaping laser 

optics will be placed inside the control 

area and can still be reused on several 

decommissioning sites, as explained 

in section 6. 

3 Decontamination 

of metal surfaces 

Laser-based cleaning of metal surfaces 

has been established within the last 

decade in applications like pre-treatment 

for welding, restauration of art 

and decoating of paint. In these cases, 

a laser beam is scanned and moved 

over the soiled surfaces and removes 

adhering unwanted species. High 

process selectivity can be achieved 

on metal surfaces, because a large 

fraction of the beam is reflected on 

blank metals as opposed to higher 

absorption on organics and oxides. 

Laser-based cleaning substitutes 

chemical as well as mechanic abrasive 

processes and results in a reduction of 

waste of factor 2.6 J/cm² for 

austenitic steel, >3.5 J/cm² for ferritic 

steel and >5.3 J/cm² for zinc plated 

Wavelength 

Average Power 

Pulse-Energy 

Peak Power 

Pulse length 

Pulse frequency 

Scanner 

1064 nm 

150 W 

11.5 mJ @ 12 kHz 

112 kW @ 12 kHz 

102 ns@ 12kHz 

12 - 40 kHz 

Spot diameter 472 µm 

Scanning width 

Mass of optics 

60 mm 

1,7 kg 

| Tab. 1. 

Specifications of Nd:YAG Laser CL150. 

| Fig. 1. 

Idealized scenarios of contamination, blue color symbolizes contamination: 

contamination of uncovered surfaces (A), 

on top of covered surface (B) and 

on as well as under covering layer (C), 

contamination within oxide layer (D). 

| Fig. 2. 

Ablation depth as function of cumulative energy with a pulse duration 

of 105 ns. 

| Fig. 3. 

Concentration of surrogates on sample surface prior and after 

decontamination. 




steel. Tested paint layers can be 

ablated at fluence above 1.8 J/cm², 

while oxide layers can be removed at 

even lower fluence. 

The achieved ablation depth for 

the above mentioned materials is 

plotted in Figure 2. 

It can be seen that the removal 

speed is about 100 times higher for 

paint than for metals, what supports 

the selective removal of the paint on 

the metal ground. A variation of laser 

fluences between 1.8 – 3.5 J/cm² has 

been tested for its decontamination 

effect on blank austenitic steel. Single 

scans resulted in a contamination 

reduction of 50 – 90 % for all surrogates. 

Similar tests have been conducted 

on painted surface applying a 

laser fluence of 4.4 J/cm², to prove 

complete removal of the covering 

layer (Figure 3). 

Both tested cases result in a 

decreased contaminant concentration 

and decontamination factors up to 

98.9 have been achieved. 

4 Laser-based PCB 

degradation 

The TU Dresden has developed a PCB 

decontamination process by utilising 

a continuous wave diode laser. 

Laboratory experiments on concrete 

surfaces coated with epoxy paint 

demonstrated a reduction of 96.83 % 

of the PCB value. The PCB decomposition 

rate on the surface and in 

the exhaust gas before filtering is 

88.75 % [16]. 

PCB decomposes at temperatures 

above 800 °C. At optimized process 

parameters the temperature in the 

laser spot is much higher [17], enable 

the full decomposition of the PCB. 

Rapid cooling of the exhaust gas is 

required to prevent the formation of 

toxic polychlorinated dioxins and 

furans (PCDD/PCDF) which can be 

ensured by the laser processing. The 

application of a fabric filter in the 

extraction and filtration unit limits 

the temperature of filtered exhaust 

gas to



| Fig. 6. 

Overview of tested samples as supplied by KTE (A– lid of a 200 liter drum, B– steel sheet, C– coated u 

profile, D– locker door part). 

| Fig. 7. 

Components for decontamination of complex metal surfaces, (A) laser tool for the decoating of flat 

surfaces and (B) laser tool for the decontamination of edges. 

transported using optical fibres to 

reach the experimental area. This 

feature allows for the protection of 

costly laser equipment from contamination 

in any case, as described in 

section 2. The user can simply equip 

the required laser tool and start the 

decontamination process. Suction 

and filtration of mobilized contamination 

at high efficiency is facilitated 

applying a commercial filtration and 

suction unit by ULT GmbH. The 

localized suction of aerosols leads to 

prevention of recontamination and 

guarantees clear vision for the user at 

all times. 

Restricted access to the experimental 

area is beneficial to simplify the 

laser safety requirements, even in case 

of a technology demonstration. 

A full-fledged risk assessment 

according to §5 of ArbSchG [18] and 

§3 of ArbStättV [19] has been filed to 

scan for any arising risks for operators 

coming along with the application of 

the whole set up. All requirements of 

the occupational safety have been 

tested and proven. The risk assessment 

covers the following subject 

area: 

p General occupational safety 

p Laser safety 

p Safety from toxic substances (PCB) 

p Radiation safety 

p Respiratory protection 

The risk assessment complies with the 

laws, guidelines (ASR), workers compensation 

board rules and provisions. 

The risk assessment was finalized by 

TU Dresden and IABG and checked by 

KTE and has been the fundamental 

milestone for approval of the on-site 

demonstration at MZFR. 

5.3 Decontamination of 

complex metal geometries 

The decontamination of metal surfaces 

at MZFR was conducted on 

paints and metals prior unknown, to 

check the ability to transfer results 

from laboratory scale to on-site tests. 

To test limits of the laser tool, samples 

of varying geometries have been 

supplied by KTE, e.g. painted steel 

sheets, bracket steel or complex surfaces 

like ridges/spline/serration, as 

shown in Figure 6. 

Different laser tools were designed 

to ensure the decontamination at 

different surface geometries, e.g. in 

case of tighter angles the tool was 

adapted to maintain the complete 

suction of particles and aerosols 

during the laser process. A tool change 

can be completed within 1.5 min, 

Figure 7. 

Similar ablation characteristics 

were found for all tested samples as 

compared to the laboratory results 

(Figure 2). This implies that generalized 

material characteristics for paint 

layers and metals can be applied. 

From practical point of view, the 

flexibility of the laser ablation process 

was verified, as all tested geometries 

were completely cleaned without 

residues (Figure 8). 

5.4 Decontamination of 

concrete walls and floor 

at MZFR 

Decontamination of concrete demands 

solutions for a multitude of 

demands, e.g. uneven surfaces, varying 

thicknesses of paint, different 

shapes of surfaces and unknown PCB 

concentration in the decontamination 

paint. The practicability of the 

developed laser system was tested on 

walls and floors at the MZFR on 

sample surfaces sized 30 x 30 cm². 

The decontamination paint on the 

wall was thin and exhibited a high 

PCB concen tration (see Table 3). The 

floor featured lower PCB concentration 

and very thick paint (approx. 

1.5 mm). 

In Figure 9 the side view of the 

laser head is shown during the decontamination. 

During the process all 

arising by-products, like visible particles 

and gases, are extracted by the 

extraction and filtration unit and do 

not endanger the working staff. 

Figure 10 displays the result 

of decontamination on a floor 

| Fig. 8. 

Steel sheet partly decoated. 

| Fig. 9. 

Laser head during the decontamination of the floor. 




| Fig. 10. 

Demonstration field floor (A) before and (B) after the decontamination 

with multiple scans. 

Sample number Position Before decontamination 

PCB in mg/kg 

demonstration field before (A) and 

after (B) the laser process. 

In Figure 11 the laser head is 

displayed with different adaptions 

for flat areas (picture A and B) and 

internal corners (picture C). The 

modular set-up enables a fast change 

of the adaptions omitting any changes 

to the leaser heads basic components. 

Using such adaptions the decontamination 

of a wall with adjoining 

corners was demonstrated, illustrating 

the clear advantage compared 

to shaving equipment, where usually 

boundary areas need an additional, 

time- consuming handling step. 

The removal of thick layers and/or 

remaining soot can be realized by 

multiple passes. Both have been 

demonstrated in Karlsruhe. 

During the demonstration an 

average laser power of 3.7 kW was 

applied, resulting in a practical 

determined ablation speed of 6 m²/h 

5.5 PCB sampling at the MZFR 

Within the demonstration PCB 

sampling was performed to prove the 

decontamination rate of the laserbased 

paint removal. This study aimed 

at the evaluation of the PCB concentration 

of the primary source. To 

release material the PCB-concentration 

must amount less than 50 mg/ 

kg. Samples with mass > 50 mg were 

extracted using an electric scraper. 

AGROLAB Labor GmbH performed 

the analysis of the samples according 

to DIN EN 15308. The samples taken 

before the decontamination consisted 

mainly of decontamination paint. 

After the decontamination no paint 

remained on the surface anymore, 

therefore the now surfacing concrete 

was sampled. 

In Figure 12 the sampled surfaces 

are shown before PCB sampling (A), 

after PCB sampling (B) and after 

decontamination and the consecutive 

PCB sampling (C). The corresponding 

PCB concentration is shown in 

Table 3. 

The laser-based decontamination 

process resulted in an average reduction 

of PCB of 98.7 % within a practical 

application. This value is even 

higher than the value of the laboratory 

experiments performed at the TU 

Dresden [16]. The remaining PCB 

concentration after the decontamination 

is mainly caused by the PCB 

inside the concrete matrix. After sealing 

the walls with PCB containing 

paint the PCB diffuses from the decontamination 

paint (primary source) 

into the concrete. Alternative laserbased 

technologies are available to 

remove and to vitrify concrete surfaces 

in a single process step [20; 21]. 

| Fig. 11. 

Laser head for concrete decontamination; 

A- laser head (round version), B- laser head (angled), C- laser head internal corner. 

After decontamination 

PCB in mg/kg 

Difference 

mg/kg 

| Fig. 12. 

Examples of a demonstration field: 

(A)- before sampling, (B)- after decontamination and (C)- after decontamination and sampling. 

6 Conclusion and future 

prospects 

The TU Dresden successful demonstrated 

the application of laser-based 

decontamination on radiologic and 

chemical-toxic contaminated metal 

and concrete surfaces at the MZFR in 

Karlsruhe. All necessary documents to 

obtain acceptance of the KTE were 

prepared by TU Dresden and approved 

by KTE. An independent work 

permit was available to TU Dresden. 

The selective decontamination on 

metal surfaces is characterized by a 

process stop after removing any coating 

(decontamination paint, oxide 

and/or contamination). The practical 

removal rate on metal is 0.42 m²/h 

using a mean laser power of 150 W 

and an average paint thickness of 

165 µm. The application of laser-based 

decontamination to PCB-contaminated 

concrete surfaces resulted in 

a mean reduction of 98.7 % of the 

PCB-concentration. During the onsite-demonstration 

a removal rate of 

6 m²/h of paint from concrete walls 

with 3.7 kW laser power were accomplished. 

The handheld laser tools 

Reduction 

% 

1 Floor 9.3 0.23 9.07 97.5 

2 Floor 30.1 0.04 30.06 99.9 

3 Wall 3,810 67.3 3,742.7 98.2 

4 Wall 2,360 19 2,341 99.2 

5 Wall 1,976 18.9 1,957.1 99.0 

6 Wall 2,510 14.2 2,495.8 99.4 

7 Wall 2,272 49.7 2,222.3 97.8 

| Tab. 3. 

PCB concentration of the sampling before and after the decontamination. 

Ø 98.7 % 






enable high flexibility and reduce 

restraining forces for the working 

staff. The dry laser cleaning process 

runs completely without abrasive or 

chemical aids thus reducing the 

amount of secondary waste drastically. 

The determined removal rates on 

concrete and metal are highly affected 

by the thickness and optical pro perties 

of the decontamination paint. The 

great advantage of lasers is the scalability 

of their power. The use of 

commercial available lasers with 

higher power would result in higher 

ablation rates without any changes to 

the applied laser tools. 

The on-site demonstration at 

MZFR Karlsruhe verified the possibilities 

of that technology for decontamination 

of contaminated surfaces. 

Robust laser tools can offer an 

alternative for surface decontamination 

in the near future. Following the 

positive results of the research project 

LaPLUS more detailed insight into the 

behavior of mobilized aerosols as well 

as further tests within control areas 

will be provided from follow up projects. 

Industrial application of the 

laser-based decontamination process 

should become feasible in the next 

years 

Acknowledgement 

The on-site demonstration at the 

MZFR in Karlsruhe was carried out 

inside the Research and Development 

Project LaPLUS financed by The 

German Federal Ministry of Education 

and Research (BMBF) under the 

contract number 15S9215A. 

We would like to thank all colleagues 

of the MZFR that supported 

the on-site-demonstration with enthusiasm 

and useful evidences as well 

as practical drive. Our special thanks 

go to the members of our technical 

staff, who were a great support during 

the exhausting time of demonstration 

and the whole project as well. 

References 

[1] Nuclear Energy Agency: R&D and Innovation Needs for 

Decommissioning of Nuclear Facilities. 2014 (7191) 

[2] BLIEDTNER, M.: Lasermaterialbearbeitung. München: 

CARL HANSER Verlag GMBH, 2013 

[3] Leitz, K.-H.; Redlingshöfer, B.; Reg, Y.; Otto, A.; Schmidt, M.: 

Metal Ablation with Short and Ultrashort Laser Pulses. In: 

Physics Procedia 12 (2011), S. 230–238 

[4] Büchter, E.: Entwicklung eines Hochleistungs-Laserstrahl- 

Reinigungsgerätes zur Ressourcen schonenden 

Entschichtung von Oberflächen. Herzogenrath, 2004 

[5] Carvalho, L.; Pacquentin, W.; Tabarant, M.; Maskrot, H.; 

Semerok, A.: Growth of micrometric oxide layers to explore 

laser decontamination of metallic surfaces. In: EPJ Nuclear 

Sciences & Technologies 3 (2017), S. 30 

[6] Delaporte, Ph.; Gastaud, M.; Marine, W.; Sentis, M.; 

Uteza, O.; Thouvenot, P.; Alcaraz, J. L.; Le Samedy, J. M.; 

Blin, D.: Radioactive oxide removal by XeCl laser. 

In: Applied Surface Science 197-198 (2002), S. 826–830 

[7] Kim, D.; Lim, H.: Laser Decontamination of Carbon Steel 

Surfaces. In: ISIJ International (2003), Nr. 43, S. 1289–1291 

[8] Leontyev, A.; Semerok, A.; Farcage, D.; Thro, P.-Y.; Grisolia, C.; 

Widdowson, A.; Coad, P.; Rubel, M.: Theoretical and 

experimental studies on molybdenum and stainless steel 

mirrors cleaning by high repetition rate laser beam. 

In: Fusion Engineering and Design 86 (2011), 

9-11, S. 1728–1731 

[9] Edelson, M. C.; Pang, H.: A laser-based solution to industrial 

decontamination problems. 1995 (ICALEO 1995 768) 

[10] Potiens, A. J.; Dellamano, J. C.; Vicente, R.; Raele, M. P.; 

Wetter, N. U.; Landulfo, E.: Laser decontamination of the 

radioactive lightning rods. In: Radiation Physics and 

Chemistry 95 (2014), S. 188–190 

[11] Sadanori, S.; Seiji, A.; Inoue, T.: Applying laser technology to 

decommissioning for nuclear power plant (Advanced 

High-Power Lasers and Applications). Osaka, Japan, 1999 

[12] Takakuni, H.; Yutaka, K.; Masato, M.: Application of a laser to 

decontamination and decommissioning of nuclear facilities 

at JAERI (Advanced High-Power Lasers and Applications). 

Osaka, Japan, 1999 

[13] Vatry, A.; Grisolia, C.; Delaporte, Ph.; Sentis, M.: Removal 

of in vessel Tokamak dust by laser techniques. In: Fusion 

Engineering and Design 86 (2011), 9-11, S. 2717–2721 

[14] Nilaya, J. P.; Raote, P.; Kumar, A.; Biswas, D. J.: Laser-assisted 

decontamination – A wavelength dependent study. In: 

Applied Surface Science 254 (2008), Nr. 22, S. 7377–7380 

[15] Delaporte, Ph.; Gastaud, M.; Marine, W.; Sentis, M.; Uteza, O.; 

Thouvenot, P.; Alcaraz, J. L.; Le Samedy, J. M.; Blin, D.: Dry 

excimer laser cleaning applied to nuclear decontamination. 

In: Applied Surface Science 208-209 (2003), S. 298–305 

[16] Anthofer, A.; Kögler, P.; Friedrich, C.; Lippmann, W.; 

Hurtado, A.: Laser decontamination and decomposition 

of PCB-containing paint. In: Optics & Laser Technology 87 

(2017), S. 31–42 

[17] Anthofer, A.: Oberflächenentschichtung mittels 

Laserstrahlung. Untersuchungen zur Dekontamination 

radioaktiv und chemisch-toxisch belasteter Betonoberflächen 

mittels Lasertechnologie. Dissertation, Dresden, 2014 

[18] Bundesministerium der Justiz und für Verbraucherschutz: 

Gesetz über die Durchführung von Maßnahmen des 

Arbeitsschutzes zur Verbesserung der Sicherheit und des 

Gesundheitsschutzes der Beschäftigten bei der Arbeit 

(idF v. 31. 8. 2015) (1996-08-07) 

[19] Bundesministerium der Justiz und für Verbraucherschutz: 

Verordnung über Arbeitsstätten 

(idF v. 18. 10. 2017) (2004-08-12) 

[20] Lippmann, W.; Herrmann, M.; Pietsch, C.; Reinecke, A.; 

Hille, C.; Wolf, R.; Zeuner, A.: LASABA II : Dekontamination 

silikatischer Oberflächen in kerntechnischen Anlagen mittels 

Laserabtrag bei gleichzeitiger Abproduktkonditionierung. 

Abschlussbericht, Dresden, 2008 

[21] Hurtado, A.; Littwin, R.; Lippmann, W.: MANOLA: Manipulator 

gesteuerter Oberflächenabtrag durch Lasertechnologie. 

Abschlussbericht, Dresden, 2011 

Authors 

Torsten Kahl, 

Georg Greifzu, 

Marion Herrmann, 

Wolfgang Lippmann, 

Antonio Hurtado 

Technische Universität Dresden 

Chair of Hydrogen and 

Nuclear Energy 

Institute of Power Engineering 

George-Bähr-Str. 3b 

01062 Dresden 

Germany 

Carsten Friedrich, 

Christian Held 

Kerntechnische 

Entsorgung Karlsruhe 

Hermann-von-Helmholtz-Platz 1 

76344 Eggenstein-Leopoldshafen 

Germany 




Scientists and Professionals from all around the World in Karlsruhe: 

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

on Nuclear Reactors “Physics, Fuels and Systems“ 

Victor Hugo Sanchez-Espinoza 

The Institute of Neutron Physics and Reactor Technology (INR) of the Karlsruhe Institute of Technology (KIT) together 

with the Commissariat à l’Énergie Atomique (CEA) hosted this year the “Frédéric Joliot/Otto Hahn (FJOH) Summer 

School“ at the Akademiehotel in Karlsruhe from August 21 st to August 30 th 2019. The topic of this year’s school was 

“Innovative Reactors: Matching the Design to Future Deployment and Energy Needs”. 

549 

REPORT 

It was organized in six technical blocks 

and one seminar devoted to “Public 

Acceptance of Energy Technologies” 

given by S. Hirschberg from PSI 

Switzerland. The first introduc tory 

block was devoted to “Various Innovative 

Reactor Concepts for Various 

Missions” and consisted in two 

lectures. The first one entitle “The 

Deciding Factors in Opting for a 

Particular Reactor Technology” was 

given by J. Guidez (CEA, France) and 

the second one entitled “Promising 

Reactor Concepts for Multiple Applications” 

given by S. Monti (IAEA). 

The second block was focused on the 

topic “Near-term-deployment Powerto-grid 

Integrated LWR Technology” 

and it consisted of three lectures as 

follows. The first lecture deals with 

“From Specific Design Criteria to an 

Advanced Modular LWR Concept: Approach, 

Methods, Validation” given by 

F. Morin (CEA, France). The second 

lecture was entitled “The Challenges 

of Designing and Licensing a First-ofa-kind 

Reactor Prototype, even a Small 

One” given by D. Delmastro (CNEA, 

Argentina) and finally the third lecture 

was entitled “The Reliability and 

Safety Case of a Reactor Equipped 

with Passive Systems” presented by 

A. Schaffrath (GRS, Germany). The 

third block was dedicated to the topic 

“Multi-purpose Molten Salt Reactors” 

and it was organized as three 

lectures to the following issues. The 

first lecture was entitled “MSR Design 

Principles, Concepts, Modelling Approaches, 

and Methods” given by T. 

Abram (Univ. of Man chester, UK). The 

second lecture was devoted to “From 

the MSR Physics Principles to a Plant 

Layout” given by E. Merle (Grenoble 

INP & CNRS) and the third lecture 

entitled “Fuel Salts Chemistry and 

Materials Compati bility” given by 

V. Ignatev (KI, Russia). The fourth 

block was devoted to “Nuclear Technology 

for Space Propulsion and 

Manned Space Exploration” and it 

consisted of two lectures. The first one 

entitled “ Nuclear Rocket Propulsion: 

Background, Physics and Methods, 

Design and Tests” given by W. Emrich 

Jr. ( NASA, USA) and the second one 

entitled “The Challenge of Fueling a 

Nuclear Engine for Space Exploration” 

given by J. Witter (BWXT, USA). 

The fifth block was devoted to 

“Minimum-intervention Long-life 

Breed-and-burn Fast Reactors” and 

it consisted of two lectures. The first 

one entitled “Physics of Breed-andburn 

Reactors, Optimized Core and 

Fuel Design, Licensing Case” given by 

K. Weaver (INL, USA) and the second 

one entitled “Cladding and Structural 

Materials for Very Long In-core 

Residence Times“ given by Y. Decarlan 

(CEA, France). The sixth block was 

devoted to “Reactor Concepts for 

Process Heat and Power-to-gas 

Applications” and it consisted of two 

lectures. The first one entitled 

“ Designing a Small Reactor to Bring 

Power to Remote Areas or to Produce 

Process Heat” given by J. Kloosterman 

(TU Delft, Netherlands) and the 

second one entitled „Techno-economic 

Assessment of Hydrogen Production 

from Nuclear Energy“ given by 

J. Witter (BWXT, USA). In the frame of 

the technical visit, a guided tour was 

organized for the TrasnetBW GmbH 

in Wendlingen, which is one of the 

largest system Control Centre for 

Baden Württemberg with headquarter 

in Stuttgart. It operates the electricity 

transmission grid in the German state 

of Baden-Württemberg, control and 

monitor the energy flows through the 

grid, and perform the necessary 

maintenance and network planning 

and development activities. This year, 

44 participants of 18 countries (EU, 

Asia, Latin America, East Europe, 

Middle East, Africa, and USA) attended 

the FJOH Summer School. 

Recognized experts of 10 different 

countries from Asia, EU, South 

America, USA and East Europe from 

Academia, industry, research and 

TSOs gave high-level lectures on topics 

of their expertise. During the ten days, 

the participants had the opportunity 

to exhaustive discussions with the 

lecturers and other participants. Apart 

| The 25 th Edition of the Frédéric Joliot/Otto Hahn Summer School on Nuclear 

Reactors “Physics, Fuels and Systems“, participants and lecturers in Karlsruhe. 

from the technical issues, another goal 

of the FJOH Summer School is to 

intensify the networking among the 

participants of different continents 

and nationalities with the common objective 

of enhanced safety worldwide. 

A well-recognized tradition during 

the FJOH Summer School is an extensive 

and diverse program with social 

events to get familiar with the German 

culture and way of life as well as to 

foster the exchange among the participants. 

This year, the Sunday trip 

consisted in the visit of the Technical 

Museum Speyer and afterwards, a 

Canoe Tour at the Old Rhine. After the 

Museum visit, the participants have 

free time to get familiar with the 

downtown of Speyer and to visit the 

Romanesque Cathedral that houses 

the grave f most important kings and 

emperors e.g. it is the burial place for 

emperor Konrad II and his wife. 

The next Summer School will be 

hosted by CEA in Aix-en-Provence, 

from August 26 th to September 4 th , 

2020 and it will be devoted to “High- 

fidelity Modelling and Simulation 

of Nuclear Reactors: Turning a 

Promise into Reality”. 

Author 

Dr.-Ing. Victor Hugo Sanchez-Espinoza 

Head of Group 

“Reactor Physics and Dynamics” 

Project Leader 

“LWR Safety Methods and Codes ” 

Karlsruhe Institute of Technology (KIT) 

Institute for Neutron Physics and 

Reactor Technology (INR) 

Hermann-von-Helmholtz-Platz 1 

76344 Eggenstein-Leopoldshafen 

Report 

The 25th Edition of the Frédéric Joliot/Otto Hahn Summer School on Nuclear Reactors “Physics, Fuels and Systems“ ı Victor Hugo Sanchez-Espinoza



550 

SPECIAL TOPIC | A JOURNEY THROUGH 50 YEARS AMNT 

Am 7. und 8. Mai 

2019 begingen wir 

das 50. Jubiläum 

unserer Jahrestagung 

Kerntechnik. Zu 

diesem Anlass öffnen 

wir unser atw-Archiv 

für Sie und präsentieren 

Ihnen in jeder 

Ausgabe einen 

historischen Beitrag. 

Überarbeitete 

Fassung eines 

Vortrags gehalten 

am 14. Mai 2002 

auf der Jahrestagung 

Kerntechnik 2002, 

Stuttgart, 

14.-16.5.2002 

Schutz von Mensch und Umwelt – 

Nukleare Anwendungen außerhalb des 

Energiesektors 

P.P. De Regge, Wien 

1 Einleitung Nukleare Technologien liefern immer mehr bedeutende Beiträge zum Schutz des Menschen und 

seiner Umwelt sowie zur Verbesserung der Lebensstandards. Die Internationale Atomenergie Organisation (IAEO) ist 

durch die Vereinten Nationen (UN) per Statut beauftragt, solche nuklearen Technologien zu entwickeln, zu fördern 

und deren Einsatz zu unterstützen. Sie erreicht dies durch Veröffentlichung von Informationen, Ausbildung, Zulieferung 

von Ausrüstung, Dienstleistungen und Erstellung von Sicherheitsnormen. 

In diesem Beitrag wird diese Rolle der IAEO erläutert 

und es wird weiter der heutige Stand der nuklearen 

Anwendungen in folgenden Bereichen dargestellt: 

p Nuklearmedizin und Gesundheitspflege 

p Veterinärmedizin und Viehzucht 

p Bodenkultivierung und Düngemittel 

p Wasserversorgung und Umweltschutz 

p Anbau von Kulturpflanzen 

p Schädlingsbekämpfung 

p sonstige Anwendungen, wie Landminenräumung und 

Schutz des kulturellen Erbes 

2 Rolle der Internationalen Atomenergie 

Organisation 

Die Ziele der IAEO hinsichtlich der Anwendung von 

nuklearen Technologien sind ein Beitrag zur nachhaltigen 

Entwicklung und zum Umweltschutz in den Bereichen 

medizinische Versorgung, Ernährung, Landwirtschaft und 

Industrie sowie Versorgung mit Wasser. Die Aktivitäten 

der IAEO in diesen Bereichen haben Grundbedürfnisse der 

Menschheit im Fokus. Nukleare Technologien sollen dabei 

wesentlichere und konkurrenzfähigere Vorteile bieten als 

vergleichbare andere Technologien. 

Entsprechende nukleare Technologien und Anwendungen 

werden im Rahmen von koordinierten, angewandten 

Forschungsprojekten entwickelt. Dazu werden 

über mehrere Jahre Beiträge von potenziellen Anwendern 

aus Industriestaaten und Entwicklungsländern koordiniert. 

Bereits einsetzbare Technologien werden verfügbar 

gemacht und interessierten Ländern bereit gestellt; über 

technischer Kooperationsprojekte wird die technische 

Ausrüstung beschafft und es wird für eine entsprechende 

notwendige Ausbildung zum Betrieb von Analgen und 

Einsatz von Technologien gesorgt. 

Im derzeitigen Gesamtbild der weltweiten Entwicklung 

erkennt man folgende wichtige Problemfelder: 

p Die Weltbevölkerung hat fast 6 Milliarden Menschen 

erreicht und wächst pro Jahr um rund 80 Millionen 

an – dies entspricht etwa der deutschen Gesamtbevölkerung. 

p Fast eine Milliarde Menschen sind chronisch unterernährt 

und es fehlt ihnen zudem an einer zuver lässigen 

Wasserversorgung. 

p Sechs Millionen Kinder in der Dritten Welt sterben 

jedes Jahr aufgrund Mangelernährung; ebenso viele 

Erwachsene sterben in den Industriestaaten jährlich an 

Krebs. Weitere Millionen sind erkrankt oder sterben an 

bakteriellen oder von Viren verursachten Infektionen 

und Krankheiten. Zum Beispiel streben allein an den 

Folgen der Malaria weltweit mehr als 2 Millionen 

Menschen pro Jahr. 

Viele heutige Aktivitäten des Menschen verschmutzen und 

verändern die Umwelt und sind auf längere Sicht nicht 

tragbar. 

Die von der IAEO entwickelten und unterstützten 

nuklearen Anwendungen außerhalb des Energiesektors 

haben daher auch 

p die Verbesserung der Ernährungssituation und Wasserversorgung, 

p die Verbesserung des Gesundheitsschutzes sowie 

p den Schutz der Umwelt durch Analyse, Vorbeugung 

und Sanierung belasteter Bereiche 

zum Ziel. Trends der IAEO-Unterstützung von Projekten 

im Rahmen der technischen Kooperation sind in Tabelle 1 

dargestellt. 

3 Gesundheitsvorsorge 

Die meist verbreiteten diagnostischen Anwendungen von 

ionisierender Strahlung und nuklearer Technologie sind 

die Röntgendiagnose sowie die Diagnostik mit radioaktiven 

biologischen Indikatoren in der Nuklearmedizin. 

Die therapeutische Anwendung ist die Strahlungsonkologie, 

bei der mit einer wirksamen Strahlungsdosis an 

bestimmten krebsgefährdeten Stellen des Körpers mittels 

einer externen Strahlungsquelle (Teletherapie) oder 

mittels einer implantierten Strahlungsquelle (Brachytherapie) 

gearbeitet wird. Auch offene Strahlungsquellen, 

wie radioaktive Substanzen zur Behandlung von Schilddrüsenkrebs, 

werden manchmal verwendet. Neue Anwendungen 

von radioaktiven Substanzen werden erforscht 

und entwickelt. 


Protection of Man and Environment – Nuclear Usage Outside of Energy Sector ı P.P. De Regge, Wien


Projekte 

Eine revolutionäre Entwicklung der diagnostischen 

Radiologie war die Computertomografie, die genaue 

Schnittbilder von bestimmten Körperregionen in allen 

anatomischen und funktionellen Einzelheiten darstellt. 

Die Nuklearmedizin verwendet Radioisotope in über 

Hundert standardisierten Methoden für diagnostische, 

therapeutische und Forschungszwecke in den Bereichen 

Onkologie, Endokrinologie, Kardiologie, Neurologie und 

Nephrologie. Industriestaaten haben derzeit etwa 

20 Gammakameras pro einer Million Einwohner zur 

Verfügung, während in Entwicklungsländern nur eine 

pro eine Million Einwohnern verfügbar ist, und dies 

nur in den größeren Metropolen. 

Molekulare Nuklearmedizin ist ein neuer Anwendungsbereich. 

Erkrankungen werden auf zellulären oder 

gene tischem Niveau untersucht und identifiziert. Polymerase 

Kettenreaktionen bilden extrem empfindliche 

Diagnosetechniken für epidemische Infektionen, Krebs 

oder weiterer Symptome und sind z. B. von der US Food 

and Drug Administration als in vitro Tests genehmigt 

worden. 

Radioimmunoassay ist eine der wichtigsten Komponenten 

der in vitro Diagnosetechniken für die Quan tifizierung 

von Proteinänderungen oder anderer erkrankungsbedingter 

Metabolismusprodukte. Ein Zuwachs des 

Weltumsatzes für diese in vitro Diagnosetechniken von 

20 Milliarden Dollar im Jahr 1999 auf bis zu 26 Milliarden 

Dollar im Jahr 2004 wird prognostiziert und spiegelt die 

Bedeutung dieser Diagnosemöglichkeit wieder. 

Die Anzahl von Krebserkrankungen, eine der häufigsten 

Todesursachen in den Industriestaaten, nimmt auch aufgrund 

der steigenden Lebenserwartung in den Ländern 

der Dritten Welt ständig zu. Von etwa 10 Millionen neuen 

Krebsfällen im Jahr 2000 entfällt jeweils die Hälfte auf 

Industriestaaten und auf Entwicklungsländer. Es wird 

erwartet, dass 10 Millionen von 15 Millionen neuer 

Jahr 

1980 1985 1990 1995 2000 

Kernenergie 26 10 10 5 5 

Gesundheitsvorsorge 11 11 11 16 23 

Physik und Chemie 25 25 20 15 10 

Hydrologie 2 3 5 6 11 

Landwirtschaft und Ernährung 24 25 22 24 18 

Industrie 5 12 12 8 8 

Strahlenschutz 6 12 17 24 20 

Abfallverarbeitung 1 2 3 2 5 

| Tab. 1. 

IAEO Unterstützung von Projekten im Rahmen der Technischen Kooperation (Prozent der Gesamtfinanzierung) 

| Teletherapie. 

Krebsfälle im Jahr 2015 in Ländern der Dritten Welt auftreten 

werden. Fünfzig Prozent der Krebspatienten werden 

mittels Radiotherapie behandelt, mit einer Überlebenschance 

von ungefähr 45%. 

Am meisten wird noch Teletherapie verwendet, jetzt 

mit Präzisionsbestrahlung von komplexen und unregelmäßigen 

Tumorformen. Für Brachytherapie mittels 

implantierter Strahlungsquellen werden derzeit Caesium- 

137 und Iridium-192 anstelle von Radium-226 verwendet. 

Die IAEO, zusammen mit der Weltgesundheitsorganisation 

(World Health Organisation, WHO), verwaltet 

eine Datenbank von Spitälern und Kliniken, in denen 

Radio therapie zur Anwendung kommt. Länder der 

Dritten Welt mit 85 % der Bevölkerung verfügen über 

ein Drittel der Radiotherapieinstallationen, etwa 2200 

Teletherapie geräte mit Kobalt-60 Quellen und 850 

Brachytherapie geräten allerdings nur einem Fünftel der 

Elektronen beschleuniger. Mit jedem Gerät können 

600 Patienten pro Jahr behandelt werden oder insgesamt 

nur 1,9 Millionen von 2,5 Millionen Patienten in diesen 

Ländern. Mehr als 5000 zusätzliche Radiotherapiegeräte 

werden daher bis zum Jahr 2015 gebraucht werden. Am 

Rande sei erwähnt, dass die IAEO und die WHO jährlich 

die ordnungsgemäße Anwendung sowie die eingesetzte 

Strahlungsdosis von etwa 600 dieser Geräte in der Dritten 

Welt überprüfen. 

Nukleare Diagnose- und Forschungtechniken auf Basis 

stabiler oder radioaktiver Isotope werden auch zu Maßnahmen 

auf dem Ernährungssektor verwendet, insbesondere 

zum Nachweisen von Vitamin-, Spuren element- und 

Mikronährstoffmangel. Diese sind mit Ursache für Anämie, 

Beeinträchtigungen des Sehver mögens, Wachstum und 

geistige Entwicklung. Differen tielle Röntgenabsorp tiometrie 

wird derzeit auch als nicht intrusive Methode zur 

Knochendichtemessungen verwendet. 

4 Landwirtschaft 

Nukleare Technologien werden in der Landwirtschaft zur 

Verbesserung und zum Schutz von Nahrungsmitteln 

verwendet sowie zur Optimierung der Viehzucht und 

Vorbeugung gegen Seuchen und zur Bekämpfung von 

Schädlingen. Biodiversität ist entstanden durch zufällige 

Mutationen, induziert durch kosmische und natürliche 

Strahlung und auch durch Transkriptionsfehler im Erbgut. 

Durch Anwendung von Strahlung kann deshalb die 

Mutationsfrequenz erhöht und beschleunigt werden 

und man kann in kurzer Zeit eine große Anzahl von 

neuen Pflanzenvarianten bilden. Die meisten erzeugten 

Mutationen sind nicht weiter lebensfähig oder sogar 

schlechter überlebensfähig als die ursprüngliche Pflanze, 

aber einige sind unter Umständen besser angepasst und 

werden daher weiter gezüchtet. Einige Beispiele werden 

551 





552 


| Züchtungen von Pflanzen: 

Höhere Erträge durch den Einsatz von Bestrahlungen. 

hier angegeben, da sie für die betreffenden Länder und 

Regionen eine erhebliche Bedeutung in der zukünftigen 

Versorgung besitzen. Reissorten, geeignet zum Wachsen 

im Wechsel von Salzwasser und Süßwasser der asiatischen 

Flussdeltas, dienen als Grundnahrungsmittel von 

Millionen Menschen in Pakistan, China und Bangladesh. 

Gerstevarianten ohne Hülse, dürretolerantes Sorghum 

und Reis werden durch strahlungsinduzierte Mutationen 

hergestellt und gezüchtet, weil sie den klimatologischen 

und geologischen Bedingungen in Peru oder auch in Mali 

besser angepasst sind. Am besten eignen sich strahlungsinduzierte 

Mutationen für die Optimierung von Pflanzen, 

die sich durch sterile Klonung vermehren, wie Bananen. 

Eine früh blühende, herrlich schmeckende und seuchenresistente 

Variante Novaria wurde auf diese Art in Malaysia 

entwickelt. Seuchenresistente Sorten bringen nicht nur 

bessere Ausbeute sondern benötigen auch weniger oder 

gar keinen Pestizideinsatz und sind deshalb sowohl billiger 

als auch umweltschonender. Bananen und Reis sind die 

wichtigsten Grundnahrungsmittel und Landwirtschaftsprodukte 

in der Dritten Welt, aber auch in der Industriestaaten 

werde fast zweitausend durch strahlungsinduzierte 

Mutationen verbesserte Produkte konsumiert. 

In Österreich wurde aus Golden Delicious die Variante 

„Golden Haidegg” gezüchtet, mit einer schöneren Farbe 

und längerer Haltbarkeit ohne Rostflecken. Japanische 

Birnen mit strahlungsinduzierter Resistenz gegen 

Schwarzfleckenseuche werden derzeit gezüchtet. Dies 

reduziert den früher notwendigen Pestizideinsatz auf ein 

Viertel. 

Nukleare Techniken werden nicht nur zur Verbesserung 

der Nahrungsmittel sondern auch zu deren Schutz 

und Konservierung verwendet. In 30 Ländern wird 

Bestrahlung von Nahrungsmitteln zur Gesundheits- und 

| Bestrahlung von Nahrungsmitteln. 

Qualitätssicherung genutzt sowie zur Einhaltung der 

Quaran täneverordnungen für bestimmte Nahrungsmittel, 

wie Fleisch, Früchte, Kräuter und getrocknete Gemüse. 

Seit Mitte 2000 wird die Bestrahlung von Hackfleisch in 

den Vereinigten Staaten aus mikrobiologischen Sicherheitsgründen 

eingesetzt. In mehr als 2000 Supermärkten 

werden diese Produkte ohne signifikante Vorbehalte der 

Verbraucher angeboten. Viele Länder, insbesondere in 

Asien, Lateinamerika, dem Mittler Osten und Afrika 

würden enorme Nutzen durch der Anwendung dieser 

Technologie zum Schutz der Nahrungsmittel erzielen. 

Die weltweit internationale Autorität für die Sicherheit 

der Nahrungsmittel, die Kodex Alimentarius Kommission, 

hat auch im Bereich der Strahlenbehandlung von Nahrungsmitteln 

Protokolle, Richtlinien und Empfehlungen 

veröffent licht. 

Im Landwirtschaftsbereich findet man auch die Anwendung 

von Radioisotopen durch den Radioimmunoassay 

bei Tieren für Hormonanalysen von Milch, Serum oder 

Plasma. Zweck ist die Optimierung der Planung und 

Diagnose von der Trächtigkeit, die frühzeitige Erkennung 

von Gesundheits- und Reproduktionsmängeln und die 

zeitgerechte Identifikation und Vorbeugung vor Seuchen, 

wie Rinderpest, Trypanosomosis, Brucellosis und Maulund-Klauen-Seuche. 

Nukleare Technologien und Isotopmarkierungen 

mit Stickstoff, Kohlenstoff und Phosphor 

finden auch verbreitet Anwendungen zur Untersuchung 

von biologischer Stickstofffixierung durch Bakterien, zur 

Optimierung der Anwendung von Düngemitteln nach 

Menge, Art und Jahreszeit in den verschiedensten Klimazonen, 

Bodentypen und Ernährungskulturen sowie zur 

Minimierung der ungenutzten Phosphat- und Stickstoffmengen 

in der Umwelt. 

| Insektensterilisierung: Schädlingsbekämpfung. 

Beim Ausrotten von Schädlingen werden Nukleare 

Technologien unter der Bezeichnung „Sterile Insekten- 

Technik“ verwendet. Diese Technik beinhaltet die 

großräumige Produktion und systematische Freisetzung 

von strahlungssterilisierten männlichen Fruchtfliegen, 

Tsetse-Fliegen, Holzwürmer usw. Freigelassen, verhalten 

sie sich auf ganz normale Weise, aber produzieren keinerlei 

Nachwuchs. Die schon geschwächte nächste Generation 

wird wieder mit neuen sterilen Männchen versorgt, 

bis sie lokal gezielt ausgerottet ist, ohne Pestizide, ohne 

Umwelt zerstörung und unter Erhalt des biologischen 

Gleich gewichtes. Insbesondere die Nachfrage und 

Produktionskapazität für sterilisierte Mittelmeerfruchtfliegen, 

notorische Schädlinge mit erheblichen Auswirkungen, 

sind im Lauf der letzten Dekade von einer auf 

drei Milliarden Fliegen pro Woche gestiegen. Am meisten 

Erfolg hatten die Ausrottungskampagnen in Mittel- und 




| Insektensterilisierung: Schädlingsbekämpfung. 

Lateinamerika, den Südlichen Vereinigten Staaten, 

Australien und Japan. Subtile genetisch eingebaute Unterschiede 

erlauben die Sterilisierung und Aussortierung der 

männlichen Insekten schon im Puppenstadium. Dementsprechend 

werden die Zuchtkosten gesenkt und die 

Effektivität wird erhöht. Auf diese Weise wird diese Technologie 

gleich einem selektiven Insektizid zur kompletten 

Ausrottung verwendet. Im Rahmen eines Projektes des 

IAEO Programms zur „Technischen Kooperation“ wurde in 

Tansania eine Produktionsanlage für Millionen sterilisierte 

Tsetse Fliegen errichtet, womit verschiedene afrikanische 

Regionen von diesen Schädlingen befreit werden konnten. 

Derzeitige Forschungsprojekte in diesem Bereich entwickeln 

weitere sterile Insekten-Techniken für Motten und 

Würmer. Die IAEO hat ein ehrgeiziges Projekt für die 

Ausrottung von Malaria Moskitos gestartet. 

Die wirtschaftlichen Erfolge dieser Technologie sind 

insgesamt herausragend. Die Bekämpfung von Fruchtfliegen 

in Mexiko mit einem Kapitaleinsatz von 

10 Millionen Dollar pro Jahr bewirkt einen Ertrag von 

einer Milliarde Dollar Wert an Zitrusfrüchten und Gemüse. 

Ähnliche Ergebnisse wurden in Chile erzielt, insbesondere 

da die Zitrusfrüchte dort während des nördlichen Winters 

produziert werden und somit einen maßgeblichen Einfluss 

auf den Export haben. 

5 Hydrologie 

Nukleare und Isotopentechniken werden auch in der 

Hydrologie und Klimatologie genutzt, wo sich aufgrund 

der natürlich entstandenen Fingerabdrücke der Isotopenzusammensetzung 

die Herkunft und das Alter der Wasservorräte 

nachweisen lässt. Isotopentechniken, basierend 

auf Messungen von stabilen und radioaktiven Wasserkomponenten, 

werden zur Modellierung von Wassersystemen 

verwendet, um die in vielen Regionen 

| Isotopentechniken in der Hydrologie: Wasservorräte erweitern. 

beschränkten und kostbaren Wasserressourcen optimal zu 

nutzen und ein lokales Gleichgewicht zwischen verbrauchten 

und erneuerbaren Wassermengen zu erzielen. 

Diese Methoden werden auch zur Identifizierung und zum 

Nachweis von unerwünschten Vermischungen der Wasserversorgung 

mit Abwässern verwendet, wie es in dichtbesiedelten 

Großmetropolen öfter vorkommt. Die 

Temperaturabhängigkeit der Wasserisotopenzusammensetzung 

beim Verdampfen und Kondensieren wird in 

weltweiten Forschungen von Klima- und Treibhauseffekt 

genutzt. So ist die Klimageschichte von zehntausenden 

Jahren aufgrund der Isotopenanalysen von Gletschereisschichten 

oder antarktischen Ablagerungen rekonstruiert 

worden. 

Beispiele zur Angabe der wirtschaftlichen und humanitären 

Bedeutung dieser nuklearen Techniken in der 

Hydrologie sind: 

p In Venezuela wurde das Trinkwasserdefizit um 30 % 

verringert durch Inbetriebnahme von 50 mit Isotopentechniken 

georteten Frischwasserquellen. 

p Durch Grundwassermessungen im Niltal und in den 

ägyptischen und äthiopischen Wüsten wurden erneuerbare 

Wasserreserven in Nubischen Sandsteinschichten 

geortet womit die Wasserversorgung um 20 % erhöht 

werden konnte. 

p Ebenfalls in Venezuela wurde aufgrund von Isotopenmessungen 

die Zuverlässigkeit eines Stauseedammes 

nachgewiesen. Eingeplante Reparaturkosten in 

Höhe von 6 Mio. US-Dollar konnten somit eingespart 

werden. 

p Mit dem Ausbau von geothermischen Energiequellen 

werden in Mittelamerika und Südostasien Millionen 

Dollar an Ölimporten gespart. Isotopentechniken 

dienen dabei zur Suche von geeigneten Standorten und 

zur Optimierung der Anlageneinrichtungen. 

6 Umweltschutz 

Nukleare Technologien und Messtechniken werden weit 

verbreitet zur Überwachung und Erforschung der Umwelt 

eingesetzt. Überhaupt werden künstliche und natürliche 

Radioisotope in der Atmosphäre, in der Hydrosphäre und 

im Boden für radiologische Zwecke überwacht. Sie sind 

zudem ausgezeichnete Indikatoren für atmosphärische 

und ozeanografische Transportprozesse, für das Verhalten 

und den Transport von nicht radioaktiven Umweltgiften, 

für die geochronologische Altersbestimmung und für 

ökologische und biologische Forschungszwecke. Einfache 

tragbare Röntgenfluoreszenzgeräte mit Radioisotopenquellen, 

Forschungsreaktoren für die Neutronenaktivierungsanalyse 

und Teilchenbeschleuniger zur 

Charakterisierung von mikroskopischen Teilchen formen 

zusammen ein Arsenal von nuklearen Techniken zum 

Identifizieren von schädliche Substanzen in der Umwelt 

und zur genauer Identifizierung ihrer Zusammensetzung 

und Herkunft. 

Nicht nur zur Diagnose von schädlichen Substanzen, 

sondern auch zur Vorbeugung oder Eliminierung ihrer 

Effekte werden nukleare Technologien verwendet. 

Elektronenbeschleuniger werden in Kohlekraftwerken 

verschiedener Länder zur Abgasbehandlung eingesetzt. 

Schwefel- und Stickstoffoxiden werden im Rauchgas 

umgewandelt und mit zugefügtem Ammoniak zu Düngemittel 

umgesetzt. Abwässer, organisch verseuchtes Wasser 

und Klärschlamm werden mit ionisierender Strahlung 

behandelt und desinfiziert. 

Zur Umweltsanierung gehört auch die Ortung und 

Entfernung von rund 60 Millionen Landminen, die in 

553 





554 


| Lokalisierung von Landminen. 

62 Ländern Jahr für Jahr Ursache für Tausende Tote und 

Verletzte sind. Nukleare Technologien werden auch für 

diese Anwendungen entwickelt. Derzeit verfügbare Geräte 

benutzen Neutronengeneratoren und Gammadetektoren 

zur Auswertung von Elementverhältnissen, insbesondere 

Wasserstoff, Stickstoff, Kohlenstoff und Sauerstoff, in 

minenverdächtigen Gegenständen im Boden. 

7 Industrie und Archäologie 

Unzählige Anwendungen von nuklearen Technologien, 

ionisierender Strahlung und Isotopen werden auch in 

allgemeinen Bereichen der Industrie genutzt, so in 

Transport- und Bauunternehmen für radiografische 

Materialkontrolle, in der Polymerverarbeitung (z. B. 

Autoreifen), für die Sterilisierung sowie das Vernichten 

von Viren, Bakterien und Pilzen im medizinischen 

Bereich. 

Minuten quadratmillimeter kleine Einzelheiten von 

Gemälden, Skulpturen, Polychromen, Münzen oder 

Keramiken zu charakterisieren und auszuwerten. Ein 

Prototyp kam im Kunsthistorischen Museum in Wien für 

die Identifizierung von Gemälden des 16ten Jahrhunderts 

und zur Katalogisierung von Etruskischen Bronzen und 

Münzen zum Einsatz. 

8 Zusammenfassung 

Nukleare Technologien, Isotopentechniken und Strahlungsanwendungen 

sind weit verbreitete Anwendungen 

und liefern unersetzbare Beiträge zur Verbesserung und 

Erhaltung des heutigen Lebensstandards sowie zum 

Schutz des Menschen und seiner Umwelt. Das Ausmaß 

dieser Anwendungen für diagnostische und thera peutische 

Zwecke im Gesundheitsbereich wurde dargestellt und 

Erfolge und Erwartungen der Entwicklungen von 

Computertomografie und molekularer Nuklearmedizin 

wurden erwähnt. Im therapeutischen Bereich ist ohne 

Zweifel eine Erweiterung der Anwendungen zu erwarten, 

inbesondere in der Dritten Welt. Anwendungen in der 

Landwirtschaft, der Nahrungsversorgung und dem Schutz 

der Nahrung, wie Pflanzenoptimierung, Ausrottung von 

Schädlingen, Vorbeugung vor Seuchen und Konservierung 

von Nahrungsmittel sind vollständig ausgereift und 

werden akzeptiert. Bei komplexeren Problemen kommen 

ebenfalls erfolgreiche Maßnahmen zum Einsatz, wie 

im Zusammenhang mit der Bekämpfung von Malaria, 

Trypanosomosis und Myasis. Anwendungen in der Hydrologie 

und im Umweltschutz sind seit Jahren etabliert und 

anerkannt. Im Bereich der Klimatologie und in der meeresoder 

landesbezogenen Ökologie werden sie erweitert 

genutzt. Nukleare Techniken und Technologien in den 

hier beschriebenen Anwendungsbereichen werden nicht 

aus reinem Interesse an der Sache oder aufgrund 

popukärer Überlegungen eingesetzt, oder weil besondere 

Förderungen zu erwarten wären, sondern weil sie nach 

vergleichender Begutachtung und Kosten-Nutzen-Analyse 

optimalere Ergebnisse versprechen als vergleichbare 

verfügbare Technologien. 

Literatur 

1. Annual Report 2000. International Atomic Energy Agency, Vienna, 2001 

2. Nuclear Technology Review 2002. Document GOV/2002/7, International Atomic Energy Agency, 

Vienna, February 2002 

3. The International Atomic Energy Agency‘s Laboratories, Meeting the Challenges of Research and 

International Cooperation in the Application of Nuclear Techniques. Document IAEA/PI/A67E 

99-02059, August 1999 

| Kulturelles Erbe: Analysen zur Bestätigung der Herkunft. 

Interessant und vielleicht weniger bekannt ist die 

Anwendung von zerstörungsfreien nuklearen Technologien 

und Messmethoden im Bereich der Archäologie und 

der Konservierung des Kulturellen Erbes. Vor kurzer Zeit 

hat die IAEO ein Projekt in Lateinamerika zur Förderung 

von interdisziplinärer Forschung und Unterstützung in 

diesem Bereich koordiniert. Jedes teilnehmende Land 

hat sich mit bestimmten ungelösten Fragen bezüglich 

Keramiken befasst und mittels Spurenanalyse durch 

Röntgenfluoreszenz- und Neutronenaktivierung das 

erwünschte Ergebnis endgültig erhalten. 

In Zusammenarbeit mit polnischen, kubanischen und 

österreichischen Wissenschaftlern hat die IAEO ein tragbares 

Röntgenfluoreszenzgerät entwickelt, das ideal für 

die zerstörungsfreie Analyse von Kunstgegenstände 

eingesetzt werden kann. Es ist mit Computer- und Lasersteuerung 

ausgestattet und in der Lage, innerhalb von 

Verfasser 

Peter P. De Regge 

Head, PCI Laboratorien, 

Internationale Atomenergie Organisation (IAEO), 

Wagramer Straße 5, 1400 Wien, Austria 




Inside 

Women in Nuclear bei Framatome/ANF in Karlstein 

Die deutsche Sektion des international agierenden Frauennetzwerks 

WiN (Women in Nuclear) traf sich am 10. und 

11. Oktober 2019 bei Framatome in Karlstein. Nach der 

Mitgliederversammlung von WiN Germany wurde das 

Treffen mit Teilnehmerinnen aus Schweden, Finnland und 

der Schweiz international. 

Unter dem Dach des gemeinnützigen Vereins Women in 

Nuclear (WiN) Germany – gegründet 2008 – möchten 

Frauen, die auf den Gebieten Kernenergie, Strahlenschutz, 

Nuklearmedizin und nukleare Wissenschaften arbeiten, 

dazu beitragen, einen transparenten Dialog über die Notwendigkeit 

der nuklearen Kompetenzen für Deutschland 

zu führen. 

Chantal Greul, Präsidentin von WiN Germany, erklärt: 

„Auch wenn sich unser Land mehrheitlich für den Ausstieg 

aus der Stromerzeugung durch Kerntechnik entschieden 

hat, bleiben das Know-how und das hohe Sicherheitsverständnis 

Deutschlands nach dem Ausstieg 2022 

wichtig: für den Rückbau, die Endlagerung – und nicht 

zuletzt als Beitrag zum Erhalt des höchstmöglichen Sicherheitsniveaus 

weltweit. Wir ermutigen junge Frauen, eine 

berufliche Laufbahn in der Nukleartechnik zu wählen. 

Denn nur durch gut ausgebildete Nachwuchskräfte bleiben 

die notwendigen Ressourcen in Deutschland langfristig 

gesichert.“ 

| Im Foyer der Framatome GmbH in Karlstein mit den Standortleitern 

Stefan Rosenberger der Framatome GmbH (links) und Matthias Gutjahr 

der Advanced Nuclear Fuels GmbH (Mitte). 

| v.l. Chantal Greul, WiN Germany Präsidentin, 

Preisträgerin Jenny Jessat, Martina Ezmuß, 

Vorstand WiN Germany. 

Einen wichtigen Beitrag zu diesem Vereinsziel leistet 

die jährliche Verleihung des WiN-Germany Preises. Für 

diesen mit 500 Euro dotierten Preis gab es in diesem Jahr 

gleich fünf Bewerberinnen. Beiträge aus Wissenschaft und 

Forschung sowie aus Unternehmen ließen eine hohe 

fachliche Kompetenz erkennen, die die Entscheidung für 

eine Bewerberin schwer fallen ließ. Jenny Jessat überzeugte 

den Kreis mit dem Thema „Studies on the interaction 

of plant cells with uranium (VI) and europium (III) 

an on stress-induced metabolite release” („Studien zur 

Wechselwirkung von Pflanzenzellen mit Uran (VI) und 

Europium (III) und zur stressinduzierten Metabolitfreisetzung“). 

Es handelte sich um die Präsentation 

ihrer Masterarbeit in Chemie an der TU Dresden. Jenny 

Jessat ist derzeit wissenschaftliche Mitarbeiterin am 

Helmholtz-Zentrum Dresden-Rossendorf, Institut für 

Ressourcenökologie und bereitet dort ihre Promotion vor. 

Neben weiteren Fachvorträgen über kerntechnische 

Themen fanden die Werksführungen sehr viel Anklang bei 

den Teilnehmerinnen. „Wir möchten uns herzlich bei den 

beiden Standortleitern der Framatome GmbH und ANF 

bedanken. Insbesondere die Standortbesuche waren ein 

wirkliches Highlight und die Teilnehmerinnen waren 

damit sehr zufrieden“, lobte Chantal Greul den Einsatz der 

Framatome für die Veranstaltung.“ 

| Aus der Komponentenfertigung der ANF 

Karlstein stellte Petra Denner die 

Entwicklungs historie des neuen GAIA 

Abstandhalters vor und zeigte an diesem 

Beispiel auf, wie die hohen Qualitätsanforderungen 

der Kunden erfüllt werden. 

Karin Reiche 

WiN 

KTG Inside 

Verantwortlich 

für den Inhalt: 

Die Autoren. 

Lektorat: 

Natalija Cobanov, 

Kerntechnische 

Gesellschaft e. V. 

(KTG) 



T: +49 30 498555-50 

F: +49 30 498555-51 

E-Mail: 

natalija.cobanov@ 

ktg.org 

555 

KTG INSIDE 

Herzlichen Glückwunsch! 

www.ktg.org 

Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag 

und wünscht ihnen weiterhin alles Gute! 

Dezember 2019 

50 Jahre | 1969 

13. Bernd Gulich, Tiefenbach-Ast 

55 Jahre | 1964 

30. Thomas Schmidt, Lörrach 

60 Jahre | 1959 

15. Axel Lenzen, Titz 

20. Martin Schlieck-Weber, Hausen 

65 Jahre | 1954 

24. Reinhold Paul, Hanau 

70 Jahre | 1949 

2. Dipl.-Ing. Berndt Standfuß, Dresden 

28. Fritz Grimm, Alzenau 

76 Jahre | 1943 

8. Dr. Dieter Herrmann, Brandis 

77 Jahre | 1942 

8. Karl Georg Weber, Neckarwestheim 

14. Günter Breiling, Weinheim 

78 Jahre | 1941 

13. Dipl.-Ing. Klaus-Dieter Hnilica, 

Rodenbach/Hanau 

Wenn Sie künftig eine 

Erwähnung Ihres 

Geburtstages in der 

atw wünschen, teilen 

Sie dies bitte der KTG- 

Geschäftsstelle mit. 

KTG Inside


556 

79 Jahre | 1940 

8. Dipl.-Ing. Wolfgang Heess, Laudenbach 

21. Dr. Jürgen Wehmeier, Springe 

94 Jahre | 1925 

10. Dr. Arthur Pilgenröther, Kleinostheim 

83 Jahre | 1937 

9. Dipl.-Ing. Werner Rossbach, 

Bergisch Gladbach 

NEWS 

79 Jahre | 1940 

16. Dipl.-Ing. Wolfgang Breyer, Buckenhof 

19 Prof. Dr. Wernt Brewitz, Wolfenbüttel 

80 Jahre |1939 

1. Dipl.-Ing. Georg Dumsky, Gräfelfing 

6. Dipl.-Ing. Hans-Henn. Kuchenbuch, 

Laboe-Brodersdorf 

27 Dr. Horst Bauer, Sigless/AT 

81 Jahre | 1938 

1. Dr. Gert Spannagel, 

Linkenheim-Hochstetten 

82 Jahre | 1937 

30. Dipl.-Ing. Wilhelm Weiss, Weinheim 

83 Jahre | 1936 

7. Dipl.-Ing. Aurel Badics, Bad Kreuznach 

17. Prof. Dr.-Ing. Rolf Theenhaus, Linnich 

86 Jahre | 1933 

10. Prof. Dr. Jürgen Vollradt, 

Unna-Königsborn 

Januar 2020 

55 Jahre | 1965 

31. Eckhard Stengert, Worms 

60 Jahre | 1960 

26. Dr. Friedhelm Funke, Dormitz 

70 Jahre | 1950 

15. Dipl.-Ing. Andreas Hüttmann, Oering 

78 Jahre | 1942 

31. Dipl.-Phys. Werner Scholtyssek, 

Stutensee 

79 Jahre | 1941 

15. Dipl.-Ing. Ulf Rösser, 

Heiligkreuzsteinach 

81 Jahre | 1939 

16. Dr. Wolfgang Kersting, Blieskastel 

82 Jahre | 1938 

7. Dipl.-Ing. Manfred Schirra, Stutensee 

12. Dipl.-Ing. Hans Dieter Adami, Rösrath 

22. Dr. Franz Müller, Erlangen 

84 Jahre | 1936 

5. Obering. Peter Vetterlein, Oberursel 

23. Prof. Dr. Hartmut Schmoock, 

Norderstedt 

30. Dipl.-Phys. Wolfgang Borkowetz, 

Rüsselsheim 

30. Dipl.-Ing. Friedrich Morgenstern, Essen 

85 Jahre |1935 

10. Dipl.-Ing. Walter Diefenbacher, 

Karlsruhe 

17. Dipl.-Ing. Helge Dyroff, Alzenau 

24. Theodor Himmel, Bad Honnef 

87 Jahre |1933 

9. Prof. Dr. Hellmut Wagner, Karlsruhe 

88 Jahre | 1932 

3. Dipl.-Ing. Fritz Kohlhaas, Kahl/Main 

91 Jahre | 1929 

20. Dr. Devana Lavrencic-Cannata, Rom/I 

93 Jahre | 1927 

1. Prof. Dr. Werner Oldekop, 

Braunschweig 

Top 

World Energy Outlook 2019 

highlights deep disparities in 

the global energy system 

(iea) Deep disparities define today’s 

energy world. The dissonance 

between well-supplied oil markets 

and growing geopolitical tensions and 

uncertainties. The gap between the 

ever-higher amounts of greenhouse 

gas emissions being produced and the 

insufficiency of stated policies to 

curb those emissions in line with 

international climate targets. The gap 

between the promise of energy for 

all and the lack of electricity access 

for 850 million people around the 

world. 

The World Energy Outlook 2019, 

the International Energy Agency’s 

flagship publication, explores these 

widening fractures in detail. It explains 

the impact of today’s decisions 

on tomorrow’s energy systems, and 

describes a pathway that enables the 

world to meet climate, energy access 

and air quality goals while maintaining 

a strong focus on the reliability 

and affordability of energy for a 

growing global population. 

As ever, decisions made by 

governments remain critical for the 

future of the energy system. This is 

evident in the divergences between 

WEO scenarios that map out different 

routes the world could follow over the 

coming decades, depending on the 

policies, investments, technologies 

and other choices that decision 

makers pursue today. Together, these 

scenarios seek to address a fun damental 

issue – how to get from where 

we are now to where we want to go. 

The path the world is on right now 

is shown by the Current Policies 

Scenario, which provides a baseline 

picture of how global energy systems 

would evolve if governments make no 

changes to their existing policies. In 

this scenario, energy demand rises by 

1.3 % a year to 2040, resulting in 

strains across all aspects of energy 

markets and a continued strong 

upward march in energy-related 

emissions. 

The Stated Policies Scenario, 

formerly known as the New Policies 

Scenario, incorporates today’s policy 

intentions and targets in addition to 

existing measures. The aim is to hold 

up a mirror to today’s plans and 

illustrate their consequences. The 

future outlined in this scenario is still 

well off track from the aim of a secure 

and sustainable energy future. It 

describes a world in 2040 where 

hundreds of millions of people still go 

without access to electricity, where 

pollution-related premature deaths 

remain around today’s elevated levels, 

and where CO 2 emissions would 

lock in severe impacts from climate 

change. 

The Sustainable Development 

Scenario indicates what needs to be 

done differently to fully achieve 

climate and other energy goals 

that policy makers around the world 

have set themselves. Achieving this 

scenario – a path fully aligned with 

the Paris Agreement aim of holding 

the rise in global temperatures to well 

below 2° C and pursuing efforts to 

limit it to 1.5° C – requires rapid and 

widespread changes across all parts of 

the energy system. Sharp emission 

cuts are achieved thanks to multiple 

fuels and technologies providing 

efficient and cost-effective energy 

services for all. 

“What comes through with crystal 

clarity in this year’s World Energy 

News


Outlook is there is no single or simple 

solution to transforming global energy 

systems,” said Dr Fatih Birol, the IEA’s 

Executive Director. “Many technologies 

and fuels have a part to play 

across all sectors of the economy. For 

this to happen, we need strong leadership 

from policy makers, as governments 

hold the clearest responsibility 

to act and have the greatest scope to 

shape the future.” 

In the Stated Policies Scenario, 

energy demand increases by 1% per 

year to 2040. Low-carbon sources, led 

by solar PV, supply more than half of 

this growth, and natural gas accounts 

for another third. Oil demand flattens 

out in the 2030s, and coal use edges 

lower. Some parts of the energy sector, 

led by electricity, undergo rapid 

transformations. Some countries, 

notably those with “net zero” 

aspirations, go far in reshaping all 

aspects of their supply and consumption. 

However, the momentum behind 

clean energy is insufficient to offset 

the effects of an expanding global 

economy and growing population. 

The rise in emissions slows but does 

not peak before 2040. 

Shale output from the United 

States is set to stay higher for longer 

than previously projected, reshaping 

global markets, trade flows and 

security. In the Stated Policies 

Scenario, annual US production 

growth slows from the breakneck 

pace seen in recent years, but the 

United States still accounts for 85 % of 

the increase in global oil production to 

2030, and for 30 % of the increase in 

gas. By 2025, total US shale output 

(oil and gas) overtakes total oil and 

gas production from Russia. 

“The shale revolution highlights 

that rapid change in the energy 

system is possible when an initial push 

to develop new technologies is 

complemented by strong market 

incentives and large-scale investment,” 

said Dr Birol. “The effects 

have been striking, with US shale now 

acting as a strong counterweight to 

efforts to manage oil markets.” 

The higher US output pushes down 

the share of OPEC members and 

Russia in total oil production, which 

drops to 47 % in 2030, from 55 % in 

the mid-2000s. But whichever pathway 

the energy system follows, the 

world is set to rely heavily on oil supply 

from the Middle East for years to 

come. 

Alongside the immense task of 

putting emissions on a sustainable 

trajectory, energy security remains 

paramount for governments around 

the globe. Traditional risks have not 

gone away, and new hazards such as 

cybersecurity and extreme weather 

require constant vigilance. Meanwhile, 

the continued transformation 

of the electricity sector requires policy 

makers to move fast to keep pace with 

technological change and the rising 

need for the flexible operation of 

power systems. 

“The world urgently needs to put a 

laser-like focus on bringing down 

global emissions. This calls for a grand 

coalition encompassing governments, 

investors, companies and everyone 

else who is committed to tackling 

climate change,” said Dr Birol. “Our 

Sustainable Development Scenario 

is tailor-made to help guide the 

members of such a coalition in their 

efforts to address the massive climate 

challenge that faces us all.” 

A sharp pick-up in energy efficiency 

improvements is the element that 

does the most to bring the world 

towards the Sustainable Development 

Scenario. Right now, efficiency 

improvements are slowing: the 1.2 % 

rate in 2018 is around half the average 

seen since 2010 and remains far below 

the 3 % rate that would be needed. 

Electricity is one of the few energy 

sources that sees rising consumption 

over the next two decades in the 

Sustainable Development Scenario. 

Electricity’s share of final consumption 

overtakes that of oil, today’s 

leader, by 2040. Wind and solar PV 

provide almost all the increase in 

electricity generation. 

Putting electricity systems on a 

sustainable path will require more 

than just adding more renewables. 

The world also needs to focus on 

the emissions that are “locked in” 

to existing systems. Over the past 

20 years, Asia has accounted for 

90 % of all coal-fired capacity built 

worldwide, and these plants potentially 

have long operational lifetimes 

ahead of them. This year’s WEO considers 

three options to bring down 

emissions from the existing global 

coal fleet: to retrofit plants with carbon 

capture, utilisation and storage or 

biomass co-firing equipment; to 

repurpose them to focus on providing 

system adequacy and flexibility; or to 

retire them earlier. 

| www.iea.org 

New report and webinar 

on the supply of medical 

radioisotopes 

(nea) The NEA hosted a webinar on 

18 November 2019 to present findings 

from a new report on the supply of 

medical radioisotopes, jointly produced 

with the Organisation for 

Economic Co-operation and Development 

(OECD) Health Committee. 

Technetium-99m (Tc-99m) is the 

most commonly used medical 

radioisotope, essential for 85 % of the 

nuclear medicine diagnostic scans 

performed worldwide. There are no 

comparable substitutes available for 

diagnoses of various cancers and for a 

range of diagnostics in children. 

Unfortunately, the global supply of 

Tc-99m is not technically and economically 

robust, and the existing 

supply-chain continues to experience 

chronic shortages. This new study 

analyses the current market structure 

and identifies barriers for the implementation 

of full cost recovery. 

Report and the webinar recording 

are available at: 

oe.cd/nea-med-rad-webinar-2019. 

| www.oecd-nea.org 

World Nuclear Performance 

Report 2019 Asia Edition 

launched 

(wna) The World Nuclear Performance 

Report 2019 Asia Edition shows 

that nuclear energy in Asia is meeting 

the growing demand for electricity, 

whilst helping to tackle 

air pollution and climate change. The 

report, published by World Nuclear Association, 

was launched today at Singapore 

International Energy 

Week.Nuclear generation in Asia 

continued its rapid growth in 2018, increasing 

by 12 %. By replacing 

coal-fired generation nuclear energy 

avoided the emission of over 

500 million tonnes of carbon dioxide 

last year.Agneta Rising, Director 

General of World Nuclear Association 

said, “Nuclear is fast, scalable and a 

long-lasting way to add clean 

557 

NEWS 

News


Operating Results July 2019 

558 

NEWS 

Plant name Country Nominal 

capacity 

Type 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 

Energy generated, gross 

[MWh] 

Month Year Since 

commissioning 

Time availability 

[%] 

Energy availability 

[%] *) Energy utilisation 

[%] *) 

Month Year Month Year Month Year 

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 

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 

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 

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 

KKB 2 Beznau 7) PWR CH 380 365 744 266 954 1 925 011 136 275 418 100.00 100.00 100.00 100.00 94.22 99.56 

KKG Gösgen 3,7) PWR CH 1060 1010 611 634 890 4 627 036 318 502 564 82.06 86.72 82.03 86.04 80.50 85.81 

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 

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 

Dukovany B1 PWR CZ 500 473 744 363 425 2 485 594 114 715 087 100.00 99.86 99.31 99.61 97.70 97.72 

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 

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 


Temelin B1 PWR CZ 1080 1030 744 798 674 3 926 675 118 287 717 100.00 72.28 99.96 71.97 99.21 71.34 

Temelin B2 1,2) PWR CZ 1080 1030 0 0 4 477 198 113 749 715 0 80.87 0 80.81 0 81.34 

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 

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 

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 

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 

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 

Tihange 2 2) PWR BE 1055 1008 693 659 639 659 639 255 311 569 93.13 13.62 87.24 12.76 84.43 12.35 

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 

Plant name 

Type 

Nominal 

capacity 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 


[MWh] 


[%] 



[%] *) 

Month Year Since Month Year Month Year Month Year 

commissioning 

KBR Brokdorf 1,2) DWR 1480 1410 544 722 179 5 437 455 356 005 265 73.08 81.11 67.73 76.05 65.25 71.95 

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 

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 

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 

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 

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 

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 

*) 

Net-based values 

(Czech and Swiss 

nuclear power 

plants gross-based) 

1) 

Refueling 

2) 

Inspection 

3) 

Repair 

4) 

Stretch-out-operation 

5) 

Stretch-in-operation 

6) 

Hereof traction supply 

7) 

Incl. steam supply 

8) 

New nominal 

capacity since 

January 2016 

9) 

Data for the Leibstadt 

(CH) NPP will 

be published in a 

further issue of atw 

BWR: Boiling 

Water Reactor 

PWR: Pressurised 

Water Reactor 

Source: VGB 

elec tricity generation.“For the 90 

reactors that have started operating 

from 2000 to today, the typical construction 

time is 5 to 7 years. Of those 

90 reactors, 27 % were built in less 

than five years – and they will provide 

clean and reliable electricity for more 

than 60 years or more.”Many reactors 

in operation today are planned to 

operate for 60-80 years. Reactors are 

already demonstrating high performance 

irrespective of how long 

they have been in operation, with 

capacity factors of around 80 % 

maintained egardless of age.The 

report profiles Tarapur 1, a reactor 

located in Palghar, India, which 

marked 50 years of operation in 

April 2019. Four other reactors will 

match this achievement in 2019, 

the first year in which reactors have 

passed this milestone. Worldwide 

nuclear generation in 2018 increased 

for the sixth successive year, reaching 

2563 TWh. This is more than 

10 % of global electricity demand. 

Overall, capacity additions for the 

period 2016-2020 are expected to 

reach the targets of the nuclear 

industry’s Harmony programme. 

But build rates will have to increase 

significantly to achieve the overall goal 

of supplying 25 % of global electricity 

demand before 2050. Agneta Rising 

said, “Nuclear energy is key to Asia 

meeting the twin challenges of a growing 

demand for electricity, and an 

urgent need to switch to less polluting, 

low-carbon generation sources. More 

and more organizations are recognizing 

that nuclear energy is vital to the 

goal of a sustainable future for people 

and the planet.” 

| www.world-nuclear.org 

World Science Day: 

International Nuclear 

Information System 

highlights open access 

(iaea) Public awareness of developments 

in scientific research is crucial 

for building a more informed global 

society, and the IAEA is providing 

access to a trove of research on the 

peaceful uses of nuclear science and 

technology. 

The IAEA’s International Nuclear 

Information System (INIS) is one of the 

world’s largest repositories of published 

research on the peaceful uses of nuclear 

science and technology, with more 

than 4.2 million bibliographic records 

and access to more than 1.6 million 

full-text documents. Each year, more 

than 1 million unique users perform 

around 2 million searches and download 

3 million pages. 

News


Operating Results August 2019 

Plant name Country Nominal 

capacity 

Type 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 


[MWh] 

Month Year Since 

commissioning 


[%] 



[%] *) 

Month Year Month Year Month Year 

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 

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 

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 

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 

KKB 2 Beznau 1,2,7) PWR CH 380 365 239 86 668 2 011 679 136 362 086 32.12 91.34 31.92 91.31 30.14 90.70 

KKG Gösgen 7) PWR CH 1060 1010 495 511 876 5 138 912 319 014 440 66.47 84.14 66.04 83.49 64.91 83.14 

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 

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 

Dukovany B1 3) PWR CZ 500 473 379 176 941 2 662 535 114 892 028 50.94 93.62 50.34 93.32 47.57 91.32 

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 



Temelin B1 PWR CZ 1080 1030 744 793 753 4 720 428 119 081 470 100.00 75.82 99.86 75.52 98.60 74.82 

Temelin B2 1) PWR CZ 1080 1030 511 537 389 5 014 587 114 287 104 68.68 79.32 66.85 79.03 66.76 79.48 

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 

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 

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 

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 

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 

Tihange 2 PWR BE 1055 1008 744 756 385 1 416 023 256 067 953 100.00 24.64 99.92 23.88 97.11 23.16 

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 

559 

NEWS 

Plant name 

Type 

Nominal 

capacity 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 


[MWh] 


[%] 



[%] *) 

Month Year Since Month Year Month Year Month Year 

commissioning 

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 

KKE Emsland DWR 1406 1335 744 1 026 542 6 728 057 353 547 026 100.00 83.78 100.00 83.66 98.11 82.06 

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 

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 

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 

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 

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 

In line with the theme of this year’s 

World Science Day for Peace and 

Development – “Open Science, leaving 

no one behind” – on 10 November, 

INIS plays a vital role in providing 

open access to information and 

research. 

“The IAEA has been an advocate of 

open access to knowledge since the 

inception of INIS in 1970,” said 

Dobrica Savic, Head of the Nuclear 

Information Section. All the documents, 

presentations and records are 

free of charge and accessible to anyone 

via the Internet. 

Open science is defined by 

UNESCO as making scientific research 

and data accessible to all. It includes 

publishing open scientific research, 

promoting open access and making 

open source software available. 

Libraries and repositories are key for 

open science as many play an active 

role in the preservation, curation, 

publication and dissemination of 

digital scientific materials, according 

to the Organisation for Economic 

Co-operation and Development. 

The INIS repository includes 

research on a wide range of nuclearrelated 

topics, from environmental science 

and energy storage to radio logy 

and many more. It’s made possible 

through an ongoing colla borative 

effort between the IAEA and experts 

from 132 countries and 24 international 

organizations. Member States 

submit peer-reviewed research from 

their country/organization to INIS, 

and approximately 120,000 bibliographic 

records and 13,000 full-text 

PDF documents are added each year. 

To be sure, most publications are 

not for the lay reader. A quick glance 

turns up titles such as “Method and 

apparatus for the nondestructive 

assay of bulk nuclear fuel using 1 keV 

to Mev range neutrons” or “Regional 

comparison of nuclear and fossil electric 

power generation costs”. Many 

were once only available in hard copy 

but are now online after a massive 

effort to convert millions of microfiche 

pages to fully searchable electronic 

files. 

Improving INIS inputs 

To continually improve the quantity 

and quality of submissions to the INIS 

repository, the IAEA periodically 

offers trainings and e-learning courses 

to specialists, librarians and scientific 

officers involved in knowledge 

management. Last month, experts 

gathered in Vienna to discuss different 

aspects of INIS operations, such as 

News


560 

NEWS 

selection criteria and descriptive 

cataloguing, and improve their skills 

for the preparation of high-quality 

input and use of the repository. The 

participants, more than half of them 

women, came from Africa, Asia, Europe 

and the Americas. 

Full-text documents available in 

the INIS repository represent almost 

entirely nuclear-related non-conventional 

literature. Non-conventional 

literature includes any literature 

which is not normally available 

through commercial distribution 

channels and which is generally 

difficult to locate. The depth and 

breadth of the INIS repository and 

the large number of daily users 

demonstrate how a well-planned and 

implemented international cooperation 

project can make a significant 

contribution to open science. 

| www.iaea.org 

Company News 

Framatome successfully 

implements innovative 

maintenance technique 

on reactor vessel 

component underwater 

(framatome) Framatome applied a 

cutting-edge maintenance technique 

on reactor vessel primary nozzles at 

Dominion Energy’s Millstone Power 

Station during the plant’s spring 2019 

outage. This was the first application 

of Framatome’s ultra-high pressure 

(UHP) cavitation peening process on 

reactor pressure vessel nozzles to 

primary pipe welds. Because it is 

deployed directly to the inner surface, 

it is uniquely suited to remediate the 

component regardless of external 

space restrictions or dose constraints. 

“Framatome’s innovative solutions 

are ensuring the efficient and reliable 

operation of today’s reactor fleet,” said 

Catherine Cornand, Framatome’s 

senior executive vice president in 

charge of the Installed Base Business 

Unit. “This new underwater application 

of UHP cavitation peening on a 

primary nozzle is another example of 

our team’s expertise and dedication to 

innovation and continuous improvement 

in servicing our customers 

worldwide.” 

To prepare for the work, Framatome 

demonstrated the qualified reactor 

vessel primary nozzle cavitation 

peening technology on a full-scale 

mock-up at the company’s world-class 

Technical Training Center in 

Lynchburg, Virginia, in early 2019. 

UHP cavitation peening is designed 

to prevent primary water stress 

corrosion cracking. The process uses 

ultra-high-pressure water jets to 

generate vapor bubbles that collapse 

with enough force to create beneficial 

compression of the components’ 

surfaces. This surface compression 

improves components’ material 

properties and enhances resistance 

to corrosion and other types of 

degradation, which reduces the 

effects of aging. 

UHP cavitation peening can extend 

the life of nuclear reactor primary 

components, including the hot leg 

primary nozzles, for up to 40 additional 

years. Additionally, the process 

reduces outage time and saves money 

by eliminating the need to replace 

components or address indications 

with traditional repair methods. UHP 

cavitation peening can be used for 

several different applications in most 

reactor designs. 

“Cavitation peening is an industry 

game-changer that was recognized in 

2017 as one of the Top Innovative 

Practices for work completed on the 

Byron and Braidwood reactor vessel 

closure heads,” said Craig Ranson, 

senior vice president of Framatome’s 

North America Installed Base Business 

Unit. “We are proud to work with 

Dominion to expand our proven 

capabilities and engineer a solution 

for this unique primary nozzle repair.” 

Located in Waterford, Connecticut, 

the Millstone Power Station’s two 

pressurized water reactors produce 

enough electricity to power 2.1 million 

homes. 

| www.framatome.com 

UK Government and industry 

champion new compact 

nuclear power station 

(rolls-royce) UK Research and Innovation 

(UKRI) has confirmed it has 

provided initial match funding to the 

consortium of companies designing a 

new type of nuclear power station in 

the UK. 

| Artist's view of the SMR design of a new UK nuclear power station project. 

The initial joint investment of 

£18million from UKRI will be matched 

by nuclear, civil engineering construction 

and manufacturing industry 

firms, who have been working on the 

preliminary design for four years. 

The power station is a compact 

design, the components for which are 

manufactured in sections in regional 

UK factories, before being transported 

to existing nuclear sites for rapid 

assembly inside a weatherproof 

canopy. This cuts costs by avoiding 

weather disruptions and secures 

gradual efficiency savings by using 

streamlined and standardised manufacturing 

processes for its components. 

By 2050 a full UK programme of up 

to 16 of these power stations could 

create: 

p Up to 40,000 jobs 

p £ 52 bn of value to the UK 

economy 

p £ 250 bn of exports 

Paul Stein, Chief Technology Officer 

for Rolls-Royce, which leads the 

consortium, said: “Tackling climate 

change requires collaboration across 

industries and governments to find 

effective, affordable and sustainable 

ways of achieving net zero by 2050. 

“The consortium’s work with the 

Government shows that action is 

being taken to decarbonise our 

economy and meet our society’s vital 

and growing power needs. This is a 

very positive step forward to this next 

phase of the programme.” 

The partners in the consortium 

are Assystem, BAM Nuttall, Laing 

O’Rourke, National Nuclear Laboratory 

(NNL), Rolls-Royce, Atkins, 

Wood, The Welding Institute (TWI) 

and Nuclear AMRC. 

The target cost for each station is 

£1.8 billion by the time five have been 

built, with further savings possible. 

Each power station will be able to 

operate for 60 years and provide 

440 MW of electricity, which is 

enough to power a city the size of 

Leeds. 

News


The shared initial investment will 

be used to progress the significant 

opportunities presented by the programme; 

prepare it for the UK’s 

regulatory Generic Design Assessment 

process; and make final decisions on 

which innovations to pursue and realise. 

It will also generate the valuable 

confidence that the supply chain needs 

to begin to prepare for a programme 

that could create around £52 billion of 

value for the UK economy. 

When licensed and supported by 

the required enabling legislation and 

siting processes, the power station 

could provide reliable low carbon 

energy from the early 2030s. 

The Government’s intent to 

support the programme was announced 

in July 2019. 

| www.rolls-royce.com 

Uranium 

Prize range: Spot market [USD*/lb(US) U 3O 8] 

140.00 

120.00 

100.00 

80.00 

60.00 

40.00 

20.00 

0.00 

1980 

Yearly average prices in real USD, base: US prices (1982 to1984) * 

1985 

1990 

1995 

2000 

2005 

) 1 

2010 

2015 

2019 

Year 

* Actual nominal USD prices, not real prices referring to a base year. Year 

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019 

* Actual nominal USD prices, not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019 

| Uranium spot market prices from 1980 to 2019 and from 2008 to 2019. The price range is shown. 

In years with U.S. trade restrictions the unrestricted uranium spot market price is shown. 

Separative work: Spot market price range [USD*/kg UTA] 

Conversion: Spot conversion price range [USD*/kgU] 

180.00 

22.00 

) 1 

160.00 

140.00 

120.00 

100.00 

80.00 

60.00 

40.00 

Uranium prize range: Spot market [USD*/lb(US) U 3O 8] 

140.00 

120.00 

100.00 

80.00 

60.00 

40.00 

20.00 

0.00 

20.00 

18.00 

16.00 

14.00 

12.00 

10.00 

Jan. 2008 

8.00 

6.00 

4.00 

Jan. 2009 

Jan. 2010 

) 1 

Jan. 2011 

Jan. 2012 

Jan. 2013 

Jan. 2014 

Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

561 

NEWS 

) 1 Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019 

Westinghouse signs 

agreement to acquire Rolls- 

Royce Civil Nuclear Systems 

and Services Business 

(west) Westinghouse Electric Company 

signed a definitive agreement to 

acquire Rolls-Royce’s Civil Nuclear 

Systems and Services business in North 

America, expanding Westinghouse’s 

global capabilities in digital, engineering 

services, plant automation and 

monitoring systems, field services and 

manufacturing. 

“Creating customer value and 

supporting our customers’ operations 

is a key driver for Westinghouse. 

Acquiring Rolls-Royce will strengthen 

our ability to serve the nuclear 

operating fleet through an expanded 

presence in our core business while 

adding new digital offerings,” said 

Patrick Fragman, Westinghouse president 

and chief executive officer. “This 

acquisition is an important step in our 

growth strategy. We look forward to 

welcoming the employees of Rolls- 

Royce to Westinghouse.” 

The acquisition of Rolls-Royce will: 

p Expand Westinghouse’s operating 

plant services capabilities 

p Enhance the company’s digital 

innovation efforts to optimize 

customer planning and maintenance, 

and provide engineering 

solutions to maximize cost 

effectiveness and obsolescence risk 

p Support both Westinghouse’s and 

Rolls-Royce’s global customer base 

through an expanded presence 

and synergies between both 

companies, enhancing customer 

offerings and experience in field 

services and plant automation 

p Further enable Westinghouse’s 

growth while supporting 

20.00 

0.00 

Jan. 2008 

Jan. 2009 

Jan. 2010 


Jan. 2012 


customers in the North American 

and European nuclear markets 

Rolls-Royce operates 11 sites in 

Canada, France, the United Kingdom 

and the United States. 

| www.westinghousenuclear.com 

Market data 

(All information is supplied without 

guarantee.) 

Nuclear Fuel Supply 

Market Data 

Information in current (nominal) 

U.S.-$. No inflation adjustment of 

prices on a base year. Separative work 

data for the formerly “secondary 

market”. Uranium prices [US-$/lb 

U 3 O 8 ; 1 lb = 453.53 g; 1 lb U 3 O 8 = 

0.385 kg U]. Conversion prices [US-$/ 

kg U], Separative work [US-$/SWU 

(Separative work unit)]. 

Jan. 2013 

Jan. 2014 

Jan. 2015 

Jan. 2016 

2017 

p Uranium: 19.25–26.50 

p Conversion: 4.50–6.75 

p Separative work: 39.00–50.00 

2018 

p Uranium: 21.75–29.20 



2019 

January 2019 

p Uranium: 28.70–29.10 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019 

2.00 

0.00 


| Separative work and conversion market price ranges from 2008 to 2019. The price range is shown. 

)1 

In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps. 

* Actual nominal USD prices, not real prices referring to a base year 

Sources: Energy Intelligence, Nukem; Bilder/Figures: atw 2019 

Jan. 2008 

Jan. 2009 

Jan. 2010 


Jan. 2012 



February 2019 

p Uranium: 27.50–29.25 



March 2019 

p Uranium: 24.85–28.25 



April 2019 

p Uranium: 25.50–25.88 



May 2019 

p Uranium: 23.90–25.25 



June 2019 

p Uranium: 24.30–25.00 



July 2019 

p Uranium: 24.50–25.60 



August 2019 

p Uranium: 24.90–25.60 



September 2019 

p Uranium: 24.80–26.00 



| Source: Energy Intelligence 

www.energyintel.com 

Jan. 2013 

Jan. 2014 

Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

News


562 

NUCLEAR TODAY 

John Shepherd is a 

freelance journalist 

and communications 

consultant who has 

covered the nuclear 

industry for the past 

20 years and is 

currently editor-in-chief 

of UK-based Energy 

Storage Publishing. 

Sources 

World Nuclear 

Association 

https://cutt.ly/weEfVJC 

IEA Sweden report 

https://cutt.ly/TeEfZIs 

Cryospheric Sciences 

Division blog 

https://cutt.ly/1eEfLyD 

Taking a Leaf out of Greta’s 

Climate Change Book 

The media loves a ‘warrior underdog’ and so journalists around the world have largely fallen in love with Swedish eco 

champion Greta Thunberg. The diminutive teenager has captured the imagination of a new generation of environmental 

activists since she burst onto the international scene around a year ago, when she protested on school days outside the 

Swedish parliament. She said her school strike was to draw attention to global warming. 

Her actions caught on and soon led to other students 

making similar protests. The rest, to coin a phrase, is 

already modern history. Thunberg has been feted worldwide 

and she even addressed the 2018 United Nations 

Climate Change Conference. 

Whatever one thinks of Thunberg’s approach, it’s hard 

to deny that she has reignited interest in the climate change 

debate and raised the political temperature on all sides. 

So could nuclear’s proponents take a leaf out of the 

young environmentalist’s book? Earlier this year, a group 

of reactor physicists, operators and politicians were among 

those who followed Thunberg’s lead and gathered outside 

the Swedish parliament to raise awareness of the benefits 

of nuclear energy. 

One of the organisers of the ‘Stand up for Nuclear’ event 

was Swedish Nuclear Society president Marcus Eriksson. 

He told me that, on the same day of the event, the group 

published an opinion article in one of Sweden’s major 

newspapers, Svenska Dagbladet, which explained the 

country’s energy situation in Sweden, making a parallel to 

Germany’s energy sector. 

The Stand up for Nuclear event is an important development 

in Sweden and one that could and should have 

wider implications for Europe. Sweden has a chequered 

past where nuclear is concerned. 

According to the World Nuclear Association, Sweden 

generated a combined 156 terawatt hours (TWh) of 

electricity in 2016, of which 63 TWh (40%) was from its 

eight nuclear power reactors and 62 TWh (40%) from 

hydro. Wind provided 15.5 TWh (10%), various fossil fuels 

2 TWh, and biofuels and waste 13 TWh (8%). Total 

installed generating capacity at the end of 2016, as 

recorded by the International Energy Agency‘s (IEA) 

‘ Electricity Information 2018’, was 40 gigawatts electric. 

In 1980, the Swedish government decided to phase out 

nuclear power but this policy was repealed by lawmakers 

in June 2010. The country‘s 1997 energy policy had 

allowed 10 reactor units to operate longer than envisaged 

by the 1980 phase-out policy. However, the policy also led 

to the premature closure of a two-unit (1,200 megawattelectric) 

plant, although some 1,600 MWe was subsequently 

added in uprates to the remaining 10 units. 

In 2015, decisions were made to close four older 

reactors by 2020. A year later, a tax discriminating against 

nuclear power was abolished. 

Nuclear’s impact on Sweden’s environmental credentials 

is impressive. The country has almost fully decarbonised its 

electricity generation, which the IEA has described as “a 

feat which is quite unique among the IEA member countries”. 

In a review of Swedish energy policy published last 

April, the IEA said the country should assess the contribution 

of nuclear power over the next 20 years and the impact 

of further potential early closures on energy security. 

I have no scientific basis for saying this, only a gut 

feeling (from reporting on nuclear and the wider energy 

industry over the past 20 years), that there is often a silent 

majority who would speak up in favour of nuclear – 

provided they are given the facts. 

It appears such a majority could exist in Sweden, if a 

survey conducted last summer by Novus Opinion is 

anything to go by. That survey indicated a majority of 

Swedes now believe that nuclear power could be a means 

to tackle the climate crisis. Which brings me back to the 

Stand up for Nuclear event organised in Stockholm. 

Marcus Eriksson told journalists covering the event: 

“Two years ago none of the political parties wanted to talk 

about nuclear power, now everyone is talking about it. It 

reflects a stronger opinion that the technology has an 

important role to play to combat climate change.” 

The Swedish Nuclear Society should be praised for 

raising awareness of the benefits of nuclear in this way. 

Every day, when we are out and about, there is a manifestation 

of some sort or another with campaigners competing 

for our attention on issues. So why not something 

that speaks up for nuclear? In these gloomy winter nights, 

perhaps there is an inbuilt advantage for the nuclear cause 

in pointing out just how dark and cold many of our communities 

would be without the benefit of atomic power. 

As we head into a new year, it’s a good time to start 

thinking about the resolutions we will make for the months 

ahead. In addition to resolution favourites, such as going 

on a diet, or giving up smoking, those involved in the 

nuclear sector should pledge to make a special effort to 

stand up for nuclear in 2020. 

The initial unassuming campaign launched by Greta 

Thunberg just a year ago is, I suggest, a template for what 

could be a ‘2020 Vision for Nuclear’ campaign for the year 

ahead. Perhaps with the advent of social media we’ve 

forgotten about the impact of ‘grassroots’ campaigning. 

Social media is of course tremendously valuable in 

reaching millions. But let’s not forget the impact that can 

be had by being out on the streets and chatting to passersby 

about the technology that keeps the lights on and can 

help clean up our local and global environments. 

There’s no time like the present to get out there and 

follow Greta’s footprint. In fact, the beloved Santa Clause, 

a figure who returns to prominence at this time of the year, 

could be recruited to underline the importance of the 

nuclear cause. 

As the Cryospheric Sciences Division of the European 

Geosciences Union has pointed out, Santa might well have 

to think about moving from his fabled home of Lapland 

because of global warming and polar amplification. In the 

absence of snow, Santa does have the possibility of 

converting his sleigh to an all-electric model. But even if 

there were enough charging points around the globe, they 

would need reliable supplies of clean electricity, which 

brings us back to nuclear’s importance for our planet’s 

energy mix. 

In wrapping up this festive article, it only remains to 

wish everyone the compliments of the season and, if not a 

white Christmas, at least an increasingly green one. 

Nuclear Today 

Taking a Leaf out of Greta’s Climate Change Book ı John Shepherd

Kommunikation und 

Training für Kerntechnik 

Strahlenschutz – Aktuell 

In Kooperation mit 

TÜV SÜD Energietechnik GmbH 

Baden-Württemberg 

Seminar: 

Das neue Strahlenschutzgesetz – 

Folgen für Recht und Praxis 

Seminarinhalte 

1. Teil | Das neue Strahlenschutzgesetz (StrlSchG) 

ı Das neue StrlSchG: Historie 

ı Inkrafttreten des StrlSchG 

ı Überblick über grundlegende Änderungen 

ı Die Entwürfe der neuen Strahlenschutzverordnung(en) (soweit zum Seminarzeitpunkt vorliegend) 

2. Teil | Auswirkungen auf die betriebliche Praxis 

ı Genehmigungen, Zuständigkeiten 

ı Begriff der Expositionssituation (geplant, bestehend, Notfall), NORM 

ı Aufsichtsprogramm, § 180 StrlSchG, Rechtfertigung 

ı Notfallpläne 

ı Änderungen für SSV/SSB 

ı Dosisgrenzwerte 

ı Strahlenschutzregister 

3. Teil | Strahlenschutz im Back End 

ı Freigabe und Entsorgung 

ı Altlasten 

ı Baustoffe 

Zielgruppe 

Die 2-tägige Schulung wendet sich an Fach- und Führungskräfte, an Projekt- und Abteilungsleiter und 

Experten aus den Bereichen Betrieb, Abfälle, Genehmigung, Strategie und Unternehmens kommunikation 

sowie an Juristen. 

Referenten 

Dr. Maria Poetsch 

Dr. Christian Raetzke 

ı Strahlenschutzexpertin bei der TÜV SÜD Energietechnik GmbH 

Baden-Württemberg 

ı Rechtsanwalt, Leipzig 

Wir freuen uns auf Ihre Teilnahme! 

Bei Fragen zur Anmeldung rufen Sie uns bitte an oder senden uns eine E-Mail. 

Termine 

2 Tage 

28. bis 29. Januar 2020 

17. bis 18. März 2020 

16. bis 17. Juni 2020 

Tag 1: 10:30 bis 17:30 Uhr 

Tag 2: 09:00 bis 16:30 Uhr 

Berlin 

Teilnahmegebühr 

1.598,– € ı zzgl. 19 % USt. 

Im Preis inbegriffen sind: 

ı Seminarunterlagen 

ı Teilnahmebescheinigung 

ı Pausenverpflegung 

inkl. Mittagessen 

Kontakt 

INFORUM 

Verlags- und Verwaltungsgesellschaft 

mbH 



Petra Dinter-Tumtzak 

Fon +49 30 498555-30 

Fax +49 30 498555-18 

Seminare@KernD.de

#51KT 

www.kerntechnik.com 

Medienpartner 

51. KERNTECHNIK 

2020 

Jetzt anmelden! 

www.kerntechnik.com/kerntechnik/registration/ 

5. – 6. Mai 2020 

Estrel Convention Center 

Berlin

atw - International Journal for Nuclear Power | 11/12.2019

You also want an ePaper? Increase the reach of your titles

Delete template?

Save as template?