atw - International Journal for Nuclear Power | 05.2020

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2020 

5 

ISSN · 1431-5254 

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The European Nuclear 

Experimental Educational 

Platform (ENEEP) for 

Education and Training 

Physical and Chemical Effects 

of Containment Debris on the 

Emergency Coolant Recirculation 

Safety Case Considerations 

for the Use of Robots 

in Nuclear Decommissioning

Competence for 

Nuclear Services 

Operational Waste and D&D 

Spent Fuel Management 

Nuclear Casks 

Calculation Services and Consulting 

Waste Processing Systems and Engineering 

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atw Vol. 65 (2020) | Issue 5 ı May 

Corona Epidemic. Nuclear Power. 

Dear readers, The worldwide spread of the Corona virus and the directly related effects are currently determining 

and changing our lives, and will probably continue to do so for a long time to come. The Covid-19 pandemic has triggered 

an unprecedented global health and economic crisis. The energy sector as a whole is also affected by this crisis. 

But this crisis also shows once again that the energy sector 

and the reliable and secure supply of energy is not only of 

central importance for our society and our lives, but is also 

the domino at the beginning of the chain for our entire 

infrastructure and is therefore part of the “critical”, 

perhaps better “essential” infrastructure. And the power 

supply – as a pillar of the important possibility of 

maintaining more or less social contacts at all in these 

days, via telephone or the manifold possibilities of the 

internet – receives special attention. To maintain its 

stability 3600s/24h/7d is a demanding challenge for 

people and technology. Employees in nuclear power plants 

and other energy companies around the world also 

contribute to mastering this challenge. 

The measures taken by governments in response to the 

Covid-19 crisis have led to a significant reduction in 

expected total electricity consumption in the range of 

10 to 25 %, particularly due to restrictions on economic 

activity in some countries. In contrast, many countries are 

experiencing a noticeable increase in this share of supply 

of up to 5 %, largely due to “home office work” in the 

private sector. Nuclear energy contributes to about 10.5 % 

of global electricity generation in over 30 countries. It is 

part of the electricity supply and has high capacity factors 

and availabilities as well as a high degree of flexibility, in 

particular to support intermittent generation and to 

technically enable its integration into the energy system. 

This results in special precautionary measures to protect 

the employees and to technically maintain operations. In 

the USA, the measures were also extended to the nuclear 

supply chain, i.e. nuclear fuel supply and service personnel. 

The nuclear industry worldwide has taken precautions 

for this eventuality. As part of their special safety culture 

based on foresight and forward planning, plans were 

already in place before the current crisis to ensure the best 

possible protection. The reports from the Wuhan outbreak 

in China at the beginning of this year were followed with 

particular attention by the nuclear industry. The industry 

was thus able to initiate the existing plans to maintain 

operations and protect all employees in a forward-looking 

manner. The measures are manifold and serve, for 

example, to minimize the risk of infection: Home office is 

the keyword for the area of administration, communication 

is conducted via electronic channels wherever 

possible, two examples for many measures in a coherent 

overall concept that is confirmed at nuclear sites 

worldwide. In addition, provisions have been made at 

plant locations to ensure that operations can continue to 

be run autonomously where necessary, i.e. to be able to 

provide support and care for employees on site as far as 

possible. 

With regard to technology, the operation of nuclear 

power plants was also adapted to the current challenges, 

i.e. revision plans were adjusted and also the operational 

management, in order to be able to contribute to the power 

supply with sufficient "residual criticality" of the nuclear 

fuel at a later point in time, if necessary. 

In the field of nuclear infrastructure, some companies 

in the nuclear fuel supply industry have decided to 

temporarily suspend their mining activities in order to 

protect their employees. Operators of new construction 

projects have made the same decisions and in some cases, 

as in the case of the foreign projects of the Russian company 

Rosatom, have withdrawn their personnel from the 

construction sites and brought them back home. The 

extent to which these measures will affect the planned 

start-ups cannot currently be estimated, but it is not urgent 

either – protection has priority. 

However, nuclear technology is also directly involved in 

coping with and combating corona infection and 

pandemic. The International Atomic Energy Agency 

(IAEA) provides such equipment to particularly affected 

countries with few facilities and equipment and supports 

the training of nuclear diagnostic specialists. This 

particularly concerns countries in Africa. Industrial 

irradiation facilities are now being used primarily for 

sterilisation and disinfection of medical equipment, and 

specialised nuclear medicine and basic nuclear research 

laboratories are working with other faculties on the 

search for and development of active substances against 

Covid-19 disease and possible vaccines. 

With expertise and the commitment of its employees 

worldwide, the nuclear sector plays its part in keeping our 

society and our lives functioning. We would like to thank 

you, as well as all the others who have to bear special 

burdens and who show extraordinary commitment! 

Christopher Weßelmann 

– Editor in Chief – 

243 

EDITORIAL 

Editorial 

Corona Epidemic. Nuclear Power.


244 

EDITORIAL 

Corona-Epidemie. Kernenergie. 

Liebe Leserinnen, liebe Leser, die weltweite Verbreitung des Corona-Virus und die damit direkt verbundenen 

Auswirkungen bestimmen und verändern aktuell unser Leben und dies voraussichtlich noch für eine lange Zeit. 

Die Covid-19-Pandemie hat eine beispiellose globale Gesundheits- und Wirtschaftskrise ausgelöst. Auch der 

Energiesektor in Gänze ist von dieser Krise betroffen. 

Diese Krise zeigt aber auch erneut, dass der Energiesektor 

und die zuverlässige und sichere Energieversorgung nicht 

nur von zentraler Bedeutung für unsere Gesellschaft und 

unser Leben ist, sondern der Dominostein am Beginn der 

Kette für unsere gesamte Infrastruktur und deshalb zur 

„kritischen“ vielleicht besser „essenziellen“ Infrastruktur 

zählt. Und der Stromversorgung – als eine Säule der in 

diesen Tagen wichtigen Möglichkeit mehr oder minder 

soziale Kontakte überhaupt noch aufrecht zu erhalten, 

über Telefon oder die vielfältigen Möglichkeiten des 

Internets – kommt ein besonderes Augenmerk zu. Ihre 

Stabilität 3600s/24h/7d aufrecht zu erhalten, ist eine 

anspruchsvolle Herausforderung an Menschen und 

Technik. Auch die Beschäftigen in den Kernkraftwerken 

und den weiteren Energieunternehmen weltweit tragen 

zur Bewältigung dieser bei. 

Die von den Regierungen ergriffenen Maßnahmen 

als Antwort auf die Covid-19-Krise haben besonders 

durch Einschränkungen des Wirtschaftslebens in einigen 

Ländern zu einem deutlichen Rückgang des erwarteten 

Gesamt-Stromverbrauchs im Bereich von 10 bis 25 % 

geführt. Im Gegensatz dazu wird in vielen Ländern 

wesentlich aufgrund der „Home-Office-Arbeit“ im privaten 

Bereich ein erkennbarer Anstieg dieses Anteils der 

Versorgung von bis zu 5 % verzeichnet. Die Kernenergie 

trägt weltweit zu etwa 10,5 % der Stromerzeugung in über 

30 Ländern bei. Sie ist Teil der Stromversorgung und von 

hohen Kapazitätsfaktoren und Verfügbarkeiten sowie 

einem hohen Maß an Flexibilität gekennzeichnet, um 

insbesondere intermittierende Erzeugung zu stützen und 

ihre Einbindung in das Energiesystem technisch zu 

ermöglichen. Daraus folgen besondere Vorsorgemaßnahmen 

zum Schutz der Mitarbeitenden und zur 

technischen Aufrechterhaltung des Betriebs. In den USA 

wurden die Maßnahmen auch auf die nukleare Lieferkette, 

d.h. die Kernbrennstoff versorgung sowie das Servicepersonal 

ausgeweitet. 

Die kerntechnische Industrie weltweit hat für diesen 

Fall vorgesorgt. Als Teil ihrer besonderen auf Weitsicht 

und Vorausplanung ausgestalteten Sicherheitskultur lagen 

schon vor der aktuellen Krise ausgearbeitete Pläne vor, um 

bestmöglichen Schutz zu gewährleisten. Die Meldungen 

Anfang dieses Jahres aus dem Ausbruchsgebiet Wuhan in 

China waren in der kerntechnischen Industrie mit 

besonderer Aufmerksamkeit verfolgt worden. So war die 

Industrie in der Lage, die vorliegenden Pläne zur Aufrechterhaltung 

des Betriebs und zum Schutz aller Beschäftigten 

vorausschauend einzuleiten. Die Maßnahmen sind vielfältig 

und dienen zum Beispiel dazu, Infektionsrisiken zu 

minimieren: Homeoffice ist das Stichwort für den Bereich 

der Verwaltung, Kommunikation wird da, wo möglich, 

über elektronische Kanäle geführt, zwei Beispiel für viele 

einzelne in einem stimmigen Gesamtkonzept, das sich an 

den kerntechnischen Standorten weltweit bestätigt. 

Darüber hinaus wurde an Anlagenstandorten Vorsorge 

dafür getroffen, den Betrieb gegebenenfalls autark weiter 

zu führen, d.h. die Beschäftigten soweit wie möglich vor 

Ort betreuen und versorgen zu können. 

Mit Blick auf die Technik wurde der Betrieb von 

Kernkraftwerken zudem auf die aktuellen Herausforderungen 

angepasst, d.h. Revisionspläne wurden 

angepasst und ebenso das Betriebsmanagement, um 

gegebenenfalls mit ausreichender „Restkritikalität“ des 

Kernbrennstoffs zu späteren Zeitpunkten zur Stromversorgung 

beitragen zu können. 

Im Bereich der nuklearen Infrastruktur haben sich 

Unternehmen der Kernbrennstoffversorgung teils dafür 

entschieden ihre Bergbauaktivitäten vorläufig einzustellen, 

um auch hier die Beschäftigten zu schützen. 

Gleiche Entscheidungen haben die Betreiber von Neubauprojekten 

getroffen und ihr Personal teilweise, wie z.B. bei 

den Auslandsprojekten des russischen Unternehmens 

Rosatom, von den Baustellen abgezogen und es nach 

Hause zurückgeholt. Inwieweit sich diese Maßnahmen 

auf die geplanten Inbetriebnahmen auswirken wird, lässt 

nicht aktuell nicht abschätzen, ist aber auch nicht vordringlich 

– Schutz hat Vorrang. 

Die Nukleartechnik ist aber auch direkt an der 

Bewältigung und Bekämpfung der Coronainfektion und 

-pandemie beteiligt. Die Internationale Atomenergie- 

Organisation (IAEO) stellt besonders betroffenen Ländern 

mit wenigen Möglichkeiten und Ausrüstung solche zur 

Verfügung und unterstützt bei der Ausbildung von 

Fach personal für die Nukleardiagnostik. Vor allem betrifft 

dies Länder Afrikas. Industrielle Bestrahlungsanlagen 

werden jetzt vordringlich zur Sterilisation und 

Desinfektion medizinischer Ausrüstung eingesetzt und in 

den spezialisierten Labors der Nuklearmedizin und 

nuklearen Grundlagenforschung wird gemeinsam mit 

anderen Fakultäten an der Suche und Entwicklung 

von Wirkstoffen gegen die Covid-19-Erkrankung und möglichen 

Impfstoffen gearbeitet. 

Der Nuklearsektor trägt mit seiner Expertise und dem 

Engagement seiner Mitarbeiterinnen und Mitarbeiter 

weltweit seinen Teil dazu bei, unsere Gesellschaft und 

unser Leben funktionsfähig zu halten. Auch Ihnen ist, wie 

allen anderen, die besondere Lasten zu tragen haben und 

sich außergewöhnlich engagieren, zu danken! 

Christopher Weßelmann 

– Chefredakteur – 

Editorial 

Corona Epidemic. Nuclear Power.

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 25.06.2020 Berlin 

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

Export kerntechnischer Produkte und Dienstleistungen – 

Chanchen und Regularien 

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

Atomrecht – Was Sie wissen müssen 

3 Kommunikation und Politik 

RA Dr. Christian Raetzke 

Akos Frank LL. M. 

11.11.2020 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: 

3 Nuclear English 

Das Strahlenschutzrecht und 

seine praktische Umsetzung 

Stilllegung und Rückbau in Recht und Praxis 

Dr. Maria Poetsch 


Dr. Stefan Kirsch 


16.06. - 17.06.2020 

29.10. - 30.10.2020 

Berlin 

23.09. - 24.09.2020 Berlin 

English for the Nuclear Industry Angela Lloyd 07.10. - 08.10.2020 Berlin 

3 Wissenstransfer und Veränderungsmanagement 

Erfolgreicher Wissenstransfer in der Kerntechnik – 

Methoden und praktische Anwendung 

Veränderungsprozesse gestalten – 

Herausforderungen meistern, Beteiligte gewinnen 

Dr. Tanja-Vera Herking 

Dr. Christien Zedler 

Dr. Tanja-Vera Herking 

Dr. Christien Zedler 

05.10. - 06.10.2020 Berlin 

24.11. - 25.11.2020 Berlin 

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

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.


246 

Issue 5 | 2020 

May 

CONTENTS 

Contents 

Editorial 

Corona Epidemic. Nuclear Power E/G . . . . . . . . . . . . . . . . . . 243 

Inside Nuclear with NucNet 

Foratom Interview: Why Europe Needs to Include Nuclear 

in Low-Carbon Energy Planning . . . . . . . . . . . . . . . . . . . . . 248 

Calendar 250 

Feature | Research and Innovation 

The European Nuclear Experimental Educational 

Platform (ENEEP) for Education and Training . . . . . . . . . . . . . 251 

Did you know...? 257 

Spotlight on Nuclear Law 

Atomic Law – Changes Over Time G . . . . . . . . . . . . . . . . . . . 258 

Research and Innovation 

BER II – The End of an Era . . . . . . . . . . . . . . . . . . . . . . . . . 259 

On the Scientific Utilisation of Low Power Research Reactors . . . 262 

The Performance of Low Activation Steel SCRAM on ACPs 

Source Term in Water- cooled Loop of Fusion Reactor ITER. . . . .268 

Fluid Structure Interaction Analysis of a Surge-line 

Using Coupled CFD-FEM . . . . . . . . . . . . . . . . . . . . . . . . . . 272 

Environment and Safety 

Physical and Chemical Effects of Containment Debris 

on the Emergency Coolant Recirculation . . . . . . . . . . . . . . . .276 

Experimental and Computational Analysis 

of a Passive Containment Cooling System 

with Closed-loop Heat Pipe Technology . . . . . . . . . . . . . . . . 280 

Decommissioning and Waste Management 

Safety Case Considerations for the Use of Robots 

in Nuclear Decommissioning . . . . . . . . . . . . . . . . . . . . . . . 287 

KTG Inside 292 

Cover: 

BER II Research Reactor 

Courtesy of Bernhard Ludewig 

G 

E/G 

= German 

= English/German 

News 292 

Nuclear Today 

Energy Providers Deserve Our Gratitude Now More Than Ever . . 298 

Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 278 

Contents


247 

Feature 


251 The European Nuclear Experimental Educational 

Platform (ENEEP) for Education and Training 

CONTENTS 

Marcella Cagnazzo, Helmuth Boeck, 

Štefan Čerba, Szabolcs Czifrus, Jan Haščík, Anže Jazbec, 

Jakub Lüley, Marcel Miglierini, Filip Osuský, 

Vladimir Radulović, Fabian Schaden, Lubomir Sklenka, 

Luka Snoj, Attila Tormási, Mario Villa, Branislav Vrban 


259 BER II – The End of an Era 

Helmholtz-Zentrum Berlin für Materialien und Energie 

262 On the Scientific Utilisation of Low Power Research Reactors 

Pavol Mikula and Pavel Strunz 


276 Physical and Chemical Effects of Containment Debris 

on the Emergency Coolant Recirculation 

Jisu Kim and Jong Woon Park 


287 Safety Case Considerations 

for the Use of Robots in Nuclear Decommissioning 

Howard Chapman, John-Patrick Richardson, 

Colin Fairbairn, Darren Potter, Stephen Shackleford and Jon Nolan 

Contents


248 

Foratom Interview: Why Europe Needs to Include Nuclear 

in Low-Carbon Energy Planning 

Industry group head Yves Desbazeille says if EU is serious about tackling climate change, 

it must make use of ‘all the best low-carbon tools’ 

INSIDE NUCLEAR WITH NUCNET 

Q: You told a press briefing in Brussels 

that 2020 will be a crucial year for the nuclear industry 

in Europe. Why is that? 

Last year ended with a few very important developments 

which will impact the future of nuclear energy. The 

European Commission’s proposal for a European Green 

Deal maintains the principle that EU member states are 

free to choose their own energy mix. And in its resolution 

ahead of the COP25 conference in Madrid, the European 

Parliament recognised the role of nuclear in fighting 

climate change. 

Also, an official memorandum following the European 

Council’s (EUCO) December summit on climate change 

mentions nuclear energy as a tool to achieve climate 

neutrality. What is more, the recently agreed classification 

system for sustainable economic activities, known as the 

taxonomy, does not exclude nuclear. The trend for 

including nuclear in future energy options was also seen at 

the end of last year when several EU member states, 

including the Czech Republic, Hungary and Poland, made 

it clear that to commit to 2050 decarbonisation targets 

they must be allowed to invest in nuclear power. 

There are signals at EU level that nuclear may not be 

treated equally with other low-carbon energy sources. 

When discussing the bloc’s future energy mix, EU decisionmakers 

tend to focus only on renewables and energy 

efficiency. So the question for the next 12 months is how 

recent positive signals will be translated into specific EU 

legislation and to what extent EU decision makers will 

recognise nuclear energy for the benefits it brings to the 

system. If the European Union is serious about tackling 

climate change, then EU decision-makers must act urgently 

and make use of all the best low-carbon tools, including 

nuclear. Only by combining renewables with nuclear 

energy can we deliver on our commitments. 

Q: The EU excluded nuclear energy from funding in its 

recent European Green Deal policy initiative, a move 

Foratom has criticised. What was the impact of the 

decision? 

The European Green Deal maintains the principle of 

leaving EU member states free to choose their own energy 

mix, including nuclear energy. Foratom supports this 

approach and welcomes the commission’s goal of becoming 

more ambitious in reducing its CO 2 emissions whilst at the 

same time ensuring that no EU citizen is left behind in the 

transition, as long as it allows member states to choose 

their own methods of decarbonisation. Expecting them to 

reduce their greenhouse gas emissions whilst preventing 

them from investing in specific low-carbon technologies 

such as nuclear would be counter-productive. 

What concerns us is the fact the commission has 

decided to exclude nuclear energy, both new build and 

decommissioning, from having access to the Just Transition 

Fund, which is one of three main sources of financing 

the Just Transition Mechanism – the EC’s key tool to 

provide member states with targeted financial support 

for their transition to low-carbon energy. We regret that 

the commission didn’t include nuclear energy in the fund. 

It’s hard for us to see the justification for this decision 

because the EU should be focusing on helping people in 

carbon- intensive regions transition into all low-carbon 

industries. 

It’s important to emphasise that nuclear hasn’t been 

excluded from the whole Just Transition Mechanism. For 

example, the European Investment Bank’s updated loan 

policy, which will be one of the sources for financing the 

‘just transition’, keeps nuclear on the list of potential 

projects that can receive funding. 

The proposals presented by the European Commission 

will now go through the legislative procedure, which 

means they could change. The commission recently 

launched a public consultation focusing on the Just 

Transition Fund. Foratom, as the voice of the European 

nuclear industry, will participate to show that nuclear 

energy should be included in the fund. 

Q: What is your view of the ‘do no significant harm’ 

policy in the commission’s taxonomy proposals? 

We want the commission to adopt a technology neutral 

and fact-based approach when it assesses energy 

technologies using this principle. The ’do no significant 

harm‘ assessment – which will enable a decision on 

whether nuclear or any other technology is eligible for 

sustainable finance or not – should be undertaken by 

experts with a strong knowledge of the nuclear life cycle. 

Foratom is confident that such a thorough and fact-based 

approach, which will evaluate selected energy sources 

using criteria like CO 2 emissions, volume and traceability 

of waste, raw material consumption and land use, will lead 

to the recognition of nuclear energy as a sustainable source 

of energy that contributes significantly to climate change 

mitigation. The same criteria should be applied equally to 

all power producing technologies. 

Q: European new-build projects have reported cost 

increases in 2019 and many industry officials have 

complained about the loss of nuclear-related industrial 

expertise in Europe. How big of a challenge is this for 

the industry? 

The nuclear industry is aware of the challenges it faces. 

Avoiding further delays in construction scheduling and 

cost increases are among them. Unfortunately, such issues 

in major construction projects, in the nuclear or in any 

other sector, are relatively common and always difficult to 

predict. That said, we believe that lessons learned from 

construction sites will enable better planning in future 

while taking into consideration the particularities of 

different projects in different countries. 

The lack of new investments in nuclear and the current 

perception of nuclear energy in the EU have definitely an 

impact on the will of young people to pursue a nuclear 

career. This is a significant challenge for us as the nuclear 

industry needs a new generation of employees. The people 

who were involved in building the first generation of 

nuclear plants, for example in France in the 1980s, are on 

the point of retiring and we will need new employees to 

replace them. 

The European nuclear industry is already undertaking 

several actions to address this challenge. One example is 

the ENEN+ project, which is funded through Horizon 

2020. The goal of this project is to attract more young 

people to a career in the nuclear sector. Unfortunately, 

more needs to be done, and not just in terms of attracting 

people into the nuclear industry, but also into science, 

technology, engineering, and mathematics subjects in 

general. We hope that the EU will also put some effort and 


Foratom Interview: Why Europe Needs to Include Nuclear in Low-Carbon Energy Planning


will work closely with the industry to ensure generation 

transition and competence transfer, as well as it will help 

the workforce adapt to new technologies. 

Q: The EU is working on a comprehensive industrial 

strategy, which aims to make European industry more 

competitive and help sustainable growth. Where is the 

place of nuclear in this? 

The nuclear industry will strive to prove that we are able 

to fit into it by showing what we have to offer and proving 

that the nuclear industry is capable of playing its part in 

the development of the European economy. The European 

nuclear industry has a lot to offer. Maintaining jobs and 

growth are among Europe’s priorities and for this it will 

need to maintain a strong industrial base with a significant 

EU-based value. Increased globalisation means Europe’s 

industries are facing strong competition from other parts 

of the world, which is in part due to higher energy costs. 

Q: Do you expect anything specific from the EC 

proposal? 

Simply to be considered part of low-carbon energy 

sources by the commission in its strategy and subsequent 

policy proposals would be enough for us. What we must 

absolutely avoid is to be explicitly excluded from these 

developments as was the case for the Just Transition Fund. 

A leaked version of the EU’s industrial strategy said Europe 

needed affordable low-carbon energy for its industry and 

to maintain competitiveness. This is what we expect to see 

nuclear be part of in the proposal. We are not asking for 

any special treatment, but rather for a level playing field 

for all low-carbon sources. 

Q: What could the nuclear industry still do to improve 

its capabilities and project record? 

In 2019, senior representatives from across the nuclear 

industry outlined – in their joint manifesto – what needs to 

be done to achieve a decarbonised Europe by 2050, whilst 

at the same time maintaining growth and jobs. The 

industry needs to deliver the required volume of nuclear 

capacity on time and at a competitive cost. To achieve that, 

the industry is working closely with the supply chain to 

maximise the benefits of replicating new build projects. 

In the manifesto, the industry underlined the 

importance of investing in and maintaining human capital. 

There is a need to work closely with national and local 

governments and other stakeholders to make the industry 

more attractive to young people and to ensure it has the 

highly skilled workforce it needs. We need to avoid any 

potential workforce gap. 

In the context of the future European industrial strategy, 

nuclear is capable of providing stable low-carbon electricity 

– compared with renewables – at an affordable cost. 

Furthermore, many industries are energy intensive and 

will need to find solutions which can help them decarbonise 

their manufacturing processes. Otherwise, Europe will 

run the risk of losing its industries due to so-called “carbon 

leakage”. Nuclear has a role to play in supporting these 

industries and helping them to remain in Europe. 

Q: Supply chain problems have been a major headache 

for nuclear new-build projects in Europe and North 

America. Is the industry working to improve the 

efficiency of the supply chain? 

We are fully supporting the optimisation of the supply 

chain. Later this year, Foratom’s Supply Chain Optimisation 

Working Group will publish a report that will include 

recommendations on what should be done to enable 

the continuous development of safety and reliability of the 

EU nuclear fleet. We want to work more closely with regulators 

to promote the better alignment of licensing and regulatory 

processes and contribute to more harmonisation 

across the European nuclear sector. 

Many of our member organisations comprise themselves 

many companies from the European supply chain, both 

locally and at an international level. They are all aware of 

the challenges and want to work to improve the efficiency of 

the sector. For example, some of our Nordic members have 

been pushing for the development of standard rules to allow 

for “off the shelf” procurement of com ponents coming from 

other industries but with applications in the nuclear sector 

as well. Here a close coordination and open discussion with 

national regulators and industrial authorities is very 

important. It is a complex and very technical matter which 

requires careful attention and largely depends on the 

individual type of equipment or components in question. 

Our supply chain report will be a step in the right 

direction. We will put forward several high-level recommendations 

and communicate them to the European 

Commission and all stakeholders. There are not going to be 

quick results overnight. Harmonisation of standards in the 

industry is going to be a lengthy, but invaluable process. 

Of course, to make sure that our capabilities match the 

EU’s targets, we should not forget about supporting 

innovation and research and development. In this respect, 

more funding for research into both current and future 

nuclear technologies such as SMRs and using nuclear to 

produce heat and hydrogen must be made available by 

Europe’s leadership 

Q: Brexit means the EU will lose one of its biggest 

nuclear power operating member states. What is the 

impact for the industry? 

Nuclear energy’s perception in Europe varies across 

different member states. At EU level we are seeing a fragile 

balance of power between countries which support nuclear 

energy and those which don’t. Countries including 

Bulgaria, the Czech Republic, Finland, France, Romania 

and Sweden see nuclear energy as essential to their energy 

mix. Others have taken the decision not to have any nuclear 

or to phase it out. In more extreme cases, some countries – 

Austria in particular – are fighting against the use of nuclear 

power in member states other than their own, making use 

of all possible legal and political means. 

The UK is pro-nuclear and its absence will definitely have 

an impact on nuclear energy’s perception in the EU. That 

said, in many countries the tide towards nuclear may be 

turning. We have countries – without nuclear energy so far 

– that are seriously considering investing in new build, such 

as Poland and Estonia. Recently, several member states 

made their commitment to more ambitious CO 2 reduction 

targets conditional on being able to invest in new nuclear 

capacity. Also, the European Council’s memorandum 

following the latest EUCO includes nuclear energy as a tool 

used by some member states to achieve climate neutrality. 

This trend shows that more and more EU member states 

consider nuclear energy an important tool in counteracting 

climate change and see a bright future for it in Europe. 

The German government’s decision to phase out 

nuclear power can be perceived by other EU member states 

in some way as a ‘lesson learnt.’ Germany is one of the most 

anti-nuclear countries in the EU and its decision to prematurely 

phase out its nuclear fleet means it will miss its 

2020 emissions targets by a wide margin. If Germany had 

decided in 2011 to phase out 20 GW of coal capacity instead 

of nuclear, it would have reached its emissions targets 

and would now be rightly recognised as the European 

climate champion. 

Author 

NucNet – The Independent Global Nuclear News Agency 

Avenue des Arts 56 2/C 

1000 Bruxelles, Belgium 

www.nucnet.org 

INSIDE NUCLEAR WITH NUCNET 249 


Foratom Interview: Why Europe Needs to Include Nuclear in Low-Carbon Energy Planning


250 

Calendar 

2020 

This is not a full list. 

Dates are subject to change. Please check the listed websites for updates. 

CALENDAR 

Cancelled 05.05. – 06.05.2020 

KERNTECHNIK 2020. 

Berlin, Germany, KernD and KTG, 

www.kerntechnik.com 

08.06. – 09.06.2020 

Decommissioning Strategy Forum. Nashville, TN, 

USA, ExchangeMonitor, 

www.decommissioningstrategy.com 

Currently working to evaluate 

all of their options. 

14.06. – 17.06.2020 

The Society for Risk Analysis – European 

Conference. Espoo, Finland, Aalto University, 

http://www.sraeurope.eu 

29.06. – 03.07.2020 

International Conference on the Safe Transport of 

Radioactive Material. Vienna, Austria, IAEA, 

www.iaea.org/events/international-conference-onthe-safe-transport-of-radioactive-material-2020 

13.07. – 16.07.2020 

46 th NITSL Conference - Fusing Power & People. 

Baltimore, MD, USA, Aalto University, www.nitsl.org 

02.08. – 06.08.2020 

ICONE 28 – 28 th International Conference on 

Nuclear Engineering. Disneyland Hotel, Anaheim, 

CA, ASME, https://event.asme.org/ICONE 

As of this date, the conference 

is currently scheduled to take place. 

20.08. – 21.08.2020 

The Power & Electricity World Africa 2020. 

Johannesburg, South Africa, Terrapinn, 

www.terrapinn.com/exhibition/power-electricityworld-africa/index.stm 

26.08.-04.09.2020 

The Frédéric Joliot/Otto Hahn Summer School 

on Nuclear Reactors “Physics, Fuels and Systems”. 

Aix-en-Provence, France, CEA & KIT, 

https://www.fjohss.eu 

01.09. – 04.09.2020 

IGORR – Standard Cooperation Event in the International 

Group on Research Reactors Conference. 

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

07.09. – 10.09.2020 

International Forum on Enhancing a Sustainable 

Nuclear Supply Chain. Helsinki, Finland, Foratom, 

www.events.foratom.org 

09.09. – 10.09.2020 

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 

14.09. – 15.09.2020 

International Nuclear Digital Experience. Paris, 

France, SFEN, www.sfen-index2020.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 

21.09.-25.09.2020 

64 th IAEA General Conference. Vienna, Austria, International 

Atomic Energy Agency IAEA, 

www.iaea.org 

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 

30.09. – 03.10.2020 

Nuclear Energy: Challenges and Prospects. Sochi, 

Russia, Pocatom, www.nsconf2020.ru 

Postponed to 11.10. – 15.10.2020 

RRFM – European Research Reactor Conference. 

Helsinki, Finland, European Nuclear Society, 

www.euronuclear.org/rrfm-2020-helsinki 

Postponed to 11.10. – 17.10.2020 

BEPU2020– Best Estimate Plus Uncertainty International 

Conference, Giardini Naxos. Sicily, Italy, 

NINE, www.nineeng.com 

12.10. – 17.10.2020 

FEC 2020 – 28 th IAEA Fusion Energy Conference. 

Nice, France, IAEA, www.iaea.org 

19.10. – 23.10.2020 

International Conference on the Management 

of Naturally Occurring Radioactive Materials 

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

www.iaea.org 

26.10. – 30.10.2020 

NuMat 2020 – 6 th Nuclear Materials Conference. 

Gent, Belgium, IAEA, www.iaea.org 

27.10. – 29.10.2020 

enlit (former European Utility Week and 

POWERGEN Europe). Milano, Italy, 

www.powergeneurope.com 

02.11. – 06.11.2020 

International Nuclear Reactor Materials 

Reliability Conference and Exhibition. 

New Orleans, Louisiana, EPRI, www.custom.cvent.com 

09.11. – 13.11.2020 

International Conference on Radiation Safety: 

Improving Radiation Protection in Practice. 

Vienna, Austria, IAEA, www.iaea.org 

Postponed to 18.11. – 19.11.2020 

INSC — International Nuclear Supply Chain 

Symposium. Munich, Germany, TÜV SÜD, 

www.tuev-sued.de 

24.11. – 26.11.2020 

ICOND 2020 – 9 th International Conference on 

Nuclear Decommissioning. Aachen, Germany, 

AiNT, www.icond.de 

07.12. – 10.12.2020 

SAMMI 2020 – Specialist Workshop on Advanced 

Measurement Method and Instrumentation 

for enhancing Severe Accident Management in 

an NPP addressing Emergency, Stabilization and 

Long-term Recovery Phases. Fukushima, Japan, 

NEA, www.sammi-2020.org 

Postponed to 08.12. – 10.12.2020 

World Nuclear Exhibition 2020. Paris Nord 

Villepinte, France, Gifen, 

www.world-nuclear-exhibition.com 

17.12. – 18.12.2020 

ICNESPP 2020 – 14. International Conference on 

Nuclear Engineering Systems and Power Plants. 

Kuala Lumpur, Malaysia, WASET, www.waset.org 

Postponed 

Date unknown 

20 th WCNDT – World Conference on 

Non-Destructive Testing. Seoul, Korea, EPRI, 

www.wcndt2020.com 

Postponed 

Date unknown 

International Conference on Operational Safety 

of Nuclear Power Plants. Beijing, China, IAEA, 

www.iaea.org 

Postponed 

Date unknown 

NDA Group Supply Chain Event. Telford, 

Shropshire, Cvent, web-eur.cvent.com/event/ 

2263a42b-a43a-4061-a960-f0715be47457/ 

summary 

Postponed or cancelled 

Date unknown 

NuclearEurope 2020 – Nuclear for a sustainable 

future. Paris, France, Foratom, 

events.foratom.org/nuclear-europe-2020 

Postponed 

Date unknown 

International Conference on Nuclear Knowledge 

Management and Human Resources Development: 

Challenges and Opportunities. Moscow, 

Russian Federation, IAEA, www.iaea.org 

Postponed to 2021 

13 th International Conference of the Croatian 

Nuclear Society. Zadar, Croatia, Croatian Nuclear 

Society, www.nuclear-option.org 

Postponed to 2021 

WNU Summer Institute 2020. Japan, World Nuclear 

University, www.world-nuclear-university.org 

Calendar


The European Nuclear Experimental 

Educational Platform (ENEEP) 

for Education and Training 

Marcella Cagnazzo, Helmuth Boeck, Štefan Čerba, Szabolcs Czifrus, Jan Haščík, Anže Jazbec, 

Jakub Lüley, Marcel Miglierini, Filip Osuský, Vladimir Radulović, Fabian Schaden, Lubomir Sklenka, 

Luka Snoj, Attila Tormási, Mario Villa, Branislav Vrban 

Introduction Research reactors played an important role for the development of nuclear technology during the 

past decades. However recently the interest of students to engage in nuclear technology has declined for several reasons 

such as very few new nuclear power projects in Europe and better careers in other technologies. In view of human 

resources development and nuclear knowledge transfer to the next generation, modern techniques in nuclear education 

and training is of utmost importance. Therefore, five institutions in Central Europe countries (Austria, Czech Republic, 

Hungary, Slovakia, Slovenia), with access to four research reactors of different designs, cooperate in an EU project 

called ENEEP with the aim to improve nuclear education in Europe. This paper describes the ENEEP offer and discusses 

the projected target. 

1 Nuclear Education & Training: 

The role of RRs 

In the second half of last century in many countries 

research reactors (RRs) were built to prepare the country 

for a follow-up nuclear power program. The Research 

Reactor Data Base (RRDB) of the International Atomic 

Energy Agency (IAEA) [1] lists that totally 880 RR were 

built with power levels from zero power up to several 10 th 

of MW. Table 1 summarises the current situation within 

Europe, showing the number of RRs in operation and the 

geographical distribution of those that perform Education 

& Training activities. According to these data, an idea of 

the impact of RRs in nuclear education is provided by the 

fact that almost 70 % of RRs in operation are utilized for 

Education & Training activities. 

Compared to nuclear power reactors, typical research 

reactors have completely other common features such as: 

p RR cores have small volume 

p Many have power less than 5 MW(t) 

p Lower operating temperatures 

p Less fresh fuel and spent fuel 

p Natural and forced cooling 

p Higher uranium enrichment 

p Very high power density in the core 

p Pulsing capability 

p Use of moderator and reflector for thermal flux 

irradiation 

To apply research reactors efficiently for education and 

training certain requirements have to be fulfilled by the 

reactor facility such as: 

p Simple construction 

p Easy access to the experimental facilities 

p Permission to manipulate fuel 

p Up-to-date digital instrumentation and control system 

p Availability of training laboratories with modern 

instruments 

p Adequate space in the reactor control room 

p Electronic textbooks in required language 

From the various types of research reactors developed in 

the past, low power research reactors, such as TRIGA 

(Training Research Isotope General Atomics), MNSR 

(Miniature Neutron Source Reactor), Slowpoke, Argonaut, 

AGN or SUR, are the most suitable reactors for education 

and training [2]. In contrast, in typical high flux reactors or 

MTR (Material Testing Reactor), such as Opal, BR2, FRM2, 

training is practically impossible because of high operational 

costs and low flexibility in the operation schedule. 

Low power research reactors are suitable for student’s 

education at all academic levels not only in nuclear 

engineering, but also in various non-nuclear engineering 

studies, such as power engineering, electrical engineering, 

natural-, medical- and physical sciences. 

Professional training is also possible at these type of 

research reactors: in this case, the specific conditions for 

training are mainly related to customers request (i.e. 

industrial companies including nuclear power plant 

Research Reactors in Europe [1] 

PLANNED, UNDER CONSTRUCTION, 

OPERATIONAL, TEMPORARY SHUTDOWN 

111 

Operational 95 

Used for Education & Training: 

Total 66 

Austria 1 

Belarus 1 

Belgium 3 

Czech Republic 3 

France 2 

Germany 5 

Greece 1 

Hungary 2 

Italy 3 

Kazakhstan 2 

Netherlands 2 

Poland 1 

Romania 2 

Russian Federation 31 

Slovenia 1 

Switzerland 1 

Turkey 1 

Ukraine 3 

Uzbekistan 1 

| Tab. 1. 

Number of research reactors (RRs) in Europe with information about some 

of the most used applications. 

251 

FEATURE | RESEARCH AND INNOVATION 

Feature 

The European Nuclear Experimental Educational Platform (ENEEP) for Education and Training ı 

M. Cagnazzo, H. Boeck, Š. Čerba, S. Czifrus, J. Haščík, A. Jazbec, J. Lüley, M. Miglierini, F. Osuský, V. Radulović, F. Schaden, L. Sklenka, L. Snoj, A. Tormási, M. Villa, B. Vrban


FEATURE | RESEARCH AND INNOVATION 252 

operators). As they are strictly cost-benefit oriented, they 

are therefore looking for high-quality training focused on 

the specific needs of their organisation at reasonable costs. 

Education and training are similar disciplines that are 

often confused. This may be because they use the same or 

very similar pedagogical methods, instruments and 

experimental equipment, but they are very different from 

the point of view of the target audience and the range of 

knowledge transferred to the audience. 

Education is a broader term and is connected only to 

students during their educational process where students 

must obtain a broad overview of the studied curricula. 

Training is a narrower term connected with a profession, 

and the main goal of training is to prepare professionals for 

a specific position. It means training young professionals at 

the beginning of their career, as well as experienced 

workers participating in lifelong learning. Training mostly 

represents short-term courses with well-defined objectives. 

Preparation of training courses for such participants must 

consider both an initial training and regular refreshment 

courses. 

Academic education in nuclear engineering, is mainly 

based on theoretical lectures and exercises supplemented 

by modelling of real or simplified reactor systems by 

various computational codes. Computer modelling is very 

cheap compared with real experiments and it can be easily 

implemented into any academic curricula without any 

need for building complex laboratories. However, it should 

be considered that, without real experimental works and 

without hands-on experiences, future nuclear engineers 

will be handicapped in their professional career of 

potential workers in this field. This situation is very similar 

to that as if a newcomer country, which is going to build a 

nuclear power plant, constructs a low power nuclear 

research reactor as a first step of its nuclear experience. 

During the building and operation of research reactors, 

engineers, physicists, chemists, regulatory body and 

governmental staff related to the nuclear field can obtain a 

real experience through various dedicated experiments 

and hand-on activities at research nuclear reactors. 

Nowadays, it is also difficult to enable access to research 

reactors for both students and their instructors to provide 

possibility to perform nuclear reactor physics experiments 

or hands-on reactor technology experience. One reason is 

the increasing security regulations for trainees working 

near or at the reactor, the second problem is the logistic 

and the financing to participate of the trainees for a course 

of several weeks including costs such as travel, local 

transportation, accommodation, visa, health insurance, 

food restrictions, cultural differences, etc. Already in 2007, 

the IAEA called for a Technical Meeting (TM) on the role 

of universities in preserving and managing nuclear 

knowledge [3]; while a few years later, in 2012, the OECD/ 

NEA published a report indicating it’s concern on nuclear 

education in Europe [4]. 

2 ENEEP 

2.1 Concept and approach 

In order to address the needs in terms of experimental 

education and hands-on activities in nuclear curricula, 

particularly in the field of nuclear safety and radiation protection, 

the European Nuclear Experimental Educational 

Platform (ENEEP) has been established by five founding 

members. 

The ENEEP is an open platform for any European 

university or European research institute that are actively 

involved in experimental nuclear education, training and 

competence building. 

The ENEEP well represents the typical activities in 

experimental nuclear education, training and competence 

building; and can count on 4 operational research reactors 

and experimental reactor courses routinely offered for 

students in Nuclear Engineering. 

The project for the development and initial demonstration 

of ENEEP is funded by the European Union under 

the topic NFRP-2018-7: “Availability and use of research 

infrastructures for education, training and competence 

building” [5]. 

The ENEEP development plan includes not only its 

establishment, but also the demonstration of ENEEP 

education and training capabilities. As a part of the project 

in fact, demonstration of educational and training 

capabilities of the ENEEP will be carried out through 

dedicated educational activities (both group and individual 

ones) organised at the ENEEP partner facilities. Two 

types of demonstration educational and training activities 

will be prepared and carried out: 

p Group activity: As a group activity, one 2-week 

educational course will be organised for a group up to 

10 students at two experimental nuclear facilities 

which belong to two of the ENEEP partners. Besides, 

one 1-week training course will be organised up to 

10 trainees at one experimental nuclear facility which 

belongs to the one of the ENEEP partners. 

p Individual activity: As an individual activity, two 

1-week individual educational/training courses will be 

organised (each course for one student/trainee) at the 

premises of other two consortium partners. 

ENEEP will enable access to research infrastructures. The 

exact number of future users will depend on the needs of 

nuclear industry at the EU level. The exact number of 

future users is difficult to predict due to high volatility of 

energy policies among the EU. Based on the experience 

from the last years, we expect that the number of students 

and trainees during five years after the project end will 

reach up to 1300 persons. 

2.2 Objectives of ENEEP 

The ENEEP will create opportunities to get access to 

nuclear experimental facilities such as research reactors 

and specific experimental laboratories for university 

students at all academic levels (bachelors, masters and 

doctoral), professors, lecturers, experts in nuclear 

education, etc. In addition to the nuclear education, 

ENEEP will allow also for specific nuclear training of 

professionals, particularly young professionals and 

post-docs at the beginning of their career. Moreover, staff 

from governmental and non-commercially oriented 

companies such as regulatory bodies, governmental 

organizations dealing with various aspects of peaceful use 

of nuclear energy, research institutions, etc. will be trained. 

The aim and the overall objective of the project is to build 

a European Nuclear Experimental Educational Platform 

(ENEEP) which fulfils the needs of European users in order 

to significantly enhance their experimental education and 

hands-on activities in nuclear curricula, particularly in the 

field of nuclear safety and radiation protection. 

ENEEP is established as an open platform for any 

European university or European research institute that is 

actively involved in experimental nuclear education, 

training and competence building. The ENEEP platform, 

aims to become the leading European platform offering 

experimental nuclear education and training activities. 

Feature 




2.3 Expected impact of ENEEP 

The ENEEP is expected to contribute within the next few 

years to the development of multi-disciplinary nuclear 

competences and increased availability of suitably 

qualified researchers, engineers and employees in crucial 

fields like nuclear safety, radiation protection, decommissioning, 

radioactive waste management, etc. 

The platform will address the need of maintaining the 

availability of experimental nuclear education, training 

and competence building at research facilities at the 

European level, which is recently ever more challenging, 

due to numerous research facilities being shut down, high 

facility operating costs, high level of retirement of 

personnel in the nuclear field, increasing complex security 

issues related to access to nuclear facilities, etc. 

The ENEEP will interconnect the partner research 

facilities in a coordinated effort to prepare and make 

available modern education, training and competence 

building activities to students and trainees communities. 

The impact and value of ENEEP is in providing access to 

nuclear experimental facilities and allowing students 

and trainees to conduct actual experimental activities. 

Experience gained through real experimental work is 

long-lasting and allows to complement and consolidate the 

knowledge in the nuclear field acquired in the framework 

of lectures and specialized courses in a long term. More 

importantly, the experience gained through ENEEP will 

broaden the young generation’s horizons in safe and secure 

operation of current and future nuclear installations. 

3 ENEEP Partners Institutions and offer 

The five ENEEP founding partners (STU – Slovak University 

of Technology in Slovakia, CTU – Czech Technical University 

in Czech Republic, TU Wien – Technische Universität Wien 

in Austria, JSI – Jožef Stefan Institute in Slovenia, and BME 

– Budapest University of Technology and Economics in 

Hungary) are themselves heavily involved in experimental 

nuclear education, training and competence building. 

Four of the project consortium partners operate small 

nuclear research reactors of different designs, which are 

easily accessible for hands-on education, training, and 

competence building. The fifth partner has specific 

laboratories for nuclear education and training. 

At present, more than 60 experiments constitute the 

offer available at the ENEEP. Table 2 shows the main 

facilities [6] used for E&T and the number of offered 

experiments at each partner institution. For the current 

number and variety of the experiments, this is considered 

satisfactory. In fact, the collected E&T activities assure a full 

coverage at a varied level, both in term of levels of recipients 

(under-graduate, graduate and post graduate) and in term 

of level of the education and training activity itself (basic, 

advanced and complex). Nevertheless, the present database 

is intended as a living container that in future will continuously 

improve taking into account changes due to updates of 

the experimental protocols, modification of the conditions 

of delivery, or for the addition of new experiments that will 

become available over time. 

To allow the ENEEP interested users to search among 

the offer and built a tailored curricula based on the specific 

interests, the relevant information for each offered 

experiment have been shaped into a standard format, that 

includes for example a summary of what the attendant will 

learn, which is the required pre-knowledge to be admitted, 

if there are limitations and how to enrol. The ENEEP web 

page [7] will provide the needed information about how to 

select and enrol for the different E&T activities. 

ENEEP Partner Facility/ies Number of E&T 

experiments 

TU Wien (Austria) TRIGA Mark II RR (250 kW) 11 

CTU (Czech Republic) Training Reactor VR-1 (100 W) 17 

BME (Hungary) Training Reactor (100 kW) 8 

STU (Slovakia) Laboratories of Nuclear Physics 11 

JSI (Slovenia) TRIGA Mark II RR (250 kW) 14 

| Tab. 2. 

ENEEP partners institutions with their main facilities and number of Education & Training (E&T) 

experiments immediately available for interested users. 

The partners institutions are here briefly described 

giving an overview about the main facilities [6] used for 

Education & Training and providing some examples of 

available exercises [8]. 

3.1 TU Wien (Austria) 

The Technische Universität Wien (TU Wien), includes, 

within the Faculty of Physics, the Atominstitut (ATI) [9] 

dedicated to today‘s broad range of research and education 

in nuclear and particle physics; neutron-, atomic-, 

quantum- physics and quantum optics; radiation- and 

reactor physics. A central facility thereby is the TRIGA 

(Training, Research, Isotope Production, General Atomic) 

Mark II research reactor (Figure 1) and the connected 

teaching and research infrastructure, which allow to 

educate and work with radioactive materials and ionizing 

radiation. An important contribution thereby is the training 

of international experts and junior safeguards trainees for 

the International Atomic Energy Agency (IAEA). 

The reactor maximum power is 250 kW (thermal) in 

steady state condition and 250 MW in pulse operation. The 

power rise is accompanied by an increase in the maximum 

neutron flux density from 1x10 13 n cm -2 s -1 (at 250 kW) to 

1x10 16 n cm -2 s -1 (at 250 MW). The TRIGA Mark II is 

equipped with a number of irradiation devices such as 

5 reflector irradiation tubes, 1 central irradiation tube, 

1 pneumatic transfer system (transfer time 3 s), 4 horizontal 

neutron beam holes, 1 thermal column, 1 neutron 

radiography facility. 

One of offered experiment at TU Wien is the so called 

Critical Experiment. 

| Fig. 1. 

The TRIGA Mark-II reactor (TU Wien, Austria). View into the reactor tank. 


Feature 





In this experiment, ten fuel elements are removed from 

the core and stored in the reactor tank. With the neutron 

source in the core and all three control rods fully out, the 

neutron count rate is measured with a very sensitive 

neutron detector. Then, fuel elements are sequentially 

loaded from centre to the outside of the core back to their 

respective core positions one by one. For each step the 

count rate from the neutron detector is recorded. When a 

certain number of fuel elements have been added, the reactor 

reaches criticality. The same procedure is performed 

while all control rods are completely in the core, in this 

case criticality is not reached after a complete core loading. 

The number of fuel elements necessary for reactor criticality 

is determined by extrapolation of the criticality curve. 

In order to do this, the reciprocal count rate has to be compared 

with the number of loaded fuel elements. Criticality 

is achieved when the reciprocal count rate approaches to 

zero. 

During this experiment the participants will learn the 

importance of the criticality condition in a nuclear reactor 

and how to acquire this information. 

3.2 CTU (Czech Republic) 

The Czech Technical University in Prague is one of the 

oldest technical universities in Europe which was found in 

1707. The Department of Nuclear Reactors, which operates 

the VR-1 Reactor, belongs to the Faculty of Nuclear 

Sciences and Physical Engineering. 

The Training Reactor VR-1 [10] (Figure 2), which is in 

operation since 1990, is a pool-type light-water reactor 

based on low enriched uranium with the maximal thermal 

power of 100 W. The reactor is equipped with standard 

experimental devices such as vertical and horizontal beam 

ports and a rabbit system. However, the reactor also 

includes experimental devices that have been developed 

especially for experimental education and training. The 

VR-1 reactor is a key experimental facility for education of 

the students of the Czech universities in the field of nuclear 

engineering and for research and development in the field 

of safe operation of nuclear installations, theoretical and 

experimental reactor and neutron physics, nuclear fuel 

cycle and fuel management, and as a source of neutrons 

for dedicated experiments. Almost 150 Czech and 

| Fig. 2. 

The Training Reactor VR-1 (CTU, Czech Republic). 

100-120 foreign students attend the education at the VR-1 

reactor every year. Foreign students come from USA, 

United Kingdom, Slovakia, Germany, Sweden, Finland, 

and Poland. 

One of the most attractive educational experiments, 

which are carried out at the reactor, is the study of 

advanced reactor kinetics. During this experiment, all 

three basic reactor characteristics including pulse, 

transient and frequency are studied. Deep understanding 

of basic processes of time-dependent reactor kinetics, i.e. 

reactor transients, are essential for safe operation of any 

reactor. An instrumentation for fast reactivity changes is 

used in the VR-1 reactor when demonstrating response to 

a pulse reactivity perturbation (pulse characteristic) and 

to a transient reactivity perturbation (transient characteristic). 

This instrumentation is based on a pneumatic 

drive which allows fast vertical movement of a small 

specimen containing neutron-absorbing or fissionable 

material from one position to another. Fast periodic 

changes of the pressurised air flow from upwards to 

downwards, i.e. inlet of the air under and over the 

pneumatic drive plunger, allow well-defined and 

controlled up-and-down movement of the specimen. 

Another instrumentation for frequency reactivity changes 

is used to study the VR-1 reactor response to frequency 

characteristics. This instrumentation is based on frequency 

reactivity changes caused by rotation of a EK-10 fuel pin 

eccentrically located in two plastic tubes. 

3.3 BME (Hungary) 

The Institute of Nuclear Techniques (NTI) of the Budapest 

University of Technology and Economics (BME) is the 

leading organization in nuclear training in Hungary. 

The Institute operates a small reactor facility, which is 

equipped with various laboratories. The Training Reactor 

[11] (Figure 3) of BME, which started operation in 1971, 

is the scene of numerous reactor and radiation related 

exercises for undergraduate and graduate students and 

serves as a neutron and gamma radiation source for 

research. This is a light water moderated and cooled 

reactor with 100 kW nominal thermal power. The core 

consists of EK-10 type fuel assemblies, containing 10 % 

enriched UO 2 in metal magnesium matrix. The main 

purpose of the facility is the training of young engineers 

and physicists. On the other hand, research projects are 

also carried out on the reactor and using the connected 

experimental devices. Neutron and gamma irradiations 

can be performed using the vertical irradiation channels, 

horizontal beam tubes, the large irradiation tunnel and the 

pneumatic rabbit systems. Radiochemical laboratories and 

a hot cell support the training and research activities. For 

the design process, the experience gained during the 

operation of several critical assemblies, and the 2 MW 

Budapest Research Reactor (originally designed by Soviet 

engineers), was effectively applied. 

One of the offered experiments at the BME Training 

Reactor is the Reactor Operation Exercise. 

The purpose of the operational exercise is to understand 

the physical processes in a nuclear reactor, its 

structure, its nuclear and technological equipment and its 

measuring and control systems. During the exercise the 

students learn how a nuclear reactor is controlled 

( measuring chains, control rods, etc.), study and perform 

maneuvers, such as reactor start-up, power increase and 

decrease, automatic and manual operation, and shutdown. 

They study the method of inserting or removing 

reactivity into or from the reactor core by moving the 

Feature 




| Fig. 3. 

The Training Reactor (BME, Hungary). 

control and safety rods, learn how the safety systems 

intervene in case an error (either human or electronic) 

occurs. They also study the water circulation system, 

which is separated into primary and secondary parts; and 

when and how these systems should be operated. During 

the exercise students also obtain information on the 

systems measuring technological parameters and their 

role in the safe operation of the reactor. 

these structural imperfections. In this way, PAS is a specific 

tool for detecting defects in the material lattice. 

The experiment Phase analysis by Mössbauer spectrometry 

offers a deep insight into the use of this nuclearbased 

analytical method for practical utilization. The 

trainees investigate iron-based samples which consist of 

more than one crystalline phase. First, Mössbauer spectra 

of pure Fe-containing phases are recorded, then an 

unknown sample is measured. Each crystallographic phase 

is characterized by its own set of hyperfine parameters 

which are reflected in the corresponding Mössbauer 

spectra, eventually resulting in several spectral components. 

Subsequently, the measured spectrum is 

decomposed, and each component/spectral line is 

assigned to a known phase based on the reference 

measurements, standards, and/or literature data. The area 

under the Mössbauer spectrum of a given phase is used to 

determine the amount of investigated material in the 

mixture. The trainees go through both theoretical and 

practical aspects of the experiment including sample 

preparation, experiment setup, spectra acquisition, spectra 

evaluation, and interpretation of the obtained spectral 

parameters. 


3.4 STU (Slovakia) 

The Slovak University of Technology in Bratislava (STU) is 

the coordinator of the ENEEP project. STU [12] is a modern 

educational institution and it is ranked as the best 

university in chemical technologies, computer and 

technical sciences in Slovakia. The Institute of Nuclear and 

Physical Engineering (INPE) is one of the 10 institutes 

working as a part of the Faculty of Electrical Engineering 

and Information Technology (FEI) of STU. It is responsible 

for university education in the area of nuclear and physical 

engineering. INPE is active in various fields of nuclear 

research and development. There are currently 16 laboratories 

devoted to nuclear physics operated at INPE 

(Figure 4). The most important ones are the following: 

The Laboratory of Reactor Physics is designed for 

neutron activation, neutron source emission rate and 

neutron diffusion length and Fermi age measurements. In 

the laboratory, Pu-Be and Am-Be neutron sources and 

apparatus for remote control and monitoring of experiments 

are used. Moreover, the laboratory deals with 

computationally complex problems in the field of reactor 

core and shielding analyses as well as nuclear data 

treatment. The Mössbauer Spectrometry Laboratory is used 

for a non-destructive material testing using the Mössbauer 

effect with a wide diagnostic potential, applicable to all 

iron-containing materials. It enables unambiguous identification 

of crystallographic sites in structurally ordered 

phases along with distributions of hyperfine interactions 

between nuclei and electron shells in amorphous 

structures. The Positron Annihilation Spectroscopy (PAS) 

Laboratory employs positrons emitted from a suitable 

radionuclide for a non-destructive testing of materials. 

Positrons are trapped at structural defects where they 

annihilate with the materials’ electrons. The subsequently 

detected annihilation photons bear information about 

| Fig. 4. 

Laboratory facilities (STU, Slovakia). 

3.5 JSI (Slovenia) 

The Jožef Stefan Institute (JSI) is the leading Slovenian 

scientific research institute, covering a broad spectrum of 

basic and applied research. The mission of the Jožef Stefan 

Institute is the accumulation and dissemination of 

knowledge at the frontiers of natural science and 

technology to the benefit of society at large through the 

pursuit of education, learning, research, and development 

of high technology at the highest international levels of 

excellence. 

The JSI TRIGA (Training, Research, Isotope production, 

General Atomics) Mark II [13] research reactor (Figure 5) 

has been in operation since 1966. It is a light water reactor, 

with solid fuel elements consisting of a homogeneous 

dispersion of 20 % enriched uranium and zirconium 

hydride moderator. The reactor core consists of about 

60 fuel elements, yielding the maximum neutron flux in 

the central thimble of about 2×10 13 n cm -2 s -1 . A 40 

position rotary specimen rack (located around the fuel 

elements), two pneumatic tube transfer rabbit systems, 

Feature 





| Fig. 5. 

The 250 kW TRIGA Mark II research reactor (JSI Slovenia). 

as well as central thimble and four extra positions in 

the core are used for irradiation of samples. Additional 

experimental facilities include two radial and two 

tangential beam tubes, a graphite thermal column and a 

thermalizing column. Since its commissioning the reactor 

has been playing an important role in developing nuclear 

technology and safety culture in Slovenia as is one of a 

few centres of modern technology in the country. Its 

international scientific cooperation and recognized 

reputation are important for promotion of the JSI, 

Slovenian science and Slovenia as a country in the world. 

One of the offered experiments at the JSI is the In-core 

flux mapping experiment. 

In this experiment, a miniature U-235 fission chamber 

with an outer diameter of 3 mm is inserted into a 6 m long 

guide tube, which is located, during the experiment, into 

several measurement positions in the reactor core. For 

each position, the fission chamber is moved vertically, 

from the guide tube bottom (below the fuel level), by 

about 70 cm (reaching well above the fuel level) in multiple 

steps, and the axial neutron flux profile is measured. The 

audience gains first-hand insight into the overall shape of 

the axial neutron flux profile in the reactor, including 

specific features due to the core heterogeneity. By repeating 

the procedure in different radial positions, the radial flux 

profile can be investigated as well. 

Although there exist a number of information platforms on 

nuclear education in Europe (e.g. ENEN [14]), the main 

purpose of the ENEEP is to standardize and simplify access 

of potential user to the best available nuclear infrastructure. 

Even though the laboratories and research 

reactors are distributed over Central Europe, the 

established platform will bring these facilities closer to 

individuals or groups like never before. Well experienced 

staff and supervisors are able to prepare user specific 

experiments and training course based on their requirements 

and target their professional needs. All of these 

aspects predetermine the ENEEP to be unique entity which 

will contribute both to nuclear knowledge competence 

building and to improve research reactor utilization. 

5 Acknowledgment 

The ENEEP project has received funding from the European 

Union‘s Horizon 2020 research and innovation programme 

under grant agreement No. 847555. 

References 

[1] International Atomic Energy Agency Research Reactor Database (RRDB) 

https://nucleus.iaea.org/RRDB/RR/ReactorSearch.aspx. 

[2] Research Reactors IAEA: https://www.iaea.org/topics/research-reactors (access March 27, 2020). 

[3] Management of nuclear knowledge, Report of IAEA Technical Meeting on the “Role of 

Universities in Preserving and Managing Nuclear Knowledge”, IAEA Vienna, Austria – INIS IAEA 

(2008) 41011598-41-03. 

[4] Nuclear Education and Training: From Concern to Capability, OECD/NEA, OECD PUBLICATIONS, 

2, rue André-Pascal, 75775 PARIS CEDEX 16 (2012) ISBN 978-92-64-17637-9. 

[5] https://cordis.europa.eu/project/id/847555 

[6] D3.1 Database of ENEEP educational and training facilities, Deliverable Report, version 1, 

2019-09-30, Copyright © ENEEP Project Consortium 2019. 

[7] http://www.eneep.org/ 

[8] D3.2 Database of ENEEP educational and training experiments, Deliverable Report, version 1, 

2020-01-31, Copyright © ENEEP Project Consortium 2019. 

[9] www.ati.ac.at 

[10] www.reaktor-vr1.cz 

[11] www.reak.bme.hu 

[12] www.stuba.sk 

[13] http://www.rcp.ijs.si/ric/index-a.htm 

[14] European Nuclear Education Network (ENEN): https://enen.eu/ 

Authors 

Marcella Cagnazzo, 

Helmuth Boeck, 

Fabian Schaden, 

Mario Villa 

Technische Universität Wien – Atominstitut, 

Stadionallee 2, 1020 Wien, Austria 

Anže Jazbec, 

Vladimir Radulović, 

Luka Snoj 

Jožef Stefan Institute, Reactor Physics Division, 

Jamova 39, 1000 Ljubljana, Slovenia 

4 Conclusions 

The European Nuclear Experimental Educational Platform 

(ENEEP) project was initiated in year 2019 funded 

by the European Union under the topic – NFRP-2018-7: 

“ Availability and use of research infrastructures for 

education, training and competence building”. The ENEEP 

is an open platform for European university and/or 

European research institute involved in experimental 

nuclear education, training and competence building is 

expected to be completed by mid of year 2022. 

The present paper illustrates the objectives, the 

partner’s institutions, the available facilities and the E&T 

activities offered by ENEEP, which are immediately 

available to the interested parties. 

From the first analysis of the current ENEEP capabilities 

(i.e. more than 60 experiments), it can be concluded that 

the number and variety of the experiments is satisfactory. 

Štefan Čerba, 

Jan Haščík, 

Jakub Lüley, 

Filip Osuský, 

Branislav Vrban 

Slovak University of Technology in Bratislava, 

Faculty of Electrical Engineering and Information 

Technology, Institute of Nuclear and Physical Engineering, 

Ilkovičova 3, 812 19 Bratislava, Slovakia 

Szabolcs Czifrus, 

Attila Tormási 

Budapest University of Technology and Economics, 

Institute of Nuclear Techniques, 

Műegyetem rkp. 3, 1111 Budapest, Hungary 

Marcel Miglierini, 

Lubomir Sklenka 

Czech Technical University in Prague, 

Faculty of Nuclear Sciences and Physical Engineering, 

Brehova 7, 115 19 Prague 1, Czech Republic 

Feature 




Did you know...? 

Charting the French Nuclear Industry – Report 2019 

At the end of March 2020 the Groupement des Industriels 

Français de l'Énergie Nucléaire (GIFEN), founded in 2018 by 

French nuclear companies and associations (200 members) and the 

Comité statégique de la filière nucléaire (CSFN), founded in 2011 

(80 members), published the updated report on the French nuclear 

sector “Cartographie de la filière nucléaire française 2019”. The 

report is based on a poll among the companies of the industry and 

provides an update to the 2014 study. 

Division of Revenues by Type of Company (in per cent) 

The main figures characterizing the industry is more than 

220,000 employees in over 3,000 companies with above average 

qualification level and significantly lower work force turnover (only 

7.8 per cent) than other French industrial sectors, 47.5 billion Euro 

turnover and 970 million Euro R&D expenses, with 53.3 per cent of 

companies active in export business which is realized to more than 

50 per cent outside of Europe. Below you can find graphs depicting 

the division of revenues by type of company and by activity. 

DID YOU EDITORIAL KNOW...? 

257 

Operators 

53.1 % 

Very small 

enterprises 

0.3 % 

Small and medium 

enterprises 

7.9 % 

Big companies 

11.7 % 

Intermediate size 

companies 

26.9 % 

Division of Revenue by Type of Activity (in per cent) 

70 

67.1 

60 

50 

40 

30 

20 

10 

0 

0.8 1.3 2.0 2.2 2.5 

Other nuclear 

power generation 

related activities 

Remediation 

activities 

Decommissioning 

and dismantling 

activities 

R&D, 

studies 

Waste 

management 

activities 

10.0 

Building of 

nuclear facilities 

13.5 

Fuel cycle 

activities 

Operation and 

maintenance 

of existing fleet 

For further details 

please contact: 

Nicolas Wendler 

KernD 

Robert-Koch-Platz 4 

10115 Berlin 

Germany 

E-mail: presse@ 

KernD.de 

www.KernD.de 

Did you know...?


Das Atomgesetz und seine Zeit 

258 

SPOTLIGHT ON NUCLEAR LAW 

Christian Raetzke 

Das Gesetz über die friedliche Verwendung der Kernenergie und den Schutz gegen ihre Gefahren (Atomgesetz) (AtG) 

wurde am 23. Dezember 1959 verabschiedet und trat am 1. Januar 1960 in Kraft. Der somit vor kurzem fällige 

60. Geburtstag wurde aber nicht groß gefeiert. Kein Wunder, denn in den Zeiten des Kernenergieausstiegs hat das 

Atomrecht allgemein nicht mehr die Bedeutung von früher. Aber selbst das relative Gewicht des AtG im Gesamtbereich 

des Atom- und Strahlenschutzrechts hat sich verringert. 

Bei seiner Verkündung war das AtG als abdeckendes 

Gesetz für den gesamten Bereich der Kernenergie und des 

Strahlenschutzes gedacht. Wie das bei Gesetzen gerade 

im technischen Umweltrecht so üblich ist, enthielt es 

von vornherein zahlreiche Verordnungsermächtigungen, 

auf deren Grundlage eine Reihe von Verordnungen mit 

Detailregelungen erlassen wurden, so etwa auch die 

Strahlenschutzverordnung. Sie waren aber aus dem AtG 

abgeleitet; das AtG war gleichsam das „Mutterschiff“ des 

gesamten Rechtsgebiets. 

Immer wenn in den darauffolgenden Jahrzehnten neue 

Regelungskomplexe im Atomrecht geschaffen wurden, 

dann wurden sie in das AtG eingefügt. So hat der Gesetzgeber 

etwa 1976 mit § 9a AtG die grundlegende Regelung 

für die Pflichtenverteilung bei der Entsorgung radioaktiver 

Abfälle getroffen. Erkennungszeichen solcher nach träglich 

in ein Gesetz eingefügten Paragraphen ist meist der kleine 

Buchstabe – dadurch entfällt die Notwendigkeit, alle nachfolgenden 

Paragraphen neu durchnummerieren zu müssen. 

Innerhalb eines Paragraphen können auf diese Weise 

auch Absätze neu eingefügt werden, wie wiederum § 9a 

zeigt, der seit 1976 mehrfach ausgebaut wurde und heute 

zum Beispiel auch die Absätze 1a bis 1e hat. Das mutet 

zuweilen recht verschachtelt an. Wie dem auch sei: so 

wuchs das AtG im Lauf der Zeit. 

In den letzten Jahren hat sich im Atomrecht aber 

zunehmend der Trend etabliert, für neue oder zu überarbeitende 

Regelungskomplexe ein jeweils eigenes Gesetz 

zu schaffen, sie also aus dem AtG auszugliedern. Das trifft 

vor allem auf zwei Bereiche zu: die Entsorgung und den 

Strahlenschutz. 

Die Regelungen zur Entsorgung sind zunächst im AtG 

weiter ausgebaut worden; davon zeugen der bereits 

erwähnte § 9a mit den Grundpflichten der Entsorgung 

sowie die nachfolgenden §§ 9b bis 9i, in denen es 

hauptsächlich um die Zulassung (Planfeststellung oder 

Genehmigung) von Bundesendlagern geht. Als jedoch im 

Jahre 2013 Gesetzgebung über den Neustart der Standortsuche 

für das Endlager für hochradioaktive Abfälle anstand, 

hat man für die gewünschte sehr ausführliche Regelung 

ein eigenes Gesetz geschaffen, das Standortauswahlgesetz. 

Ähnlich ging es 2017 mit den Regelungen zur Neuordnung 

der Verantwortung in der kerntechnischen Entsorgung, die 

auf die Arbeit der „Kommission zur Überprüfung der Finanzierung 

des Kernenergieausstiegs“ (KFK) zurückgingen: 

zwar wurde auch das AtG angepasst, zur Aufnahme der 

wesentlichen Regelungen wurden jedoch mehrere neue 

Gesetze geschaffen, vor allem das Entsorgungsfondsgesetz 

und das Entsorgungsübergangsgesetz. 

Der Bereich des Strahlenschutzes hat sich, wie allgemein 

bekannt und in dieser Rubrik auch bereits mehrfach 

angesprochen, 2017/2018 ebenfalls vom AtG emanzipiert 

und hat mit dem Strahlenschutzgesetz (StrlSchG) vom 27. 

Juni 2017 (in Kraft getreten in zwei Phasen bis 31.12.2018) 

seine eigene höchst bedeutsame gesetzliche Regelung 

bekommen. Da das AtG selbst nur wenige punktuelle – oft 

auch erst später eingefügte – Regelungen speziell zum 

Strahlenschutz enthielt, musste es nicht weitläufig 

„ amputiert“ werden; ein paar Para graphen nur wanderten 

ins neue StrlSchG hinüber. „Schlimmer“ aus Sicht des AtG 

war der Verlust der Strahlen schutzverordnung (StrlSchV). 

Ihre Inhalte finden sich nun teils im StrlSchG selbst wieder 

und teils in der neuen StrlSchV vom 29. November 2018, 

die aber ihrerseits nunmehr ganz überwiegend auf 

Ermächtigungen beruht, die im StrlSchG und nicht im AtG 

enthalten sind. 

Der Jubilar hat also zwischen seinem 50. (wo er noch 

recht kräftig war) und 60. Geburtstag stark abgebaut; 

ehrgeizige Abkömmlinge haben sich vorgedrängt und dem 

Patriarchen neue und teils auch alte Aufgaben abgenommen. 

Kann man ihn also langsam abschreiben? Das 

wäre verfrüht. Noch steht viel Wichtiges im AtG. Noch auf 

Jahrzehnte wird es laufende und in Stilllegung befindliche 

kerntechnische Anlagen nach § 7 AtG geben. Der Fokus hat 

sich aber insgesamt natürlich auf die Entsorgung verschoben. 

§ 9a AtG bleibt in diesem Bereich die Grundnorm, 

auf der die neuen Spezialgesetze aufbauen. Die im 

AtG enthaltenen Regelungen zu Genehmigung und 

Aufsicht über Anlagen, die der Lagerung und Behandlung 

von kernbrennstoffhaltigen Abfällen und der Endlagerung 

aller Arten von radioaktiven Abfällen dienen, werden noch 

auf eine lange Epoche relevant sein. Ein anderer wichtiger 

Abschnitt des AtG ist das Haftungskapitel (beginnend mit 

§ 25), das für alle Bereiche des Atom- und Strahlenschutzrechts 

gilt; das StrlSchG etwa verweist in seinem § 176 

einfach darauf. Wichtig bleiben auch die auf das AtG 

gestützten Verordnungen wie die Atomrechtliche Verfahrensverordnung 

(AtVfV) oder die Atomrechtliche 

Deckungs vorsorge-Verordnung (AtDeckV). Im Bereich der 

Entsorgung hat das AtG sogar Zuwachs in Gestalt einer 

neuen Verordnung bekommen, der Atomrechtlichen Entsorgungsverordnung 

(AtEV), deren Regelungen inhaltlich 

aus der alten StrlSchV (§§ 72-79) übernommen wurden. 

Gerade mit Blick auf die Entsorgung könnte unser 

Jubilar also (mit Schillers König Philipp) sagen: „Die Welt 

ist noch auf einen Abend mein“. Dennoch hat er seine 

exklusive Stellung als allmächtiger Patriarch seines 

Rechtsgebiets eingebüßt. Das Atom- und Strahlenschutzrecht 

hat sich weiter ausdifferenziert; neue Gesetze 

sprießen empor; die gelbe Atomrechts-Textsammlung des 

Nomos-Verlags wird mit jeder Auflage dicker und enthält 

mehr Nummern (in der 10. Auflage von 1986, von der der 

Verfasser ein Exemplar antiquarisch erstanden hat, sind es 

zwölf Rechtstexte, in der gegenwärtigen 36. Auflage sind 

es 30). Man könnte also sogar, wenn man den Fokus vom 

AtG weg auf das gesamte Atom- und Strahlenschutzrecht 

richtet, zum Eindruck gelangen, dieses Rechtsgebiet 

wachse und gedeihe immer mehr. Aber das ist ein anderes 

Thema. 

Author 

Rechtsanwalt Dr. Christian Raetzke 

Beethovenstr. 19 

04107 Leipzig 

Spotlight on Nuclear Law 

Atomic Law – Changes Over Time ı Christian Raetzke


BER II – The End of an Era 

Helmholtz-Zentrum Berlin für Materialien und Energie 

46 Years in a Nut Shell December 9, 1973 – the day on which the 

BER II went into operation. For almost 50 years he shaped the research. 

Over the decades, scientists have repeatedly set new research priorities. 

Technical developments and political framework conditions also had an 

impact on the operation. 

Structural research was established at the former Hahn Meitner 

Institute with BER II. The new research focus had replaced the previous 

focus on nuclear chemistry. It got a decisive boost when the BER II 

changed its face again significantly in the early 1990s. The conversion 

from 5 MW to 10 MW of reactor output made it possible to establish 

international user operations and thus develop research with neutrons in 

Berlin on a completely new basis. The BER II became the most modern 

device in Germany for experiments with neutron scattering. 

Overview of all major milestones 

from research with neutrons at BER II 

December 9, 1973 

The Beginning with 5 MW of Power 

| View into the experimental hall. 

The planning for the construction of BER II already began in 1966. The aim was to focus on a new branch of research, the structural 

research. This increasingly replaced the nuclear chemistry that had previously characterized the experiments at the reactor. 

The operation of the predecessor, BER I, was discontinued in 1971. 

259 

RESEARCH AND INNOVATION 

1981 

Start of Cooperation with the Gemäldegalerie Berlin 

Discussions with the Rathgen Research Laboratory of the State Museums and the Gemäldegalerie Berlin started and led to a 

long-term cooperation. An experiment site for neutron activation analysis was set up and between 1984 and 1985 nine paintings 

by Rembrandt and his school were examined. Dis covering that the “man with the gold helmet” did not originate from Rembrandt 

was sensational. 

1986 – 1991 

Modification of BER II 

The first considerations for the conversion were made in 1975; later in 1982 it was an approvable concept. The construction began 

four years later. The reactor output was increased to 10 MW and a beryllium reflector was added to the core. This made it possible 

to significantly increase the neutron flux. At the same time, a cold neutron source was installed, a pressure vessel in which cryogenic 

hydrogen additionally breaks the neutrons. At BER II, slow, so-called cold neutrons could be generated for the first time – a great 

benefit for the research. 

1991 

Restart and Establishment of User Operation 

Due to politically determined delays in the approval process, the restart after the renovation took significantly longer than planned. 

However, from 1991 on the upgrade made completely new experiments possible. In addition, new experimental stations were 

set up to carry out more experiments at the same time. At the same time, the Working Group „Sample Environment“ was established 

to support users in their demanding experiments. The foundation for an internationally competitive user company was thus laid. 

It was organized in BENSC, the “Berlin Neutron Scattering Center”, which was founded in 1991 as a virtual institute at the HMI. 

Within a short time, BENSC had earned an excellent reputation worldwide for its user support. 

2000 

Conversation to low-enriched uranium 

At the turn of the century, reactor operations were switched from 

high-enriched uranium (HEU) to low-enriched uranium. The fuel 

elements were gradually replaced. In March 2000, a reactor core 

went into operation, which was operated for the first time 

completely without highly enriched uranium. 

2006 

Opening of neutron guide hall II 

The neutron guide hall II was built from 2004 to 2006. Among 

other things, the high field magnet later found its place here. 

| View into the neutron guide hall II. 


BER II – The End of an Era ı Helmholtz-Zentrum Berlin für Materialien und Energie


RESEARCH AND INNOVATION 260 

2009 

Fusion of HMI and BESSY – establishment of a common user service 

In January 2009, the Hahn Meitner Institute (HMI) and the Berliner Elektronenspeicherring-Gesellschaft für Synchrotronstrahlung 

(BESSY) merged to form the Helmholtz Center Berlin (HZB) for materials and energy. The fusion promoted the combined use of 

photons and neutrons in one location. Numerous research fields benefited from this, including the photovoltaics and materials 

research established at the HMI. In November 2009, the HZB invited to the “First Joint BER II and BESSY II Users’ Meeting”. More 

than 350 participants from all over the world accepted the invitation and had a cross-disciplinary exchange. 

2010 – 2012 

Replacement of the conical beam tube 

and upgrade of the neutron guide 

The long-planned replacement of the conical nozzle became 

necessary because the maximum service life for this component 

would have been reached in 2011. Other components, such as 

the cold neutron source with a moderator cell through which 

the neutrons fly, were also replaced. The scientists also used the 

break to improve the instruments and replace the neutron 

guides. They received a super mirror coating, three were 

widened and an additional (sixth) neutron guide was built. 

These improvements significantly increased the neutron flux – 

an enormous improvement for science that kept the BER II 

internationally competitive. 

2013 

Shut down decision 

| Exchange of the conical nozzle. 

On June 25, 2013, the supervisory board of the HZB decided to end science operations at the research reactor BER II at the end of 

2019. With the early announcement of the shutdown date, both the scientific users of BER II and the management were given 

planning security to set the course for a successful reorientation of research. The first plans for the dismantling of BER II already 

began in 2014. At the same time, interested parties in other neutron sources were addressed so that the neutron instruments are 

still available for research after the BER II has been switched off. 

2015 

Commissioning of the high field magnet 

After eight years of construction and development, the world’s 

strongest magnet for materials research with neutrons was 

put into operation. It operates with a hybrid magnet system 

and produces magnetic fields up to a strength of 26 Tesla. 

A normally conductive and a superconducting coil are connected 

in series. Cooperation partners from several countries were 

involved in the development of the high field magnet. The 

high field magnet laboratory in Tallahassee, USA was the lead 

partner in the partner consortium. Even if the high field magnet 

at HZB was only in operation for about five years, its construction 

is considered a pioneering achievement and scientific 

experiments have shown which questions can be investi gated 

with such high magnetic fields. 

2017 

Application for shut down 

In April 2017, the HZB submitted the basic application for 

decommis sioning and dismantling of BER II, which initiated 

the extensive approval process. In order to enable early public 

participation, the HZB invited to an information event at the 

end of 2017, attended by over a hundred interested parties. 

| Commissioning of the high field magnet. 

The dialogue group that subsequently formed has been working regularly since then and supports the dismantling process. 

2018 

The last neutron school 

The last neutron school at BER II took place in February 2018. After 38 successful years, the school has continued at the Australian 

institution ANSTO – with the participation of the HZB – since 2019. 

December 11, 2019 

End of operation 

The last time the reactor delivered neutrons was on December 11, 2019. Until then, eighteen neutron instruments were still 

in operation, ten of which were in full user operation. The measuring time was fully booked up to the last shift: In the last year 

of operation, there were more than 600 visits by users to BER II. 




Research Highlights 

SOLID-STATE PHYSICS AND MAGNETISM 

Magnetic monopoles discovered 

A sensational discovery was made at HZB in 2009: physicists led by HZB 

researcher Alan Tennant have demonstrated for the first time that magnetic 

monopoles can form under very special conditions. The North Pole and South 

Pole are separated from each other as far as it normally never happens! The 

exotic observation was achieved at temperatures almost at absolute zero in 

a dysprosium titanite crystal. With the help of neutron scattering, the HZB 

researchers were able to show that the magnetic moments inside the crystal 

are arranged in so-called spin spaghetti, at the ends of which the north and 

south poles are located. And because these are so far apart, the spin 

spaghetti behave like mono polies. The existence of such magnetic monopoles 

is predicted by quantum physics, but has never been observed before. 

The golden cut exists also in the quantum world 

At BER II, scientists have dis covered previously unknown symmetry properties 

in solid matter. The “golden ratio” is known from art and architecture. The 

researchers have now found its characteristics in the atomic structure of a 

crystal made of cobalt niobate. 

Exotic material state: “Liquid” quantum spins 

A team at the HZB has experimentally detected a so-called quantum spin 

liquid in a single crystal made of calcium chromium oxide. It is a new kind of 

state of matter. 

The huge advantage about the dis covery: According to popular beliefs, the 

quantum phenomenon should not have occurred in this material. The work 

extends the understanding of condensed matter and could also be important 

for the future development of quantum computers. 

A new condition of water: Like ice, but moveable 

Water is liquid at room temperature. But enclosed in the tiny channels of a 

zeolite structure, the water flows much tougher. A new state of matter of 

water in zeolite has now been discovered on the time-of-flight spectrometer 

NEAT at the neutron source BER II: in the nanochannels of the zeolite structure, 

the water molecules arrange themselves like in the ice crystal, but still 

remain as mobile as in a liquid. The inclusion in nanochannels enhances 

cooperative inter actions between water molecules. The results are important 

for the design of zeolite storage tanks, which are used as energy-saving air 

conditioning units for cooling. 

3-D imaging – first insight in magnetic fields 

3-D images are not only generated in medicine, for example with the help of 

X-ray or magnetic resonance imaging. Materials scientists also like to look 

inside a body. A team at the HMI has now succeeded for the first time in threedimensional 

representation of magnetic fields inside massive, non-transparent 

materials using polarized neutrons at the neutron source BER II. 

FACTS & FIGURES 

p Year of construction: 1972, 

reopening after renovation and approval: 1991 

p Termination of the operation: 

December 2019 (decision of the HZB supervisory board) 

p Type: open, light water moderated swimming pool reactor 

p Pool measurements: 

200 m 3 water capacity, two pools each 3.5 m in diameter and 

11 m deep linked by a channel 2 m wide 

p Delivery: 

p 10 MW of thermal power 

p about 2 x 10 14 neutrons per square centimetre 

and second in the core 

p Fuel elements: 24 standard elements each with 322 g of U-235 and 

6 elements for receiving the control rods each with 238 g of U-235 

p Control rods: 6 neutron absorbers 

p Reflector: 32 cm beryllium jacket 

ENERGY RESEARCH 

Transport processes in fuel cells 

How liquid water is distributed inside a fuel cell is crucial for its efficiency and 

service life. With neutron tomography at BER II, fuel cells can be analyzed in 

operando, i.e. while hydrogen and oxygen react to water. The scientists were also 

able to investigate the influence of membranes and different electrodes. 

Kesterite solar cells 

Kesterites are semiconductor compounds made up of several abundant elements. 

They can be used in solar cells to convert light into electrical energy. A team at the 

HZB produced kesterite samples and varied the composition. With neutron diffraction 

at BER II, they were able to determine how the different material 

com position affects defects and thus the efficiency of the solar cells. Further 

research showed that Germanium can improve the optoelectronic properties of 

the material. 

Batteries with silicon anodes 

In theory, silicon anodes could store ten times more lithium ions than the graphite 

anodes that have been used in commercial lithium batteries for many years. In 

practice, however, the capacity of silicon anodes drops sharply with every further 

charge-discharge cycle. A HZB team used neutron experiments at BER II and the 

Institut Laue-Langevin in Grenoble to clarify what happens on the surface of the 

silicon anode during charging and which processes reduce capacity: when 

charging, a blocking build-up occurs on the silicon surface layer that prevents the 

penetration of lithium ions. Now developers can specifically look for ways to 

break down or prevent this layer. 

HEALTH & LIFE 


ART & CULTURE 

Painting research: “Young Woman with a Dish of Fruit” 

The “Young Woman with a Dish of Fruit” was painted by Titian in Venice in 

the 16th century. The picture at BER II was examined with neutrons on behalf 

of the Gemäldegalerie. The neutrons stimulate the colour pigments so that 

the type of pigments can be inferred from them. The investigation revealed a 

surprise: Titian had already used Naples yellow for the girl’s gold- embroidered 

dress in 1555. This colour is only mentioned in the literature from 1702! It 

shows how far the powerful commercial power of Venice was internationally 

networked. (2001) 

Surprising finding in the snout of a fossil 

Scientists from the Natural History Museum Berlin have examined a petrified 

Lystrosaurus skull with neutron tomography at the HZB. This enabled them to 

create a three-dimensional image layer by layer, in which harder and softer 

components in the skull could be distinguished from one another. In the area 

of the snout they found traces of soft cartilage tissue, which indicate the 

existence of sinuses. A surprise, because the Lystrosaurus was already on the 

way to becoming a warm-blooded animal. 

How toxic proteins intend in nervous cells 

“Senile plaques” are found as typical deposits in the brains of deceased 

Alzheimer’s patients. However, these are probably not the cause, but rather the 

result of Alzheimer’s disease. Perhaps the plaques even serve as protection 

because they bind harmful proteins that would otherwise float freely. Smaller 

aggregates of the protein β-amyloid could be toxic. At BER II, a team with 

neutron diffraction investigated how β- amyloid can penetrate the membrane of 

nerve cells. The results made it possible to determine the position and mo bility 

of the toxic protein and confirmed the assumption that β-amyloid can penetrate 

nerve cells. 

Compatible joint prostheses 

In joints, the bones are equipped with cartilage and a layer of lipid membranes 

and move against each other in a liquid-filled capsule. This joint lubrication 

ensures painless mobility. At the neutron source BER II, researchers have 

investi gated this situation in a model system with synthetic lipid membranes and 

synthetic joint lubrication. They were able to measure how the distances between 

the individual lipid membranes of the “bone” coating increase with increasing 

temperature and how the surface of the artificial joint behaves under different 

pressure and shear forces. The results are interesting for the development of 

compatible joint prostheses. 





On the Scientific Utilisation 

of Low Power Research Reactors 

Pavol Mikula and Pavel Strunz 

At the previous conferences it has been reported about the effective utilisation of the Rez research reactor LVR-15 in 

basic, interdisciplinary and applied research. Now, in our contribution we will focus our attention on the scientific utilisation 

of the beam tubes at the low power research reactor. Namely, it will be reported about the neutron scattering 

instrumentation development and the educational possibilities at the low power neutron sources. The feasibility of 

carrying out the methodology and instrumental development research at the low power neutron sources will be demonstrated 

on designs of several high resolution and high luminosity neutron scattering instruments exploiting Bragg diffraction 

optics. Some of them have been already realized e.g. for small angle neutron scattering studies or residual 

strain/stress measurements. As the mentioned instrumental development and testing can be carried out at the low 

power neutron sources, due to the much lower safety requirements in comparison with the medium and high flux sources, 

they offer excellent educational and training programmes in neutron scattering or imaging for students. 

1 Introduction 

The present reactor LVR-15 was originally 

introduced in the operation in 

1957 at 2 MW power. Later on, after 

two reconstructions the present tank 

type light water reactor has used the 

uranium fuel enriched to 36 and 

finally 20 percent in uranium-235 and 

can operate at any power up to the 

licensed ceiling of 10 MW. It operates 

on average about 170 days per year 

with a pattern of operating cycles of 

three weeks plus one week for maintenance 

and instrumentation development. 

The thermal neutron flux in the 

core is the most of about 9x10 13 

n.cm 2 .s -1 (Table 1) and can be considered 

as a low power reactor. At 

present, it belongs to the Research 

Centre Rez, Ltd. and is operated 

mainly on a commercial basis. Research 

and development in Research 

Centre Rez, Ltd. is focused on the area 

of nuclear energy, nuclear reactor 

physics, chemistry and materials. The 

irradiation service uses the reactor 

namely for: Modification of Physical 

Charac teristics of Materials, Production 

of Radionuclides for Radiopharmacy 

and Production of Radionuclide 

Emitters. Crucial for research 

and development of the reactor are 

technological circuits – experimental 

loops for modelling of experimental 

conditions in the reactor core and the 

connected reactor cooling circuits. 

These loops allow mechanical, thermal-hydraulic, 

material, corrosion 

and further research at parameters 

and under operating conditions of the 

reactor concept under development. 

By placing a loop in the experimental 

reactor, all the above-mentioned 

physical and chemical influences of 

reactor coolant are supplemented by 

radiation conditions. The results are 

used in services for both Czech and 

foreign related organizations. On the 

other hand, Neutron Physics Laboratory 

(NPL) of Nuclear Physics Institute 

of the Czech Academy of Sciences performs 

effectively neutron physics experiments 

when using horizontal and 

vertical irradiation beam channels of 

the reactor [1, 2]. 

In total, NPL operates 8 instruments 

installed at 5 radial horizontal beam 

tubes (for experiments in nuclear 

physics, solid state physics and 

materials research) and two vertical 

irradiation channels (for neutron 

activation analysis) which are hired at 

Research Centre Rez, Ltd. A good 

Mean reactor power 

10 MW 

Maximum thermal neutron flux in the core 1∙10 18 n∙m -2 ∙s -1 

Maximum fast neutrons flux in the core 3∙10 18 n∙m -2 ∙s -1 

Maximum thermal flux in reflector 

(mix of Be + H 2 O) 

5∙10 17 n∙m -2 ∙s -1 

Maximum thermal neutron flux in the tubes 1∙10 12 n∙m -2 ∙s -1 

Maximum thermal flux 

at the exit of the tubes (100/60 mm) 

1∙10 8 n∙m -2 ∙s -1 

Irradiation channel - in fuel 1∙10 14 n∙m -2 ∙s -1 

Irradiation channel - at core periphery 7∙10 13 n∙m -2 ∙s -1 

Doped silicon facility 1∙10 13 n∙m -2 ∙s -1 

High pressure water loops 5∙10 13 n∙m -2 ∙s -1 

| Tab. 1. 

Reactor parameters. 

| Fig. 1. 

Schematic sketch of neutron scattering instruments installed at the reactor LVR-15. 


On the Scientific Utilisation of Low Power Research Reactors ı Pavol Mikula and Pavel Strunz


| Fig. 2. 

Photo of the experimental chamber used for the NDP and an example of the depth profiling of Boron in CaF 2 as implanted (390 keV B, 10 16 at. cm -2 ) and 

annealed at 600ºC. 

quality of the experiments carried out 

at the reactor in Řež is documented by 

the fact that NPL laboratory participated 

in the EU Project – ACCESS 

(Transnational Access to Large Facilities) 

in the frame of FP7-NMI3 programme 

which finished in January. 

2016. The following instruments are 

used at this low power research 

reactor at a good level (Figure 1): 

Two strain/stress scanners (HK4+ 

HK9), Small-angle neutron scattering 

(SANS) diffracto meter (HK8a), Neutron 

powder diffractometer MEREDIT 

(HK6), Thermal neutron depth profiling 

facility (HK3), Neutron activation 

analysis facility (NAA), Neutron 

optics diffractometer (HK8b). Effectiveness 

of the neutron scattering 

instruments is supported by employment 

of neutron optics devices in 

combination with position sensitive 

detectors (PSD). The powder diffractometer 

installed at the horizontal 

channel HK2 is operated by the 

Faculty of Nuclear Sciences and 

Physical Engineering of the Czech 

Technical University in Prague. 

2 Experimental activities 

at the reactor LVR-15 

2.1 Neutron depth profiling 

(NDP) 

NDP is the nuclear analytical technique 

available to determine depth 

profiles of light elements in solids 

(i.e., 3 He, 6 Li, 10 B, 14N, etc.). It utilizes 

the existence of isotopes of elements 

that produce prompt mono energetic 

charged particles upon capture of 

thermal neutrons. The related multidetector 

spectrometer consists of a 

large vacuum chamber, automatic target 

holders and several different data 

acquisition systems which can be used 

at the same time (Figure 2). From the 

energy loss spectra of emitted products 

the depth distri butions of light 

Nuclide 

Natural 

abundance 

or activity* 

[at/mCi] 

| Tab. 2. 

List of the NDP relevant isotopes. 

Nuclear 

reaction 

elements can be reconstructed. The 

NDP method is an excellent tool for 

studies of numerous problems in solidstate 

physics (diffusion, sputtering), 

material science (corrosion), electronics, 

optronics, life sciences, etc. Its 

applicability and efficiency has 

steadily expanded. This method uses 

the following parameters of the 

neutron beam: cross section – the 

height 4 mm and the width up to 

90 mm, intensity of the thermal 

neutron beam – 10 7 cm -2 s -1 , Cd ratio – 

10 5 , collimation – in the verical plane 

~1° and in the horizontal plane ~ 1°, 

beam homogeneity – inhomogeneous 

due to girland and zig-zag reflections. 

The list of the isotopes which can be 

used in the NDP method are shown in 

Table 2. Figure 2 shows also an 

example of the depth profiling of 

Boron in CaF 2 as implanted and after 

an anneling [3]. In general, NDP is a 

non- destructive method that leaves 

only trace amount of residual radioactivity, 

and examined samples can 

thus be measured repeatedly. Concentrations 

down to a ppm (with a 1D 

Cross 

section 

[barn] 

Energy 

of reaction 

products 

[keV] 

Detection 

limit 

[at/cm 2 ] 

3 He 0.13 x 10 -3 3 He(n,p) 3 H 5326 573 191 3.1 x 10 13 

6 Li 7.42 6 Li(n,a) 3 H 940 2051 2734 1.8 x 10 14 

7 Be* 2.5 x 10 14 7 Be(n,p) 7 Li 48000 1438 207 3.5 x 10 12 

10 B 19.6 10 B(n,γa) 7 Li 3606 1471 839 4.3 x 10 13 

10 B 19.6 10 B(n,a) 7 Li 230 1775 1014 6.7 x 10 14 

14 N 99.64 14 N(n,p) 14 C 1.81 584 42 9.1 x 10 16 

22 Na* 4.4 x 10 15 22 Na(n,p) 22 Ne 31000 2247 103 4.7 x 10 12 

33 S 0.76 33 S(n,a) 30 Si 0.14 3091 412 1.2 x 10 18 

35 Cl 75.5 35 Cl(n,p) 35 S 0.49 598 17 3.4 x 10 17 

59 Ni* 1.3 x 10 20 59 Ni(n,a) 56 Fe 12.3 4757 340 1.4 x 10 16 

List of the NDP relevant isotopes – detection limits are based on the charged particle counting rate 0.01 s -1 , 

detector → sample solid angle 0.03 Sr, and intensity of the neutron beam Φ th = 10 7 cm -2 s -1 . 

mode) or even ppb (with a 2D mode) 

level can be determined, depending on 

the element and the matrix. Pro filing 

to depths of about 15 mm (e.g. Li in 

metals) or even 60 mm (Li in polymers) 

can be obtained, with a depth resolution 

to a few nanometers only (for 

glancing angle geometry). The 

examined samples have to be solid (or 

liquid with very low volatility), flat 

with a smooth surface (with roughness 

of few nm only) and minimum 

area of at least a few mm 2 . Depending 

on the nuclides and the used substrates 

the analysis takes a few tens of minutes 

to a few tens of hours. The NDP 

technique is applicable only to the 

elements with a relevant cross- sections 

and energy of reactions [4]. 

2.2 Neutron Activation 

Analysis (NAA) 

Both short and long time irradiation 

for NAA can be carried out in vertical 

channels H1, H5, H6 and H8 of the 

LVR-15 reactor (Figure 3). Neutron 

fluence rates available in these 

channels is given in Table IV. For the 






Channel H1 H5 H6 H8 

Energy Fluence rate / n.cm -2 .s -1 

(0.0 – 0.501 eV) 3.38E+13 6.95E+13 5.98E+13 4.02E+13 

(0.501 eV – 10 keV) 1.49E+13 7.95E+13 6.80E+13 7.50E+12 

(10 keV – 0.1 MeV) 3.50E+12 2.12E+13 1.76E+13 1.81E+12 

(0.1 MeV. – 20 MeV) 1.08E+13 5.87E+13 7.16E+13 6.27E+12 

| Tab. 3. 

Neutron fluence rates in channels for NAA irradiation at the reactor LVR-15. 

short-time NAA the channel H1 is 

connected with the laboratory by a 

pneumatic system with the transport 

time of 3.5 s. Irradiation is carried out 

in a polyethylene (PE) rabbit for 10 to 

180 s. The channels H5, H6 and H8 

are used for long-time irradiation 

(0.5 h – several days) in 100 mm long 

Al-cans. In channels H5 and H8 

“ narrow” (inner diameter 35 mm) 

Al-cans are used, which accommodate 

up to 35 samples packed in disk 

shaped PE capsules, in channel H6 

“broad” (inner diameter 56 mm) Al 

cans are used, which accommodate up 

to 15 quartz vials with a 8 mm outer 

diameter. For Epithermal Neutron 

Activation Analysis (ENAA) both 

short- and long-time irradiation are 

performed behind a 1-mm Cd shield 

allowing for selective activation with 

epithermal neutrons. The laboratory 

is equipped with several high 

resolution and high efficiency HPGe 

coaxial detectors. Both relative and k 0 

– standardisation can be used for 

quantification of results as well as 

conventional g-ray spectrometry. 

The NAA methods provide a large 

variety of applications: Investigations 

of environmental and historical 

materials (determination of up 40 

elements in aerosol, fly ash, soil, 

sediment, etc., samples by a combination 

of Instrumental Neutron 

Activation Analysis (INAA) and 

ENAA) [5], geo- and cosmochemical 

samples (elemental characterization 

of rocks, tektites, namely moldavites, 

and meteorites by a combination 

of INAA, ENAA, and Radiochemical 

Neutron Activation Analysis (RNAA)), 

in biomedicine (determination of 

essential and toxic trace elements in 

selected human and animal tissues by 

a combination of INAA and RNAA to 

achieve the lowest element detection 

limits possible), in forensic science 

(determination of poisonous elements 

in selected tissues of investigated 

cases of contemporary and historical 

persons) and in chemical metrology 

[6] (certification of element contents 

in reference materials prepared by the 

most important producers, such as 

U.S. NIST, IRMM, IAEA, etc.). From 

the recent NAA investigations, let us 

introduce several of them. INAA was 

used to determine contents of more 

than 30 elements in meteorites 

Morávka [7] and Jesenice [8]. 

Environ mental research was focused 

on the determination of 129 I and 

the 129 I/ 127 I ratio in biomonitors, 

namely, in bovine thyroid and moss, 

collected in the vicinity of the Temelín 

nuclear power plant (NPP) in south 

Bohemia using NAA in several modes 

(NAA with pre-irradiation separation 

followed by RNAA, and ENAA). No 

significant differences of 129 I levels 

and the 129 I/ 127 I ratios in the thyroids 

collected prior to the start and after 

several years of the NPP operation 

have been indicated [9]. For agricultural 

and nutritional research, we 

used a RNAA procedure to study the 

Se-transfer from soil or seed to wheat 

plants [10] and the ability of bread 

and durum wheat to accumulate Se 

via a soil-addition procedure at 

sowing time [11] to increase the 

desired uptake of the element in the 

Portuguese population. Silicon is an 

important trace element in humans, 

because it reduces the absorption of 

aluminium in human gastrointestinal 

tract. The daily intake of silicon should 

be about 10–25 mg, and its most 

readily absorbable form is H 4 SiO 4 , 

which is contained in beer. Using 

INAA, we found that Si-concentrations 

in Czech lager beer(s) varied in 

the range of 13.7 to 44.2 mg L -1 [12]. 

Concerning the cultural heritage, 

in 2010, the grave of the famous 

astronomer Tycho Brahe was opened 

by a Czech-Danish research consortium 

and samples of his bones, 

hair, and teeth were procured for 

scientific investigation. We carried out 

mercury determination in segmented 

hair samples by RNAA. The results 

showed that in the last 2 months of 

Brahe's life, he was not exposed to 

lethal (or fatal) doses of mercury, 

as was previously speculated [13]. 

Furthermore, graphene is another 

example of a material difficult to assay 

by classical analytical techniques. 

Therefore, elemental impurities were 

determined by INAA in graphene 

samples prepared by various oxidation 

procedures of graphite to graphite 

oxide followed by various reduction 

processes [14]. On the corresponding 

website one can find many other NAA 

results usually taken in international 

collaboration. 

2.3 Neutron powder 

diffraction 

The medium resolution powder 

diffracto meter (MEREDIT) installed 

at the beam channel HK6 consists 

Monochromator Reflection Wavelength 

Å 

Minimum 

Dd/d 

(x 10 -3 ) 

Neutron 

flux 

n.cm -2 .s -1 

Beam 

size 

cm 2 

3 bent Si single crystals (422) 1.271 3.9 (at 56° 2θ) ~8.8 x 10 5 2 x 4 

(311) 1.877 4.4 (at 59° 2θ) ~8.6 x 10 5 2 x 4 

3 mosaic Cu crystals (220) 1.460 4.9 (at 71° 2θ) ~3.6 x 10 6 4 x 4 

| Fig. 3. 

LVR-15 active core layout. 

| Tab. 4. 

Monochromators and beam parameters. 




| Fig. 4. 

Powder diffraction spectrum of La 9.33+x (Si 1-y MyO 4 ) 6 O 2+3x/2 ; M = {Fe, Al, Mg}. 

basically of 3 changeable monochromators 

placed in a massive shielding, 

two large HUBER goniometer circles 

and a multi-detector bank which 

is mounted in a moulded neutron 

shielding made from boron carbide 

powder in epoxy resin. The bank 

contains 35 3 He counters with 

corresponding 10’ Soller collimators. 

The detector bank moves on air pads, 

which provide together with the 

stepping motor smooth positioning 

of this heavy loaded bank. Diffraction 

patterns can be collected in the 

angular range from 2° to 148° in 2 θ S 

with the step down to 0.02° and step 

delay controlled by strict time or 

neutron flux read by a monitor. Monochromator 

and beam parameters are 

shown in Table 4. The diffractometer 

is mainly used for non destructive 

structure phase identification, crystalline 

structure determination, magnetic 

structure determination, temperature 

dependent phase transition, 

quantitative multi-phase analysis and 

also for in-situ internal stress-strain 

evolution. The following sample 

environments are at a disposal: close 

cycle cryostat for 10 K → 300 K, 

vacuum furnace for 300 K → 1,300 K, 

light furnace for 300 K → 1,300 K, 

Euler goniometer, automatic 6 

samples exchanger for RT and a 

deformation rig. As an application 

example, Figure 4 shows diffraction 

spectrum serving for identification 

of deformation of oxygen ion conductive 

channels (Lanthanum silicates 

La 9.33+x (Si 1-y M y O 4 ) 6 O 2+3x/2 ; M = {Fe, 

Al, Mg} with apatite like crystal structure 

with space group P6 3 /m 

are interesting material due to the 

high oxygen ion conductivity for 

fuel cell applications) and Figure 5 

shows the result of the nondestructive 

phase analysis of the Roman cavalry 

helmet from 2 nd century A.D., where 

phase analysis of the surface corrosion 

products and an estimation of Zn 

content to be of 18 wt %. in the brass 

material was carried out [15]. 

2.5 Small-angle neutron 

scattering 

Small-angle neutron scattering investigations 

are carried out on the 

double- crystal diffractometer MAUD 

designed for the measurements in the 

high momentum transfer Q-resolution 

range. In contrast to conventional 

double- crystal arrangements, the 

fully asymmetric diffraction geometry 

on the elastically bent Si analyzer is 

employed to transfer the angular 

distribution of the scattered neutrons 

to the spatial distribution and to 

analyze the whole scattering curve 

by a one-dimensional position sensitive 

detector (Figure 6) [16]. It 

reduces the exposition time per 

sample typically to 0.5 to 5 hours 

( depending on the Q-resolution and 

| Fig. 5. 

Powder diffraction spectrum from a fragment of the Roman cavalry helmet. 

sample cross- section). The remote 

control of the curvatures of the monochromator 

and analyzer crystals 

makes possible to tune the instrument 

resolution in the DQ range from 10 -4 

to 10 -3 Å -1 , according to the expected 

size of investigated inhomogeneities. 

An absolute calibration of scattering 

cross- sections is possible by measuring 

the intensity of the direct beam 

(no calibration samples are required). 

The instrument operates in fully 

automatic mode, including a sample 

exchanger. The SANS diffractometer 

is in our case mainly used for studies 

of inho mo geneities in the size range 

from 50 nm to 1,000 nm i.e. large 

precipitates in alloys (superalloys), 

porous materials (superplastic 

ceramics, ceramic thermal barrier 

coatings), nano-particles in ceramicintermetallic 

compounds (MoSi 2 

with Si 3 N 4 and SiC particles) and 

large inhomogeneities in polymers/ 

microemulsions (dimethyl-formamide-cyclohexan 

domains segra gated 

by diblock copolymer). As application 

| Fig. 6. 

Schematic sketch of the double- crystal SANS diffractometer operating in combination with PSD. 






| Fig. 7. 

Nanoporosity development 

in metalic membrane. 

| Fig. 8. 

Precipitate dissolution in CMSX4 singlecrystal nickel-based superalloy. 

examples Figure 7 and Figure 8 show 

the results of studies of the nanoporosity 

in metallic membrane (where 

the aim of the experiment was to 

determine a dependence of the pore 

depth on the etching time by using 

SANS) and in-situ studies of hightemperature 

microstructure (precipitate 

dissolution in CMSX4 

single-crystal nickel-based superalloy 

was investigated), respectively [17, 

18]. 

2.6 Strain/stress scanning in 

polycrystalline materials 

The dedicated two-axis diffractometer 

installed at the channel HK4 is 

equipped with bent Si and Ge perfect 

single crystal monochromators which 

are easily changeable according to 

the experimental requirements. The 

diffractometer is usually used for 

macro strain scanning of polycrystalline 

materials. An easy change of the 

instrument parameters permits one to 

use it also for another type of experiments, 

e.g. Bragg diffraction optics 

experiments. The diffractometer uses 

advantages coming from focusing 

both in real and momentum space and 

yields good resolution and luminosity, 

especially for samples of small dimensions 

[19]. The resolution properties 

of the device are reached in a limited 

range of momentum transfer for 

which the focusing conditions are 

optimized. The corresponding optimization 

can be done easily by using 

a remote control of the curvature of 

the monochromator. In the case of 

the strain scanning of the sample, the 

gauge volume is determined by two 

fixed Cd slits (2 to 5) mm x (3 to 30) 

mm in the incident and diffracted 

beams and the measurements are 

performed in the vicinity of the 

scattering angle of 2q S = 90°. For 

scanning the sample a x-y-z stage or 

ABB robot system (see Figure 9) can 

be used. The instrument is controlled 

by PC. The diffractometer has a 

changeable monochromator take-off 

angle and can be set and operate at a 

suitably chosen – neutron wavelength 

in the thermal neutron range from 

0.1 nm to 0.235 nm. In the case 

of a-Fe and g-Fe samples it usually 

operates at the neutron wavelength of 

0.235 nm, when providing a maximum 

detector signal and good resolution 

after diffraction on a-Fe(110) and/or 

g-Fe(111) lattice planes. By recent 

installation of the 2D-PSD (20 x 

20 cm 2 , 2 mm spatial resolution), the 

acquisition of the data has been 

increased by a factor of 4. Depending 

on the sample-detector distance and 

the required resolution the PSD detector 

can cover from 10° to15° of the 

scattering angle 2q S . The quality of 

the instrument are supported by 

the experimental results of stress 

measurements obtained on the 

welded test-sample shown in Figures 

10 and 11 [20]. The aim of the 

performed residual stress studies was 

to find optimum composition of the 

additive material for electrodes in 

order to decrease residual stresses in 

the vicinity of the foot of the welding 

joint and consequently, to increase the 

fatigue strength. For samples we used 

parent material Weldox 700/S690QL 

and X2CrNi12 (1.4003) and D4-6547 

filler for electrodes. 

2.7 High resolution diffraction 

for materials research 

Another high-resolution two-axis 

diffractometer optimized for investigation 

of elastic and plastic deformation 

studies in polycrystalline 

materials is installed at the channel 

HK9. The instrument is used especially 

for in-situ thermo-mechanical 

testing of materials, i.e. to study the 

deformation and transformation 

mechanisms of modern types of 

newly developed materials. Neutron 

diffraction performed in situ upon 

external loads brings a wide range of 

valuable structural and sub-structural 

parameters of the studied material 

which is easy to correlate with 

the parameters of the external load. 

The obtained microstructural parameters 

of the examined material 

can be directly compared with the 

para meters of micromechanical models. 

This approach brings a deeper understanding 

of processes ongoing in 

materials upon deformations, thermal 

treatments or phase transformations. 

The instrumental parameters are as 

| Fig. 9. 

Photo showing the ABB robot system and 

2D-PSD detector. 

| Fig. 10. 

Photo of the fatigue test specimen and its dimensions. 




| Fig. 11. 

Residual stress distribution in the vicinity of the welds welded by D4-6547 filler material. Parent materials: Weldox 700/S690QL and X2CrNi12 (1.4003). 

follows: Horizontally and vertically 

focusing monochromator employing 

elastically bent Si single crystals, neutron 

wavelength – 1 Å ≤ l ≤ 2.7 Å, 

neutron flux at the sample position – 

10 5 n.cm 2 .s -1 at l=2.3 Å, angular 

range of scattering angles – 25°< 

2q S < 90° and resolution – 2·10 -3 ≤ 

Dd/d ≤ 3·10 -3 . The following sample 

environments are at a disposal: Two 

deformation rigs for uni-axial loading 

(tension or pressure; ± 20 kN and 

± 60 kN), resistance heating (T < 

1,200 °C) or hot-air heating (T < 

300 °C), miniature deformation 

machine for uni-axial loading (tension, 

pressure; ± 10 kN) inside an 

Eulerian cradle, Eulerian cradle 

( inner diameter of 400 mm, 0°< 

c



The Performance of Low Activation Steel 

SCRAM on ACPs Source Term in Watercooled 

Loop of Fusion Reactor ITER 

Weifeng Lyu, Jingyu Zhang and Shouhai Yang 

In water-cooled loops of International Thermonuclear Experimental Reactor (ITER), most Occupational Radiation 

Exposure (ORE) of personnel is due to Activated Corrosion Products (ACPs) in the cooling loops. The corrosion products 

come from the corrosion of water on steel used in the cooling loops. In order to reduce neutron activation of steel and 

the corresponding ORE, the Super-clean Reduced Activation Martensitic (SCRAM) steel is recently developed in China 

to replace the traditional steel SS316, whose performance on ACPs source term needs to be analyzed. In this paper, the 

corrosion rate of SCRAM under fusion reactor operation condition was measured using a high-temperature flowingwater 

corrosion experiment loop, which was introduced into ACPs source term analysis code CATE. Then based on 

LIM-OBB cooling loop of ITER, ACPs activity and dose rate of SCRAM were calculated and compared with that of SS316. 

The calculation results showed that during the operation phase of the reactor, SCRAM produced higher activity and 

dose rate than SS316 due to its bad corrosion-resistance, while after shutting down the reactor, SCRAM performed 

better on ORE decrease than SS316 due to its good activation-resistance. 


According to the surveillance data of 

French PWR plants, more than 90 % 

of occupational radiation exposure 

(ORE) of personnel under normal 

operation is due to the activated corrosion 

products (ACPs) in the primary 

coolant circuit [1]. And for the watercooled 

loops of fusion reactor ITER, 

the gamma ray from ACPs is also a 

major contributor to ORE [2]. 

In the water-cooled loops of ITER, 

the corrosion is caused by the contact 

of water and metal material. For 

instance, the pipe material of heat 

exchanger is corroded by the coolant 

and plenty of corrosion products 

(CPs) are produced. Some CPs are 

released into coolant, and transported 

to the regions under neutron radiation 

by the coolant, such as first wall, 

blanket, divertor, vacuum chamber, 

etc. Here, these CPs absorb neutrons 

and become radioactive, which are 

called activated corrosion products 

(ACPs). Some ACPs are transported to 

the regions out of neutron radiation 

by the coolant, such as heat exchanger, 

pipe, valve, pump, filter and so on, 

where parts of ACPs adhere to the 

internal surface of the pipe and continuously 

decay and emit gamma ray. 

When the workers inspect or repair 

these devices, they have to bear the 

radiological dose. 

As we know, the fission products 

do not exist in the fusion reactor, so 

ACPs become the dominant source 

term and should be reduced as 

much as possible. Neutron-induced 

activation of metal components in 

fusion reactor can be effectively 

controlled by proper selection of 

structure materials [3]. For meeting 

the demand of low activation, the 

Super- clean Reduced Activation 

Martensitic (SCRAM) steel is recently 

developed in China. Considering that 

it is martensitic steel, although it is 

developed to obtain reduced activation, 

it may cause more serious corrosion 

problem and increase ACPs and 

ORE, compared with the traditional 

austenitic steel SS316. This paper 

focuses on studying the influence on 

ACPs and ORE from using SCRAM and 

SS316. 

The description of the material, the 

calculation method and code are 

presented in the following two 

sections. The description of ITER 

LIM-OBB loop and the corresponding 

calculation results and discussions 

are presented in the fourth section. 

In the last section, a comprehensive 

comment is given. 

2 Description of material 

2.1 The material composition 

In the files of ITER technical basis [4], 

SS316 is claimed as structure material 

widely used in first wall, blanket, 

divertor, etc. But as we know, the 

neutron activation of SS316 is serious. 

So, for decreasing the material radioactivity 

to a level where economical 

waste disposal or recycling is feasible 

in


Parameter 

Value 

Average coolant 

temperature (°C) 

Dissolved oxygen 

contents (mg/kg) 

Coolant pH 

Pressure (MPa) 1 

Average coolant 

velocity (m/s) 

The experiment contained seven 

time points for sample measurement, 

which are 100 h, 200 h, 400 h, 800 h, 

1,000 h, 1,200 h and 1,500 h. At these 

time points, parts of the samples were 

taken out of the experiment loop, and 

then several times cleaning respectively 

with hydrochloric acid, acetone 

and water were performed to dissolve 

and remove the corrosion products in 

the samples. After drying process, the 

residual mass of base metal in the 

samples was weighed. The corrosion 

rate can be measured through monitoring 

the mass decrease of base metal 

in the samples, and the power model 

of non-linear regression [7] was 

adopted to fit the curve of corrosion 

rate, as follows. The corrosion rate of 

SS316 is quoted from Reference [8]. 

It can be seen that the corrosion rate 

of SCRAM is obviously higher than 

SS316, which is an expected weakness 

of ferritic/martensitic steel compared 

with austenitic steel. 

SCRAM: 

SS316: 

150 

less than 0.01 

7 (20 °C) 

| Tab. 2. 

The operation parameters of the experiment 

loop. 

3 Method and code 

of calculation 

(1) 

(2) 

3.1 The ACPs calculation 

The code CATE [9] is developed by 

North China Electric Power University. 

It is capable to analyze the nuclide 

composition and spatial distribution 

of ACPs along the cooling loops. The 

European Activation File EAF-2007 

[10,11] is introduced into CATE, 

which includes the nuclear data of 

2231 nuclides and makes CATE 

capable to calculate any activation 

product of any material. 

The simulation of ACPs transport 

in CATE is based on the theory that 

6 

| Fig. 1. 

Schematic of ACPs transport in the cooling loop. 

the main driving force is the temperature 

change of the coolant 

throughout the loop and the resulting 

change in metal ion solubility in 

the coolant, which is presented in 

Figure 1. 

The pipe surfaces with high neutron 

flux and resulting high temperature, 

such as first wall, blanket, 

diverter, are named “In-Flux” surface 

node, while the other pipe surface 

without neutron flux and with relative 

low temperature, such as pipe, pump, 

valve, heat exchanger, are named 

“Out-Flux” surface node. Considering 

the velocity of coolant is as fast as 

6 m/s, ACPs in the coolant will be 

mixed rapidly, so it can be assumed 

that the coolant is a homogeneous 

node. And the model above for ACPs 

transport is named three-node model. 

3.2 Dose rate calculation 

In the field of radiation protection, 

dose rate is an intuitive parameter to 

evaluate material behavior. The 

ARShield code is developed by North 

China Electric Power University to 

calculate the dose rate caused by 

ACPs. ARShield is an advanced version 

of the point kernel integration code 

QAD-CG developed by Los Alamos 

National Laboratory. It provides the 

pre-job for visualization of large-scale 

radiation field and virtual roaming in 

nuclear plant, by breaking the restrictions 

of the traditional point kernel 

integration code. The detailed characteristics 

of ARShield can be seen from 

Reference [12]. 

The composition and activity of 

ACPs calculated by CATE are introduced 

into ARShield, and then the 

dose rate is calculated using the point 

kernel integration method, which is as 

follows. 

(3) 

where, 

r, point at which gamma dose 

rate is to be calculated; 

r', 

location of source in volume 

V; 

V, volume of source region; 

μ, total attenuation coefficient at 

energy E; 

, distance between source point 

and point at which gamma 

intensity is to be calculated; 

K, flux-to-dose conversion factor; 

B, dose buildup factor. 

4 Calculation of activity 

and dose rate of ACPs 

in ITER LIM-OBB loop 

4.1 Description of 

ITER LIM-OBB loop 

The schematic of the cooling loop 

is shown in Figure 2. The main equipment 

includes blanket module, heat 

exchanger, hot leg pipe, cold leg pipe, 

pump, resin and filter. The blanket 

module belongs to the In-Flux region, 

others belong to the Out-Flux region. 

The main design parameters of 

ITER LIM-OBB loop is presented in 

Reference [8]. The calculations are 

based on the ITER SA1 operation 

scenario [13], corresponding to 432 

full power day operation with several 

dwell and burn periods. Because of 

the limitation of electric field and 

magnetic field, the plasma can’t 

sustain for a long time, and has to be 

operated under pulse mode. In the 

calculation process of neutron activation, 

the time step should be smaller 

than the pulse time, which makes 

the simulation time-consuming. In 



The Performance of Low Activation Steel SCRAM on ACPs Source Term in Water- cooled Loop of Fusion Reactor ITER ı Weifeng Lyu, Jingyu Zhang and Shouhai Yang



| Fig. 2. 

Schematic of ITER LIM-OBB loop. 

Results SCRAM SS316 

CPs mass on In-Flux surface 

(kg) 

CPs mass on Out-Flux surface 

(kg) 

ACPs activity on In-Flux surface 

(Bq/m 2 ) 

ACPs activity on Out-Flux surface 

(Bq/m 2 ) 

Contact dose rate of Out-Flux region 

(mSv/h) 

Contact dose rate of Out-Flux region 

after shutdown for 10 days (mSv/h) 

| Tab. 3. 

The calculation results of ACPs in ITER LIM-OBB loop. 

this paper, the continuous pulse 

method (CP) [14] is adopted to treat 

the pulses. The CP method is assumed 

to consist of a continuous irradiation 

period followed by only several pulses 

before shutdown, which has been 

proved accurate and efficient. 

4.2 The calculation results 

and discussions 

The calculation results of ACPs activity 

in ITER LIM-OBB loop with CATE 

code are presented in Table 3. 

The structure materials in the 

In-Flux region lie well within the 

biological shield, and they pose no 

direct radiation hazard to operating 

personnel. While the equipments 

and loops in the Out-Flux region 

become more important, because 

the workers have to access to them 

for periodic inspection and maintenance, 

and exposed to γ radiation 

3.087E+01 

4.599E+01 

2.225E+12 

4.557E+09 

3.621E+00 

7.852E-02 

2.078E+01 

2.817E+01 

1.700E+12 

3.233E+09 

2.314E+00 

1.422E-01 

from ACPs there, which may be a 

major contributor to ORE [15]. 

Therefore, only the dose rate of 

Out-Flux region is calculated, as 

follows in Table 4. 

From the above tables, some 

conclusions are drawn as follows. 

(1) During the operation phase of the 

reactor, the mass of corrosion 

products produced by SCRAM is 

more than SS316, which means 

the corrosion resistance of the 

traditional austenitic steel SS316 is 

better than SCRAM. 

(2) During the operation phase of 

the reactor, the activity of ACPs 

produced by SCRAM is more than 

SS316. This trend is consistent 

with the mass of corrosion 

products, which means ACPs 

activity is determined by the mass 

of corrosion products to some 

degree. 

(3) During the operation phase of 

the reactor, the contact dose rate 

of Out-Flux region produced by 

SCRAM is more than SS316. But 

during shutting down the reactor 

for 10 days, the contact dose rate of 

Out-Flux region produced by 

SS316 is more than SCRAM. This 

means SCRAM can present advantage 

on ORE decrease during the 

shutdown phase rather than the 

operation phase, compared with 

the traditional austenitic steel 

SS316. 

(4) The fast decrease of dose rate 

of SCRAM after shutdown is due 

to its nuclide composition of 

ACPs. SCRAM doesn’t contain 

Co element, and thus activation 

products of Co (long-lived nuclides 

Co-57, Co-58 and Co-60) do not 

exist in ACPs. But for SS316 

with Co element, the activation 

products of Co, especially Co-58, 

contribute a large part to the dose 

rate, and result in the relative slow 

decrease of dose rate after shutdown. 

4 Conclusions 

In this paper, firstly the corrosion 

rate of Super-clean Reduced Activation 

Martensitic (SCRAM) steel 

under ITER operation conditions was 

measured using a high-temperature 

flowing-water corrosion experiment 

loop. Then the model of ITER 

LIM-OBB loop was simulated by the 

ACPs source term analysis code CATE, 

and the nuclide composition and 

space distribution of ACPs of SCRAM 

and SS316 were calculated. These 

results were introduced to the dose 

rate analysis code ARShield, and 

the contact dose rate of Out-Flux 

region in ITER LIM-OBB loop caused 

by ACPs was calculated. At last, some 

comparisons among SCRAM and 

SS316 were made from the point of 

view of ORE. The results showed that 

compared with SS316, SCRAM produced 

higher dose rate during 

the operation phase of the reactor 

because of the bad corrosion resistance, 

but during the shutdown 

phase, it presented advantage on 

ORE decrease due to its good activation 

resistance. 

Acknowledgments 

The authors would like to express 

their gratitude for the support: Project 

11605058 supported by National 

Natural Science Foundation of China 

and Project 2017MS041 supported by 

the Fundamental Research Funds for 

the Central Universities. 




Nuclide 

References 

Half-life 

(h) 

[5] A. Rocher, J. L. Bretelle, M. Berger: Impact of main radiological 

pollutants on contamination risks (ALARA) 

optimization of physico chemical environment and retention 

technics during operation and shutdown, in Proceedings of 

the European Workshop on Occupational Exposure 

Management at NPPs (ISOE 04), Session 2, EDF, Lyon, France, 

March 2004. 

[6] C.B.A. Forty, P.J. Karditsas: Preliminary cooling circuit 

activation and ORE assessment for ITER, Paper presented 

at 19 th SOFT, Lisbon, September16-20, 1996. 

[7] Q. Huang, J. Li, Y. Chen: Study of irradiation effects in China 

low activation martensitic steel CLAM, Journal of Nuclear 

Materials, vol. 329, pp. 268-272, 2004. 

[8] ITER Group: ITER technical basis, ITER EDA Documentation 

Series No. 24, IAEA, Vienna, 2002. 

[9] T. Muroga, M. Gasparotto, and S.J. Zinkle: Overview of 

materials research for fusion reactors, Fusion Engineering 

and Design, vol. 61, pp. 13-25, 2002. 

[10] Y. Wen, S. Jin, Z. Yang, F. Luo, Z. Zheng, L. Guo, J. Suo: 

Positron beam Doppler broadening spectra and nanohardness 

study on helium and hydrogen irradiated 

RAFM steel, Radiation Physics and Chemistry, vol. 107, 

pp. 19-22, 2015. 

[11] V. Belous, G. Kalinin, P. Lorenzetto, S. Velikopolskiy: 

Assessment of the corrosion behaviour of structural materials 

in the water coolant of ITER, Journal of Nuclear Materials, 

vol. 258, pp. 351-356, 1998. 

[12] P. J. Karditsas: Activation product transport using TRACT: 

ORE estimation of an ITER cooling loop, Fusion Engineering 

and Design, vol. 45, no. 2, pp. 169-185, 1999. 

[13] J. Zhang, L. Li, Y. Chen: Application of CATE 2.0 Code on 

Evaluating Activated Corrosion Products in a PWR Cooling 

Loop, ATW-International Journal for Nuclear Power, vol. 62, 

no. 3, pp. 181-185, 2017. 

[14] R.A. Forrest: The European Activation File: EAF-2007 decay 

data library, UKAEA FUS 537, 2007. 

[15] R.A. Forrest, J. Kopecky, J-Ch. Sublet: The European Activation 

File: EAF-2007 neutron-induced cross section library, UKAEA 

FUS 535, 2007. 

[16] S. He, Q. Zang, J. Zhang, H. Zhang, M. Wang, Y. Chen: 

Development and Validation of an Interactive Efficient 

Dose Rates Distribution Calculation Program Arshield for 

Visualization of Radiation Field in Nuclear Power Plants, 

Radiation Protection Dosimetry, vol. 174, no. 2, pp. 159-166, 

2017. 

[17] ITER-JCT: ITER Generic Site Safety Report (GSSR), 

Volume III: Radiological and energy source terms, 

ITER G84 RI 3 01-07-13 R 1.0, 2001. 

[18] J. Sanz, O. Cabellos, P. Yuste, S. Reyes, J.F. Latkowski: 

Pulsed activation of structural materials in IFE chambers, 

Fusion Engineering and Design, vol. 60, no. 1, pp. 45-53, 

2002. 

[19] C.B.A Forty, J.D. Firth, G.J. Butterworth: Influence of materials 

choice on occupational radiation exposure in ITER, Journal of 

Nuclear Materials, vol. 258, pp. 335-338, 1998. 

Contact dose rate 

(mSv/h) 

Authors 

Weifeng Lyu 

Shouhai Yang 

Senior engineers 

State Key Laboratory of Nuclear 

Power Safety Monitoring 

Technology and Equipment, 

Shenzhen, Guangdong, 518172, 

China 

Jingyu Zhang 

Associate professor 

School of Nuclear Science and 

Engineering, North China Electric 

Power University 

No.2, Beinong Road, Changping 

District, Beijing 102206, China 

Contact dose rate 10 days after shutdown 

(mSv/h) 

SCRAM SS316 SCRAM SS316 

MN-56 2.579E+00 3.518E+00 1.907E+00 -- -- 

NI-57 3.560E+01 -- 1.805E-01 -- -- 

V-52 6.238E-02 1.809E-03 2.470E-03 -- -- 

CR-51 6.648E+02 6.614E-03 7.827E-03 4.412E-03 4.257E-03 

MN-54 7.489E+03 8.692E-02 8.410E-02 7.282E-02 5.746E-02 

CO-57 6.522E+03 -- 1.447E-03 -- 9.851E-04 

CO-58 1.701E+03 -- 1.135E-01 -- 7.189E-02 

CO-60 4.618E+04 -- 8.450E-03 -- 5.882E-03 

FE-59 1.068E+03 1.720E-03 1.407E-03 1.074E-03 8.408E-04 

| Tab. 4. 

The main contributors to dose rate. 






Physical properties 

Inner diameter surge-line 

Outer diameter surge-line 

Length of surge line center arc 

Length of main pipe section 

Outer diameter main-pipe 

Inner diameter main-pipe 

Inner diameter surge-line 

Outer diameter surge-line 

Fluid Structure Interaction Analysis 

of a Surge-line Using Coupled CFD-FEM 

Muhammad Abdus Samad, Xiang bin li and Hong lei Ai 

The mixing with different-temperature water in the pressurizer surge line may result in thermal stratification, then 

the significant deformation of the solid structure due to different thermal expansion at different parts of the structure 

perhaps occur, which will be a threat for the plant safety. To better understand the coupling mechanism, the 

corresponding characteristics in a pressurizer surge line is analyzed using CFD software (ANSYS CFX) and FEM solver, 

(ANSYS MECHANICAL). The fluid temperature distribution is calculated first, then the corresponding thermal and 

mechanical characteristics are analyzed. It is found that a large steady state stress present at the edges of the main pipe 

and the pressurizer, the consequent deformation showed large displacement at the center of the surge line. 

Introduction 

The surge line is the pipe that connects 

the pressurizer with the hot leg of the 

primary loop. As the controlling of the 

pressure takes place in the pressurizer, 

surge line acts as the in between of 

main pipe and the pressurizer. As a 

consequence of pressure control 

the surge line experiences thermal 

stresses along its life time. These 

stresses are often cyclic in nature 

given the nature of the load that the 

power plant experiences, such cyclic 

stress over time cause material fatigue 

and in worst cases can cause significant 

damage and are therefore, a big 

factor for plant safety design. The 

deformation of these line can cause 

rupture and the subsequent leakage 

may have undesirable effects on plant 

working. 

There have been many reports on 

the damage of piping in PWR plants 

due to thermal stresses. In the US 

284.20 mm 

360 mm 

19.187 m 

4.23 m 

870 mm 

736 mm 

284.20 mm 

60 mm 

there have been reports from Trojan 

plants regarding unusually large 

piping displacements due to thermal 

stratification, which resulted in 

crushed insulations, decreased gaps 

among rupture restraints and heavier 

pipe support loads, Beaver valley 2 

also had unexpected pipe displacements 

which caused the snubbers to 

stroke out. In Slovakia the surge line 

elbow at Bouhnice 3 had to be 

replaced because the calculated 

cumulative fatigue usage factor was 

high [1]. Piping in PWR plants have 

been undergoing unwanted thermal 

stress for quite a long time, the reports 

of unwanted movement in pipes as a 

result of inadequate calculations of 

| Fig. 1. 

Mesh of surge line structure. 

material and fluid interaction have 

been available in the literature from as 

early as 1995 when PWR plants in 

France reported experiencing thermal 

stratification due to the geometry 

which were un-accounted for in the 

design calculations. This stratification 

continued in steady state and the 

stresses were calculated by 1d-2d 

method developed by FRAMOTOME 

[2]. The same year a German PWR 

presented its own study on the existence 

of stratification in PWR reactors 

especially in the horizontal regions, in 

his paper they used ADINA code to 

calculate the stress in the surge line 

[3]. The Atomic Regulatory Board of 

India worked on developing an 

Analytical model for induced stress 

using intermixing layer they validated 

their model by testing it on a surge 

line [4]. In recent literature Korean 

Institute of Nuclear Safety worked on 

these stresses present in Surge lines in 

detail and performed several analysis 

to calculate the thermal stress in 

in-surge out-surge cases using commercially 

available ANSYS codes [5]. 

Also similar techniques were used 

at Beijing university of Chemical 

Engineering, Harbin University of 

Engineering, Xian Jiao tang University 

to evaluate thermal stresses and the 

consequent effects on the surge line 

[6–8]. 

| Tab. 1. 

Physical properties of surge line. 

Mesh 

Id 

Maximum 

Element Size 

Mesh 

cells 

Average 

Mesh quality 

1 0.003 9011907 0.85209 

2 0.004 3966425 0.85091 

3 0.006 1371084 0.84793 

4 0.008 739633 0.84512 

| Tab. 2. 

Sensitivity Analysis. 

| Fig. 2. 

Mesh of fluid and solid structure. 


Fluid Structure Interaction Analysis of a Surge-line Using Coupled CFD-FEM ı Muhammad Abdus Samad, Xiang bin li and Hong lei Ai


Although much work has been 

done on the different cases of transient 

there is severe lack of work on steady 

state of thermal stratification present 

in the surge line as first observe in 

France. In this paper conjugate heat 

transfer analysis is performed on a 

surgeline PWR in ANSYS CFX, then 

the steady state temperature profile is 

then transferred to ANSYS Mechanical 

to calculate the stress acting on the 

surge line in steady state. 

Model 

Physical Model 

As shown in the Figure 1, the concerned 

structure is a pipe of diameter 

360 mm connected to main pipe with 

a diameter 870 mm perpendicularly. 

The material of the pipe is stainless 

steel and it has physical parameters as 

defined in Table 1. 

For this simulation the working 

fluid is water. The water in surge line 

comes from the pressurizer where the 

temperature is around 270 °Celsius. 

While cold water at 120 °C flows 

through the main pipe at an average 

wave velocity of 15.6 m/s. The flow in 

the surge line is taken as 0.1 m/s 

towards the main pipe. In this study 

the steady state analysis is done on the 

surge line so the initial condition of 

the pipe are taken as the temperature 

of the main pipe. The simulation was 

tested in increasingly refined mesh to 

test the independence of the mesh. 

The results were then compared in 

ANSYS CFD. 

The sensitivity analysis as well as 

all the other data in this paper is 

plotted along the length of four lines 

on the surface of the structure, these 

lines run parallel to the axis of the 

surge line and are labeled by the angle 

at which they end near the pressurizer 

end of the surge line as seen in Figure 

4. These lines start from the part of 

sure line near the main pipe, the data 

is plotted along the length of the line 

and at the end point the face of pipe is 

considered for their names. 

| Fig. 3. 

Position of lines with respect to pipe. 


Meshing and Sensitivity 

Analysis 

The accuracy of the results in any 

discrete simulation depends significantly 

on the mesh. The solution 

space should be defined in such a 

manner that the simulation could be 

com pleted accurately and with low 

amount of utilized resources. For 

evaluating the temperature profile of 

the structure under discussion our 

region of interest was the connection 

connecting portion between the two 

the pipes. So a separate body was 

assigned and the mesh in that body 

was refined step by step until desired 

quality of results were achieved. The 

details of the mesh are provided in the 

table the mesh was made using ICEM, 

the connection of interest was sized 

using the body sizing function to 

achieve the max element size as 

shown in table. The fluid inside of 

the structure was meshed separately 

as it is required in conjugate heat 

transfer for the solid and fluid domains 

to be defined separately 

| Fig. 4. 

Sensitivity Analysis of mesh. 

| Fig. 5. 

Temperature of Surge line fluid and structure. 





| Fig. 6. 

Temperature Contours of structure and fluid. 

Results 

Considering the results of the sensitivity 

analysis of the mesh it can be 

seen that the results of Mesh id 1 and 

Mesh id 2 have converged, therefore 

mesh id 2 was used to perform further 

analysis in order to reduce the 

computational cost. 

T structure (l line , T fluid ) = p 00 + p 10 l line + p 01 T fluid + p 20 l line 

2 

+ p 11 l line T fluid + p 02 T fluid 

2 

Line 1 

p 00 = -276.7 (-652.7, 99.2) 

p 10 = -0.8854 (-1.385, -0.3856) 

p 01 = 7.531 (3.653, 11.41) 

p 20 = -9.293e-05 (-0.0001075, -7.835e-05) 

p 11 = 0.00458 (0.002426, 0.006734) 

p 02 = -0.02453 (-0.03519, -0.01387) 

Line 2 

p 00 = -1222 (-2119, -324.9) 

p 10 = -0.5261 (-0.7273, -0.325) 

p 01 = 14.59 (6.563, 22.61) 

p 20 = -3.921e-05 (-4.695e-05, -3.147e-05) 

p 11 = 0.002607 (0.001698, 0.003516) 

p 02 = -0.0366 (-0.05454, -0.01866) 

Line 3 

p 00 = 228.1 (191, 265.2) 

p 10 = 0.04682 (-0.03542, 0.1291) 

p 01 = -0.1856 (-0.3752, 0.004024) 

p 20 = -0.0001034 (-0.0001209, -8.582e-05) 

p 11 = 0.0005824 (0.0002221, 0.0009427) 

p 02 = -0.0005446 (-0.001159, 6.955e-05) 

Line 4 

p 00 = 204.4 (174.8, 234) 

p 10 = -0.1004 (-0.1327, -0.06796) 

p 01 = 0.657 (0.3585, 0.9555) 

p 20 = -6.994e-05 (-7.681e-05, -6.307e-05) 

p 11 = 0.001042 (0.0008915, 0.001193) 

p 02 = -0.003653 (-0.004466, -0.002839) 

| Tab. 3. 

Temperature relations. 

Goodness of fit: 

SSE: 2048 

R-square: 0.9404 

Adjusted R-square: 0.9352 

RMSE: 5.942 


SSE: 156.3 



RMSE: 1.642 


SSE: 3139 



RMSE: 7.357 


SSE: 342.5 



RMSE: 2.43 

As we are interested in the effects 

on structure under surge line operational 

conditions we first carried out 

the conjugate heat transfer analysis 

on the pipe, during this analysis the 

effects of both convection of liquid 

and the consequent heat conduction 

with the structure are considered and 

we are provided with a comprehensive 

temperature profile of the structure 

,which considering the thickness 

of pipe is necessary for an accurate 

analysis as the surface temperature 

of the fluid doesn’t provides the 

complete picture. 

These temperature are then 

exported to ANSYS mechanical where 

further analysis on the stress resulting 

from these conditions were calcu lated, 

further the deformations as a result of 

the stress were also com puted. 

Temperature of Structure 

The temperature of the surgeline 

surface is given in Figure 5. It can be 

observed from the graph that at the 

starting point of plot a severe case of 

thermal stratification is present, as the 

O° line experiences lower temperature 

while 180° line is sub jected to higher 

temperature. This is as expected 

and has been widely reported in the 

literature. As the section near the hot 

leg is the location where the mixing of 

fluids takes place. 

As we move upwards along the 

pipe away from the main pipe the 

thermal stratification reduces and at 

0.5 m distance all the temperature 

achieves the uniform temperature, 

which is the temperature of the fluid 

entering from the pressurizer. 

Temperature Relations 

The structure temperature is the temperature 

we are interested in however 

in the working conditions the temperature 

sensors are present in the fluid 

instead of the structure so it is useful to 

have a relation that gives an approximate 

temperature for the structure at a 

particular point if temperature of the 

fluid is available from the sensors. 

Using the simulated data a second 

order equation was developed that 

provides the temperature of the structure 

corresponding to the line length 

and the temperature of fluid at that 

point and are given as follows, the 

co-efficient provided are with 95 % 

confidence interval, Table 3. 

Equivalent stress 

To compute the stress in the structure 

the thermal temperature were loaded 

onto the structure, as we are only 

interested in the thermal stress the 




mechanical stresses due to fluid flow 

were ignored. The sections of surge line 

where it is connected with the pressurizer 

and main pipe are con sidered as 

fixed supports for this analysis. 

The results of stress can be divided 

into three regions broadly, the first 

section is the section that is near the 

main pipe, this section experiences 

very high stress as expected, we can 

also see that in Figure 7, where180° 

line experiences lower stress as compared 

to the other lines however after 

reaching a minimum value it starts to 

rise and then we see that all the lines 

having a same general trend with 180° 

line and 0° line experiencing more 

stress than 90° line and 270° line. 

In section 2 graph this can be 

observed even more clearly as we can 

see a clear division between the stresses 

experienced by one section of the pipe 

as compared to the other section. In 

section 3 we observe that the previous 

trend reaching the end at 14 m where 

the pipe experiences a sharp turn and a 

new trend of extremely high stress is 

observed due to the incoming stream of 

hot water from the pressurizer. 

Deformation 

In the total Figure 8 we can observe 

the deformation experienced by the 

structure under the above mention 

stresses, the deflection is mostly 

observed in the middle region of the 

pipe which is the unsupported region 

of the pipe, in our model this region 

was considered as unsupported but 

in an actual plants these regions 

movements are usually limited by 

supporting structures. 

Conclusions 

The thermal stresses in the surge lines 

due to thermal stratification is a 

widely observed phenomenon, in this 

paper a steady state analysis of the 

flow in surge line was conducted to 

analyze a long term outlook of surge 

line under continued stress, the results 

are concluded in the flowing points. 

1. The stresses in the surge line are 

present in the steady state especially 

in the section of the surge 

line near the main pipe, these 

stress exists due to the thermal 

stratification where the mixing of 

hot and cold water takes place. 

2. The equivalent stress show that as 

we move further away from the hot 

leg of the main pipe the stresses 

first decrease and then start to 

reach a very high value near the 

pressurizer opening, this is due to 

the extremely high temperature at 

the inlet of the surgeline. 

| Fig. 7. 

Equivalent stress in surge line. 

| Fig. 8. 

Deformation in surge line. 

3. The deformation resulting from 

these stresses effect mostly the 

middle of the surge line pipe as 

there is no support between the 

endpoints in a considerably large 

structure, for practical purposes 

support of some kind are recommended 

in between the pressurizer 

and the main pipe. 

4. An approximation of the outer surface 

structure temperature based 

on the temperature of fluid at the 

boundary was also calculated from 

the simulated results, this can be 

useful in practical implementation 

where fluid data from sensors is 

generally available. 

References 

[1] NEA, 2005. Thermal Cycling in LWR Components in OECD-NEA 

Member Countries, NEA/CSNI/R(2005)8, NEA CSNI, CSNI Integrity 

and Ageing Working Group. Organization for Economic 

Co-operation and Development. 

[2] Grebner, H. and Höfler, A. (1995). Investigation of stratification 

effects on the surge line of a pressurized water reactor. Computers 

& Structures, 56(2-3), pp.425-437. 

[3] Ensel, C., Colas, A. and Barthez, M. (1995). Stress analysis of a 

900 MW pressurizer surge line including stratification effects. 

Nuclear Engineering and Design, 153(2-3), pp.197-203. 

[4] Kumar, R., Jadhav, P., Gupta, S. and Gaikwad, A. (2014). Evalu a- 

tion of Thermal Stratification Induced Stress in Pipe and its Impact 

on Fatigue Design. Procedia Engineering, 86, pp.342-349. 

[5] Kang, D., Jhung, M. and Chang, S. (2011). Fluid-structure interac 

tion analysis for pressurizer surge line subjected to thermal 

stratification. Nuclear Engineering and Design, 241(1), 

pp.257-269. 

[6] Zhang, Y. and Lu, T. (2017). Unsteady-state thermal stress and 

thermal deformation analysis for a pressurizer surge line 

subjected to thermal stratification based on a coupled CFD-FEM 

method. Annals of Nuclear Energy, 108, pp.253-267. [7] Similar 

shaped models 

[7] Cai, B., Gu, H., Weng, Y., Qin, X., Wang, Y., Qiao, S. and Wang, 

H. (2017). Numerical investigation on the thermal stratification 

in a pressurizer surge line. Annals of Nuclear Energy, 101, 

pp.293-300. 

[8] Baik, S. (n.d.). [online] Inis.iaea.org. Available at: 

https://inis.iaea.org/collection/NCLCollectionStore/_ 

Public/32/068/32068795.pdf [Accessed 19 Dec. 2018]. 

[9] Schuler, X., & Herter, K.H. (2004). Thermal fatigue due to stratification 

and thermal shock loading of piping. 30 MPA-Seminar 

'Safety and reliability in energy technology' in conjunction with 

the 9th German-Japanese seminar Vol 1 (Papers 1-26), (p. 464). 

[10] NEA, 2005. Thermal Cycling in LWR Components in OECD-NEA 

Member Countries,NEA/CSNI/R(2005)8, NEA CSNI, CSNI 

Integrity and Ageing Working Group. Organization for Economic 

Co-operation and Development. 

[11] Grebner, H. and Höfler, A. (1995). Investigation of stratification 

effects on the surge line of a pressurized water reactor. 

Computers & Structures, 56(2-3), pp.425-437. 

[12] Ensel, C., Colas, A. and Barthez, M. (1995). Stress analysis of a 

900 MW pressurizer surge line including stratification effects. 

Nuclear Engineering and Design, 153(2-3), pp.197-203. 

[13] Sang-Nyung Kim, Seon-Hong Hwang,Ki-Hoon Yoon. Experiments 

on the Thermal Stratification in the Branch of NPP Journal 

of Mechanical Science and Technology ( KSME Int. J.), 2005, 

19(5):1206-1215 

[14] Sang-Nyung Kim, Cheol-Hong Kim, Bum-Su Youn, Hag-Ki Yum, 

Experiments on Thermal Stratification in Inlet Nozzle of Steam 

Generator, Journal of Mechanical Science and Technology, 

2007(21):654-663 

[15] T.H.Liu, E.L.Cranford. An Investigation of Thermal Stress Ranges 

Under stratification Loadings [J]. Transactions of the ASME 326/ 

Vol. 113, 1991 

Authors 

Muhammad Abdus Samad 

Xiang bin li 

School of Nuclear Science 

and Engineering 

North China Electric Power 

University 

Beijing, China 

Hong lei Ai 

Nuclear Power Institute of China 

Sichuan, China 





276 

ENVIRONMENT AND SAFETY 

Physical and Chemical Effects of 

Containment Debris on the Emergency 

Coolant Recirculation 

Jisu Kim and Jong Woon Park 

Physical and chemical effects of containment debris on the performance of emergency coolant recirculation are 

investigated to get insight on the cost-effective plant modifications to resolve USNRC’s Generic Safety Issue-191. The 

effects of debris sources on the sump screen performance are evaluated through the head loss calculation using NUREG/ 

CR-6224 correlation. The amount of three predominant types of precipitates, i.e., sodium aluminum silicate 

( NaAlSi3O8), aluminum oxyhydroxide (AlOOH), calcium phosphate (Ca3(PO4)2) after 30 days of ECCS mission time 

are evaluated under various environmental conditions using WCAP-16530-NP chemical models. The debris interceptor 

is considered as a viable design option to reduce particulate debris such as unqualified coatings. The key parameters of 

each effect are deduced and recommendations for reducing their adverse effects are made through the present analysis: 

(a) The amount of unqualified coating debris is a major source of particulate debris and has a great adverse effect on the 

sump screen head loss by reducing porosity in the fibrous insulation, (b) The Cal-Sil insulation reacts with TSP buffer 

and significantly increases the generation of a gum-like chemical precipitant, (c) Spray time increases the chemical 

byproducts but the effect is smaller than that of buffer agent type and unqualified coating, (d) The debris interceptor, 

when verified, may play a vital role reducing head loss generated by coatings and fibrous debris mix. 


A primary safety issue regarding 

long-term recirculation core cooling 

following a LOCA (Loss of Coolant 

Accident) is that LOCA-generated 

debris may be transported to the 

recirculation sump screen, resulting in 

adverse blockage on the sump screen 

and deterioration of available NPSH 

(Net Positive Suction Head) of ECCS 

(Emergency Core Cooling System). 

USNRC identified this as Generic 

Safety Issue (GSI) 191 [1] and issued 

the Generic Letter 04-02 [2] to resolve 

the issue. The GL required that all 

PWR owners perform an engineering 

assessment of their containment 

recirculation sumps to ensure they will 

not suffer from excessive blockage. The 

guidance report (GR) [3] for PWR 

sump performance evaluation has 

been developed by NEI (Nuclear 

Energy Institute) and approved by the 

USNRC [4]. 

The objective of the assessment is to 

derive required plant modifications 

including new insulation, sump 

screen, etc. of a Korean nuclear power 

plant for 10-year life extension. To 

derive the cost-effective modifications, 

the effects of physical and chemical 

conditions on the performance of 

the ECCS recirculation sump with 

respect to head loss are parametrically 

investigated. The physical and 

chemical conditions are debris 

source, containment environments 

and debris interceptor as a candidate 

design option. 

2 Analysis methods 

2.1 NUREG/CR-6224 head loss 

correlation 

The effects of debris sources on the 

sump screen are evaluated through 

head loss calculations using NUREG/ 

CR-6224 correlation [5]. This is 

applicable for laminar, transient, and 

turbulent flow regimes through mixed 

debris beds (i.e., debris beds composed 

of fibrous and particulate debris). This 

correlation is approved by USNRC for 

the determination of head loss [4] and 

is given by: 

 

(1) 

The fluid velocity (U), is given by 

simply in terms of the volumetric flow 

rate (Q) and the effective screen 

area (A) as: 

(2) 

The mixed debris bed solidity a m is 

given by: 

(3) 

For debris deposition on a flat surface 

of a constant size, the compression 

rate (c) relates the actual debris bed 

thickness (DL m ) and the theoretical 

fibrous debris bed thickness (DL o ) via 

the relation: 

(4) 

Compression of the fibrous bed due 

to the pressure gradient across the 

bed is also accounted, which must be 

satisfied in parallel to the previous 

head loss equation, Eq.(1), is given by 

(valid for ratios of DH/DL o > 0.5 

ft-water/inch-insulation): 

 

(5) 

where “K” is a constant that depends 

on the insulation type. The value of K 

is 1.0 for NUKON fiber. Test data or 

a similitude analysis are required to 

determine “K” for fibrous materials 

that are dissimilar to NUKON insulations. 

Each constituent of debris has a 

surface-to-volume ratio associated 

with it based on the characteristic 

shape of that debris type. For typical 

debris type, we have [3]: 

Cylindrically shaped debris: 

S v = 5/diam 

Spherically shaped debris: 

S v = 6/daim 

Flakes (flat plates): 

S v = 2/thick 

where “diam” is the diameter in feet of 

the fiber or spherical particle, and 

“thick” is the thickness in feet of the 

flake/chip. 


Physical and Chemical Effects of Containment Debris on the Emergency Coolant Recirculation ı Jisu Kim and Jong Woon Park


The average surface-to-volume ratio 

of various types of debris constituents 

is calculated as: 

(6) 

where v is the microscopic volume of 

the constituent and the subscript “n” 

refers to the n-th constituent. 

2.2 WCAP-16530 post-accident 

chemical effects evaluation 

method 

The materials inside containment may 

dissolve or corrode when exposed 

to the reactor coolant and spray 

solutions. This would produce oxide 

particulate corrosion products and a 

potential for formation of precipitates 

due to chemical reactions with other 

dissolved materials. These chemical 

products may become another source 

of debris loading to be considered in 

sump screen performance and downstream 

effects. 

Recently Westinghouse Owners 

Group (WOG) proposed a four step 

process for evaluating the postaccident 

chemical effects in containment 

sump fluids to support 

GSI-191 [6]. As shown in Figure 1, 

using ICET test results and plant data, 

the chemistry bench tests are performed 

and a chemical model is 

developed to identify the type and 

amount of chemical products that are 

produced. This chemical product 

information generated from the bench 

testing and the chemical model is 

used as an input to performance 

testing to be conducted by licenses 

and vendors of replacement sump 

screens. 

Through bench test, three types of 

predominant chemical precipitates 

are identified for the plant using 

NaOH or TSP (Tri-Sodium Phosphate) 

as a buffer agent: 

p Sodium aluminum silicate 

(NaAlSi 3 O 8 ) 

p Aluminum oxyhydroxide (AlOOH) 

p Calcium phosphate (Ca 3 (PO 4 ) 2 ) 

(if TSP is used) 

Each quantity of precipitate generated 

can be calculated as followings: 

where for each chemical species, 

concentration data generated during 

the single-effect bench testing at 

specific chemistry conditions is used 

in a regression analysis to develop 

release equations as a function of 

temperature, pH, and the concentration 

of that species. Equations are 

developed for each predominant 

source material for each chemical 

species (Ca, Al, and Si). The detailed 

information about the equations for 

the material release rate is described 

in the reference 6. 

3 Results and discussion 

3.1 Effects of debris sources 

The sump screen head loss calculations 

with various debris loadings 

on the sump screen are performed 

using USNRC’s NUREG/CR-6224 

correlation [5]. The screen area is 

assumed as 1,000 ft 2 with maximum 

ECCS flow rate of 7,000 gpm. The 

sump pool temperature and pressure 

are assumed as 212 °F and 14.7 psi, 

respectively. The debris source and 

their characteristics are summarized 

in Table 1. It is assumed that the 

particulate debris mixture consists of 

85 % of coatings, 10 % CalSil and 5 % 

of latent dust/dirt debris by mass. 

The head loss by RMI (Reflective 

Metal Insulation) debris bed is not 

Debris Type 

As-fabricated Density 

[lbm/ft 3 ] 

| Tab. 1. 

Debris source and characteristics [3,4]. 

| Fig. 1. 

WCAP-16350 Post-Accident Chemical Effects Evaluation Methodology. 

con sidered because its effect on the 

head loss is negligible 

The head loss with various loadings 

of fiber and particulate mixture 

debris is shown in Figure 2. The head 

losses by fiber only debris beds are 

significantly lower than mixed debris 

beds of fiber and particulate. The head 

loss appreciably increases with the 

amount of particulate debris. Head 

loss drastically increases with the 

decrease of the amount of fiber debris 

because the mass ratio of particulate 

to fiber increases. The debris packing 

Particle Density 

[lbm/ft 3 ] 

Sv 

[ft -1 ] 

Fiber 2.4 175 171,700 

Coatings - 94 183,000 

CalSil - 115 600,000 

Dirt/Dust - 169 106,000 

ENVIRONMENT AND SAFETY 277 

[NaAlSi 3 O 8 ] 

= 3.11 [Si], if [Si] < 3.12 [Al] 

= 9.72 [Al], if [Si] > 3.12 [Al] 

(7) 

[AlOOH] = 2.22 {[Al] - 0.32 [Si]} 

(8) 

[Ca 3 (PO 4 ) 2 ] 

= 2.58 [Ca] (if TSP is used) 

(9) 

| Fig. 2. 

Head loss with various mixed debris loading on the 1,000 ft 2 sump screen (7000 gpm ECCS flow). 





limit is shown in Figure 2, which is 

closely related to the maximum 

solidity limit. Above this limit, the 

particulate is the predominant 

ingredient and the fiber is embedded 

in the matrix. Such a condition of the 

debris bed is physically unacceptable. 

This situation can arise in the plant 

with a large particulate debris source. 

Therefore the mass ratio of particulate 

to fibrous debris is an important 

parameter in the evaluation of the 

head loss. Major source of particulate 

debris is the unqualified coatings in 

the containment because the generation 

and transport of unqualified 

coatings are assumed to be 100 % in 

the GR [3]. 

The 4-in and 6-in initial fibrous 

debris loading lines are shown in 

Figure 2. USNRC recommended that 

NUREG/CR-6224 correlations be 

used within the range of 1/8 to 

4 inches initial fibrous debris loading 

(for NUKON) because it is not fully 

validated in the range exceeding 

4 inches. In that case, the screen 

size should be increased or NUKON 

insulation should be replaced by an 

alternate insulation (i.e., RMI). As 

mentioned above, the head loss is 

closely related to the particulate- tofibrous 

debris mass ratio. The screen 

size should be determined carefully 

Amount of Cal-Sil insulation [ft 3 ] 0, 50, 100 

Spray Termination Time [sec] Short: 95,000 

Long; 1,000,000 

Buffer Agent 

| Tab. 2. 

Environmental conditions in the present study. 

considering the cost effects between 

the reductions of the amounts of 

fibrous and particulate debris. 

3.2 Effects of containment 

environmental conditions 

The prediction model for head loss 

by chemical products is currently 

not available and its effect on the 

head loss is evaluated only by 

the screen vendor’s performance 

testing. In the present analysis, the 

amounts of the potential chemical 

precipitates in the various containment 

environments are evaluated 

using WCAP-16530-NP methodology 

[6]. The amount of three predo minant 

types of pre cipitates, i.e., sodium 

aluminum silicate (NaAlSi 3 O 8 ), aluminum 

oxyhydroxide (AlOOH), calcium 

phosphate (Ca 3 (PO 4 ) 2 ) after 30 

days of ECCS mission time are 

evaluated under various environmental 

conditions as shown in Table 2. 

The typical envi ronmental conditions 

of a Korean Westinghouse two loop 

NaOH, TSP 

nuclear power plant are used as the 

following input data in the present 

analysis: 

(a) pH and temperature profiles of 

sump and containment during 

30 days, 

(b) The maximum pool volume during 

the recirculation phase after 

LBLOCA, 

(c) Amount of metallic aluminum 

(submerged and unsubmerged), 

(d) Amount of E-glass within ZOI 

(such as NUKON) 

(e) Concrete surface area within ZOI 

(submerged and unsubmerged). 

The presence of Cal-Sil insulation 

increases the releases of calcium and 

silicate as shown in Figure 3. For 

plants with sodium hydroxide (NaOH) 

buffer, sodium aluminum silicate is 

the principal precipitant since the 

insulation mix is a main source of 

silicon. 

Both of the presence of Cal-Sil 

and long spray time could drastically 

increase the amount of sodium 

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| Fig. 3. 

Comparison of the chemical precipitates with various plant environments. 

aluminum silicate precipitates because 

long spray time increases 

dissolution of aluminum as well as 

silicon. The precipitation of aluminum 

oxy hydroxide is negligible in the 

present analysis because the concentration 

of aluminum is not sufficiently 

lager than that of silicon. The 

quantity of sodium aluminum silicate 

is limited by the amount of silicon 

available in solution, if the silicon 

concentration is less than 3.12 times 

the aluminum concentration as shown 

in Eq. (7). Any remaining aluminum 

in solution will precipitate as aluminum 

oxyhyroxide. If the concentration 

of silicon is greater than 3.12 times the 

aluminum concentration, then the 

quantity of sodium aluminum silicate 

generated is limited by the concentration 

of aluminum available. 

For plants using trisodium 

phosphate (TSP), calcium phosphate 

is generated in addition to sodium 

aluminum silicate and aluminum 

oxyhydroxide. When CalSil and TSP 

co-exist, significant amount of calcium 

phosphate is generated with increasing 

amount of Cal-Sil as shown in 

Figure 3. It has been known that 

calcium phosphate has a significant 

impact on the head loss because its 

characteristics are like a gum. 

Therefore, above-mentioned three 

parameters, i.e., the amount of Cal-Sil, 

long spray termination time, use of 

TSP as a buffer agent has a most 

negative effect on the recirculation 

sump performance. Through the 

present study, the following recommendations 

are made: 

(a) Reduce the amount of Cal-Sil 

insulation 

(b) Replace TSP with alternate buffer 

agent 

(c) Reduce spray time by operation 

3.3 Debris interceptor 

There may be various design options 

to reduce the head loss across the 

sump screen such as an active strainer 

and a specially designed screen 

surface to prevent the thin bed effects. 

The adoption of a debris interceptor is 

another viable option for reducing 

particulate debris such as unqualified 

coatings. The detailed effects on the 

debris transport are dependent on the 

specific debris interceptor design. For 

example, if the debris interceptor can 

reduce the particulate debris from 

3,000 lbm to 1,000 lbm, the head loss 

can be reduced form 3.98 ft to 0.87 ft 

for 400 ft 3 fiber debris loading as 

shown in Figure 1. 

4 Conclusions 

A parametric study is performed on 

the effects of debris source, containment 

environments and debris interceptor 

on the overall performance of 

ECCS recirculation sump. The key 

parameters of each effect are deduced 

and the recommendations for reducing 

their adverse effects are made 

through the present ana lysis. Following 

conclusions can be made: 

(a) The amount of unqualified coating 

debris has a great adverse effect on 

the screen head loss by reducing 

porosity in the fibrous insulation, 

(b) Cal-Sil insulation reacts with TSP 

buffer and significantly increases 

the generation of a gum-like 

chemical precipitate, 

(c) Spray time increases the chemical 

by-products but the effect is 

smaller than that of buffer agent 

type and unqualified coating. 

(d) The debris interceptor, when 

verified, may play a vital role reducing 

head loss generated by 

coatings and fibrous debris mix. 

The cost of reducing debris sources, 

removal of Cal-Sil insulation and 

installation of debris interceptor 

should be compared with the benefit 

of reducing number of suction 

strainers to select design change 

options for a particular plant. For this, 

the present parametric analysis 

method are being used for Korean 

nuclear power plants. 

Acknowledgments 

This work was supported by a grant 

from the nuclear safety research 

program of the Korea Foundation of 

Nuclear Safety with funding by the 

Korean Government’s Nuclear Safety 

and Security Commission (Grant 

Code: 1307008-0719-CG100). 

Nomenclature 

A effective screen area [ft 2 ] 

K 

constant depending insulation type 

S v surface-to-volume ratio of the debris [ft 2 /ft 3 ] 

U 

Q 

c 

v 

fluid approach velocity [ft/sec] 

volumetric flow rate [ft 3 /sec] 

compression rate 

microscopic volume of the debris constituent 

∆H head loss [ft-water] 

∆L m actual mixed debris bed thickness [in] 

∆L o theoretical fibrous debris bed thickness 

L conversion factor 

L = 1 for SI units, and 

L = 4.1528E-05 (ft-water/inch)/(lbm/ft 2 /sec 2 ) 

for English units. 

a o solidity of the original fiber blanket (i.e., the “as-fabricated” solidity) 

a m mixed debris bed solidity 

h 

μ 

m p /m f , the particulate-to-mass ratio in the debris bed 

(i.e., total particulate mass/total fibrous mass) 

dynamic viscosity of water [lbm/ft/sec] 

r density of water [lbm/ft 3 ] 

r ƒ fiber density [lbm/ft 3 ] 

r p average particulate material density [lbm/ft 3 ] 

References 

[1] NUREG-0933, A Prioritization of Generic Safety Issues, 

Supplements 28, USNRC, Aug. 2004. 

[2] USNRC Generic Letter 2004-02, Potential Impact of Debris 

Blockage on Emergency Recirculation during DBA at PWR, 

USNRC, Sep. 13, 2004. 

[3] NEI 04-07, Pressurized Water Reactor Sump Performance 

Evaluation Methodology, Rev. 1, Nov. 2004. 

[4] Safety Evaluation by the USNRC related to NRC NRC Generic 

Letter 2004-2 NEI Guidance Report (NEI-04-07) Pressurized 

Water Reactor Sump Performance Evaluation Methodology, 

Dec. 2004. 

[5] NUREG/CR-6224, Parametric Study of the Potential for BWR 

ECCS Strainer Blockage due to LOCA Generated Debris, 

Sep. 1995. 

[6] WCAP-16530-NP, Evaluation of Post-Accident Chemical 

Effects in Containment Sump Fluids to Support GSI-191, 

Westinghouse, Feb. 2006. 

Authors 

Jisu Kim 

Jong Woon Park 

Dongguk University, 

123 Dongdae-ro, Gyeongju, 

Gyeongbuk 38066, South Korea 

kimjs@dongguk.ac.kr 






Experimental and Computational 

Analysis of a Passive Containment 

Cooling System with Closed-loop 

Heat Pipe Technology 

Lu Changdong, Ji Wenying, Yang Jiang, Cai Wei, Wang Ting, Cheng Cheng and Xiao Hon 

A conceptual design of Passive Containment Cooling System with Closed-Loop Heat Pipe Technology (PCSHP) is 

studied using both experimental and computational methods. By studying on the thermal-hydraulic parameters in 

system running, such as temperature, pressure and flow rate, the paper mainly focuses on the start-up characteristics, 

the steady-state operating characteristics, the heat transfer capacity and the natural circulation capacity of the system. 

Hence, the principle experiment and GOTHIC simulation are carried out under start-up conditions, steady-state 

conditions and decay heat simulation conditions. The applicability and conservatism of the GOTHIC model is evaluated 

by comparing the simulating results with the experimental results. The rationality of the system design is validated by 

both the principle experiment and GOTHIC simulation. It is preliminarily judged that the heat pipe technology is 

feasible to apply to the Passive Containment Cooling System (PCCS) of nuclear power plant. 


Passive safety systems are adopted in 

the design of the third generation 

nuclear power plants represented by 

AP1000. Passive safety systems can 

enhance safety, reliability and economy 

of nuclear power plant by using 

natural driving forces such as gravity, 

natural circulation and natural convection 

[IAEA, 2009]. 

For instance, Passive Containment 

Cooling System (PCCS) is used to cool 

down and depressurize the containment 

in case of an accident and thus 

ensures the containment integrity. 

The passive containment cooling 

system of AP1000 uses the containment 

vessel as the heat transfer 

surface. The heat is released to the 

interior of the containment vessel by 

condensation of vapor, and then 

removed by means of an evaporating 

| Fig. 1. 

A conceptual design of Passive Containment Cooling System 

with Closed-Loop Heat Pipe Technology (PCSHP). 

water film combined with a natural 

circulation of air outside the containment 

vessel [Westinghouse, 2010; 

Li et al., 2017]. 

Unlike AP1000, a conceptual 

design of Passive Containment Cooling 

System with Closed-Loop Heat 

Pipe Technology (PCSHP) is shown in 

Figure 1. Loop heat pipe technology 

has been widely employed in the 

thermal management of spacecraft 

and electronic component. Although 

it has not been practically applied 

in nuclear power plant, its future 

prospect is remarkable. Loop heat 

pipes, as highly-effective passive heat 

transfer devices which transfer heat 

by internal phase change, have many 

advantages such as good heat transfer 

capacity, long transmission distance 

and flexible application [Chenlong et 

al., 2013; Jean et al., 2005]. Comparing 

to the passive containment 

cooling system of AP1000, PCSHP has 

simpler structure and better heat 

transfer efficiency. Besides, it is 

applicable to most existing nuclear 

power plants without major changes 

in the containment structure. 

The heat in the containment is 

firstly absorbed by the heat exchanger 

inside the containment (evaporator). 

The heated coolant boils and evaporates 

in the evaporator. Driven by the 

density difference, the vapor goes up 

along the loop to the heat exchanger 

in the cooling water storage tank 

outside the containment (condenser) 

where it is condensed and returns to 

the evaporator because of gravity. At 

this point, a closed-loop circulation is 

completed. The cooling water storage 

tank is constantly heated till its 

ebullition. Steam in the tank is released 

to atmosphere, the final heat 

sink, and thus derives heat from the 

containment. The heat exchanger inside 

the containment is in a half-full 

state and should maintain certain vacuum 

degree. Apart from the activation 

and isolation valves, the whole system 

does not comprise other active components 

or any components requiring 

alternating current power support. 

In order to validate the rationality 

of the design of PCSHP and to provide 

essential inputs for the design improvement 

and safety analysis of the 

system, the principle experiment was 

carried out. The principle experiment 

focuses on the start-up characteristics, 

steady-state operating characteristics, 

heat transfer capacity and natural 

circulation capacity of the system by 

studying on the thermal-hydraulic 

parameters in system running, such as 

temperature, pressure and flow rate. 

Based on conservative assumptions, 

we developed a GOTHIC model 

of PCSHP to study the thermalhydraulic 

characteristics of the 

system. The applicability and conservatism 

of this model is evaluated 

by comparing the simulating results 

with the experimental results. The 

GOTHIC model can be used to 

simulate the operating characteristics 

of PCSHP, optimize system design 

and provide a foundation model for 

accident analysis and containment 

pressure and temperature response 

analysis. 

2 Principle experiment 

The principle experiment is of significance 

to study the key technology of 


Experimental and Computational Analysis of a Passive Containment Cooling System with Closed-loop Heat Pipe Technology ı Lu Changdong, Ji Wenying, Yang Jiang, Cai Wei, Wang Ting, Cheng Cheng and Xiao Hon


PCSHP. By the principle experiment, 

we study the system characteristics, 

validate the rationality of the system 

design and lay a foundation for 

subsequent design improvements and 

safety analysis. 

The scheme of the principle expe riment 

is shown in Figure 2. Main 

parameters are listed in Table 1. Main 

experimental equipments include a 

heater simulator, a condensing tank, 

two heat pipe heat exchangers, a 

containment simulator and a steel 

platform which support experimental 

bench. The containment simulator is 

used to simulate the containment, the 

heater simulator is used to simulate 

heat source inside the containment, 

the condensing tank is used to simulate 

the cooling water storage tank 

outside containment and the two heat 

pipe heat exchangers are used to 

model the evaporator and condenser 

of PCSHP. 

The objectives of PCSHP principle 

experiment are: 

a) To study the influence of the initial 

containment pressure on the startup 

characteristics of PCSHP and 

obtain operating parameters such 

as the pressure, temperature and 

start-up flow rate of the system. 

b) To study the influence of the containment 

pressure and the initial 

vacuum degree of the loop on the 

steady-state operating characteristics 

such as the natural circulation 

flow rate and the overall heat 

transfer capacity. 

c) To verify whether the system can 

reach the steady state of natural 

circulation after the core decay 

heat decreases by the step change 

of heating power which simulates 

the decline of the core decay heat. 

To achieve the listed objectives, the 

experiment is divided into three parts: 

start-up conditions, steady-state conditions 

and decay heat simulation 

conditions. 

Before the experiment begins, the 

experimental bench is at environmental 

condition. The system filling rate is 

adjusted to 50 % and the containment 

simulator and the condensing tank are 

adjusted to the specified water level. 

Initial conditions of the start-up 

conditions are shown in Table 2. At 

the beginning, the initial pressure 

(vacuum degree) of the loop is set to 

0.045 MPa by the vacuum pump. The 

heater simulator is turned on and 

maintains containment pressure at 

0.35 MPa. Then the heating power is 

adjusted to 109 kW and the isolation 

valves are turned on simultaneously. 

Main operating parameters such as 

| Fig. 2. 

The scheme of Passive Containment Cooling System with Closed-Loop Heat Pipe Technology (PCSHP)principle experiment. 

Equipments 

Evaporator 

Condenser 

Height difference between 

evaporator and condenser 

Ascending leg 

Descending leg 

System filling rate 

temperature, pressure, flow rate are 

collected by computer during system 

start-up. By repeating the experiment 

at initial containment pressure of 

0.40 MPa, 0.45 MPa and 0.52 MPa, 

the start-up characteristics of the 

system are studied. 

As for the study of the steadystate 

characteristics of PCSHP with 

different containment pressure and 

Parameters 

Outer diameter of single heat exchange tube: 57 mm 

Wall thickness of single heat exchange tube: 2.5 mm 

Number of heat exchange tubes: 37 

Length of single heat exchange tubes: 1.2 m 

Outer diameter of single heat exchange tube: 57 mm 

Wall thickness of single heat exchange tube: 2.5 mm 

Number of heat exchange tubes: 37 

Length of single heat exchange tubes: 1.2 m 

5.5 m 

Inner diameter of tube: 150 mm 

Inner diameter of tube: 20 mm 

50% (Volume faction of the heat exchange tubes 

in the Evaporator) 

Initial vacuum degree Start-up condition: 0.045 MPa 1 

Steady condition: various value 

Water level of condensing tank 2.6 m (total height: 3.0 m) 

Diameter of condensing tank 1.5m 

Inner diameter of containment simulator 

Height of containment simulator 

Water level of containment simulator 

2.0 m (wall thickness: 12 mm) 

3.0 m 

1.3 m 

| Tab. 1. 

System parameters of Passive Containment Cooling System with Closed-Loop Heat Pipe Technology (PCSHP) principle experiment. 

Initial containment 

pressure 

| Tab. 2. 

Start-up conditions. 

Heating power 

in the containment 

Initial vacuum degree 

of the loop 

different initial vacuum degree, the 

initial conditions are listed in Table 3. 

The initial temperature of the condensing 

tank is saturated temperature 

(about 100 °C). In order to study the 

influence of the containment pressure 

and the initial vacuum degree on the 

natural circulation flow rate of the 

system, the initial vacuum degree of 

the loop is set and the containment 

Initial temperature 

of consensing tank 

0.35 MPa 109 kW 0.045 MPa Ambient temperature 




1) Unless otherwise 

specified, pressure in 

this paper refers to 

absolute pressure. 






Initial 

vacuum degree 

of the loop 

Initial vacuum degree 

of the loop 

Containment 

pressure 

pressure is maintained constant. 

When the temperature and flow rate 

of the loop reach a steady state, the 

natural circulation flow rate of the 

system is recorded. 

Initial conditions of the decay heat 

simulation conditions are shown in 

Table 4. The decay heat simulation 

conditions study on the steady-state 

characteristics of PCSHP under different 

input power. The decline of the 

core decay heat is simulated by the 

step change of heating power to verify 

whether the system can reach the 

steady state of natural circulation 

after the core decay heat decreases. In 

order to study the influence of heating 

power and initial vacuum degree on 

the steady-state characteristics, the 

heating power of the heater simulator 

and the initial vacuum degree of 

the loop are set as Table 4. Before 

the system starts, the condensing tank 

is at saturated temperature (about 

100 °C), the containment is at environmental 

condition. Once the temperature 

and flow rate of the loop 

reach a steady state, the natural 

circulation flow rate of the system is 

recorded. 

Initial temperature 

of consensing tank 

0.021 MPa 0.30 MPa ~ 0.52MPa Saturated temperature 




| Tab. 3. 

Steady-state conditions. 

Initial 

containment 

pressure 

| Tab. 4. 

Decay heat simulation conditions. 

Initial 

containment 

temperature 

Initial 

temperature of the 

condensing tank 

Heating 

power 

0.021 MPa 1 atm Ambient temperature Saturated temperature 10 kW ~ 80 kW 




well as the heat transfer inside and 

between solids and fluids [EPRI, 

2014]. 

The GOTHIC version 8.1 code is 

used to model the principle experiment. 

The model diagram is shown in 

Figure 3. Wherein, the boundary 

condition 1P refers to the environment, 

the control volume 1 is the containment, 

the control volume 2 is the 

heat exchanger in the containment 

(evaporator), and the control volume 

5 is the heat exchanger in the cooling 

water storage tank outside the containment 

(condenser), and the control 

volume 8 is the cooling water storage 

tank outside the containment, control 

volume 3, control volume 4, control 

volume 6 and control volume 7 

indicate connected pipes [Hui-Un et 

al., 2013; Philipp et al., 2011]. 

In order to accurately simulate the 

thermal stratification effect and 

natural circulation in the containment, 

the evaporator and the condenser 

(control volume 1, control 

volume 2 and control volume 5) are 

subdivided. That is to say, a large 

control volume is divided into many 

small subdivided volumes. Conversely, 

other control volumes are 

simu lated using lumped parameters. 

The flow between the above control 

volumes is simulated by the flow path, 

wherein the cooling water storage 

tank outside containment is connected 

to the environment with flow 

path 7, which simulates the chimney 

structure of the storage tank so that 

the heated steam in the storage tank 

can be smoothly discharged. The 

heating power in the containment is 

simulated using a heater component. 

Valves are placed upstream and downstream 

of the evaporator to model 

system start-up and shutdown. 

The heat transfer between the condenser 

and the cooling water storage 

tank outside containment is simulated 

by the thermal conductor 1. The heat 

transfer between the condenser and 

the thermal conductor 1 uses a diffusion 

layer model (DLM) to simulate 

steam condensation. The temperature 

rise of the cooling water storage tank 

3 GOTHIC modeling of the 

principle experiment 

GOTHIC (Generation of Thermal Hydraulic 

Information for Containment) 

is a general purpose code for thermalhydraulic 

calculation, mainly used for 

containment design, license application, 

safety analysis and operational 

analysis of nuclear power plants. 

GOTHIC is capable of modeling multiphase 

fluid flow involving steam, gases, 

pools, droplets, bubbles and ice 

and the interaction between phases, 

including sublimation, evaporation, 

condensation, boiling and flashing as 

| Fig. 3. 

GOTHIC model of Passive Containment Cooling System with Closed-Loop Heat Pipe Technology (PCSHP). 




(a) Experimental results 

| Fig. 4. 

Containment pressure in start-up conditions. 

(b) GOTHIC simulation results 


(a) Experiment results 

| Fig. 5. 

Flow rate of the loop in start-up conditions. 


outside containment is simulated by 

the FILM heat transfer model that 

GOTHIC built in. At this point, the 

GOTHIC code automatically selects the 

single-phase liquid heat transfer relationship 

built in the FILM model. The 

heat transfer between the evaporator 

and the containment is simu lated by 

the thermal conductor 2, wherein the 

condensation of the containment simulated 

with the Uchida relation. And 

the boiling heat transfer of the evaporator 

is simulated with the FILM heat 

transfer model that GOTHIC built in. 

There may be multiple heat transfer 

modes such as single-phase liquid natural 

convection heat transfer, subcooled 

nucleate boiling heat transfer, 

saturated nucleate boiling heat transfer, 

film boiling heat transfer, and 

single- phase steam natural convection 

heat transfer in the evaporator. The 

GOTHIC code is able to recognize the 

above mentioned heat transfer modes 

and automatically switches relations. 

The heat transfer inside both 2 thermal 

conductors is heat conduction. Since 

the pipe materials are all made of stainless 

steel, the properties of the stainless 

steel are used such as density, thermal 

conductivity and specific heat. 

In the model, the ambient temperature 

is conservatively set at 30 °C, 

and atmospheric pressure is taken as 

101 kPa. Conservatively, the heat 

absorption of the equipment, walls, 

floors or ceilings isn’t considered, 

neither does the heat loss of the 

experimental system to the environment. 

Other initial conditions and 

parameters are exactly the same as 

those in the experiments. 

4 Simulation results 

and analysis 

The principle experiment is modeled 

using the GOTHIC code, and the startup 

conditions, steady-state conditions 

and decay thermal simulation conditions 

are simulated and analyzed. 

4.1 Start-up conditions 

In the start-up conditions, under 

different initial containment pressures, 

the curves of containment 

pressure over time are shown in 

Figure 4, where (a) shows the experimental 

results and (b) shows GOTHIC 

simulation results (P1 indicates the 

initial containment pressure). When 

the heat transfer capacity of PCSHP is 

greater than the heating power, the 

containment pressure will gradually 

decrease. Conversely, when the heat 

transfer capacity of PCSHP is less than 

the heating power, the containment 

pressure will gradually increase. 

According to the experimental results, 

when the initial containment pressure 

is 0.35 MPa and 0.40 MPa, the containment 

pressure begins to rise after 

a brief decline. When the initial containment 

pressure is 0.45 MPa and 

0.52 MPa, the containment pressure 

will gradually decrease after the 

brief decline. However the GOTHIC 





simulation results shows that the containment 

pressure at all conditions 

begin to rise after a brief decline. 

Never theless, the trend of the containment 

pressure is about the same in the 

simulation results and the experimental 

result, but the simulation 

results underestimate the heat transfer 

capacity of PCSHP compares to 

experimental results. 

In the start-up conditions, under 

different initial containment pressure, 

the curves of the flow rate of the loop 

over time are shown in Figure 5. After 

the loop heat pipe is activated, the 

natural circulation flow rate reaches a 

large value instantly and decreaces 

rapidly at a certain moment, then 

the flow rate oscillates and gradually 

stabilizes. Under same boundary conditions, 

the flow rate in the loop tends 

to increase with the rise of initial containment 

pressure. The comparison of 

the simulation results and the experimental 

results shows that the trend of 

the flow rate of the loop is about the 

same. However, the flow rate obtained 

by the simulation results is relatively 

small compared to the experimental 

results. Generally, the heat transfer 

capacity of the system is positive correlated 

with the circu lating flow rate, 

which also indicates that the simulation 

results under estimate the heat 

transfer capacity of PCSHP. 

In the start-up condition, under 

different initial containment pressures, 

the curves of loop pressure over 

time are shown in Figure 6. After the 

system is started, the loop pressure 

drops rapidly, and it rises at a certain 

moment and gradually stabilizes, 

maintaining a tendency to rise slowly. 

Under the same boundary condition, 

the flow rate in the loop tends to 

increase with the rise of initial containment 

pressure. The comparison 

between the simulation results and 

the experimental results shows that 

the trend of the loop pressure is about 

the same, but the loop pressure 

obtained by simulation results are 

relatively small compared to the 

experimental results. 

In the start-up condition, with 

initial containment pressure at 

0.35 MPa, the curves of the fluid 

temperature at various positions 

of the loop over time are shown in 

Figure 7. 

The evaporator outlet temperature 

is substantially equal to the condenser 

inlet temperature, and is stabilized at 

about 110 °C. The condenser outlet 

temperature is slightly higher than 

30 °C. The comparison of the results 

shows that the simulation results are 

in good agreement with the experimental 

results of the above three 

parameters, but the evaporator inlet 

temperature differs greatly. In the 

experimental results, the evaporator 

inlet temperature decreases rapidly 

after the start-up of PCSHP and 

remains equal to the condenser outlet. 

| Fig. 6. 

Loop pressure in start-up conditions. 




| Fig. 7. 

Fluid temperature at various positions of the loop in start-up conditions. 






| Fig. 8. 

Steady natural circulation flow rate of the loop over containment pressure in steady-state conditions. 




| Fig. 9. 

Steady natural circulation flow rate of the loop over heating power in decay heat simulation conditions. 


However, in the simulation results, 

the evaporator inlet temperature rises 

after a brief drop. This is mainly 

because the evaporator inlet temperature 

is greatly affected by the evaporator 

in the GOTHIC simulation. 

4.2 Steady-state conditions 

In the steady-state conditions, under 

different initial vacuum degrees of the 

loop, the curves of the steady natural 

circulation flow rate over containment 

pressure are shown in Figure 8 (P 0 

indicates the initial vacuum degree of 

the loop). With the same vacuum 

degree, when the containment pressure 

rises, the steady natural circulation 

flow rate of the loop tends to increase. 

This is mainly because the containment 

temperature goes up with the rise of 

containment pressure and the temperature 

difference between the evaporator 

and the containment increases. Larger 

tem perature difference results in a 

higher heat transfer coefficient in the 

evaporator, which leads to an increased 

loop circulating dive head, and a larger 

natural circulation flow rate is observed 

accordingly. This demonstrates that the 

heat transfer capacity of the system 

is highly adaptive with changes in 

containment pressure. 

It can also be seen from Figure 8 

that when at same containment, the 

higher the initial vacuum degree of 

the loop (the lower the initial loop 

pressure), the larger the steady 

natural circulation flow rate. 

It can be seen from the comparison 

of the results that the simulation 

results agree well with the experimental 

results. However, the natural 

circulation flow rate obtained by the 

simulation is relatively small compares 

to the experimental results. 

4.3 Decay heat simulation 

conditions 

In the decay heat simulation condition, 

under different initial vacuum 

degrees of loop, the curves of the 

steady natural circulation flow rate 

over heating power are shown in 

Figure 9. The simulation results are in 

good agreement with the experimental 

results. Both the simulation 

results and the experimental results 

indicate that under same initial 

vacuum degree, the steady natural 

circulation flow rate of the loop increases 

linearly with the increase of 

heating power. Generally, the heat 

transfer capacity of the system is 

positive correlated with the circulating 

flow rate. Hence, the heat transfer 

capacity of the system is highly 

adaptive with the change of decay 

heat. 

Under the same heating power, 

the steady natural circulation flow 

rate of the loop with different initial 

vacuum is basically the same. The 

initial vacuum of the heat pipe has 

little effect on the natural circulation 

flow rate when the system is stable. 





This indicates from another angle 

that heat pipe system is an adaptive 

system. Certainly, the initial vacuum 

degree has a certain influence on 

the heat transfer capacity of the heat 

pipe system, but it cannot change 

the inherent characteristics of the 

system. 

The applicability of the heat pipe 

technology on PCCS is validated by 

both the principle experiment and 

GOTHIC simulation. PCSHP is able 

to cool down and depressurize the 

containment. It is preliminarily 

judged that the heat pipe technology 

is feasible to apply to the Passive 

Containment Cooling System (PCCS) 

of nuclear power plant. 

The simulation results of the 

GOTHIC model are in good agreement 

with the experimental results. 

However, compares to the principle 

experiment, the GOTHIC model 

underestimates the heat transfer 

capacity of the PCSHP because some 

conservative assumptions are made in 

the model. 

5 Conclusion 

In this paper, a conceptual design of 

PCSHP is studied using both experimental 

and computational methods. 

The rationality of the system design 

is validated by both the principle 

experiment and GOTHIC simulation. 

It is preliminarily judged that the heat 

pipe technology is feasible to apply to 

the PCCS of nuclear power plant. 

The applicability and conservatism 

of the GOTHIC model is evaluated 

by comparing the simulating results 

with the experimental results. The 

simulation results of the GOTHIC 

model are in good agreement with the 

experimental results and the main 

parameters are within reasonable and 

credible range. The GOTHIC model 

can be used to simulate the operating 

characteristics of PCSHP, optimize 

system design and provide a foundation 

model for accident analysis and 

containment pressure and temperature 

response analysis. 

The main conclusions drawn from 

this paper are as follows: 

a) The simulation results of the 

GOTHIC model are in good 

agreement with the experimental 

results, which sufficiently verifies 

the applicability and rationality of 

the GOTHIC model; 

b) The simulation results of the 

GOTHIC model underestimates 

the overall heat transfer capacity 

of the system, which indicates 

the conservatism of the GOTHIC 

model; 

c) At the same initial vacuum degree 

of the loop, the steady natural 

circulation flow rate tends to 

increase with the rise of containment 

pressure, which shows that 

system heat transfer capacity is 

highly adaptive with changes in 

containment pressure; 

d) When the containment pressure is 

stabilized at the same value, the 

higher the initial vacuum degree of 

the loop (the lower the initial loop 

pressure), the larger the steady 

natural circulation flow rate. 

e) At the same initial vacuum degree 

of the loop, the system flow rate 

increases linearly with the rise of 

heating power, showing that the 

heat transfer capacity of the system 

is highly adaptive with the change 

of decay heat; 

f) With the same heating power, the 

steady flow rate of the loop under 

different initial vacuum is basically 

the same, which demonstrates 

from another angle that the heat 

pipe system is an adaptive system. 

The initial vacuum has a certain 

influence on the heat transfer 

capacity of the heat pipe system, 

but it cannot change the inherent 

characteristics of the system. 

References 

[1] IAEA. 2009. Passive Safety Systems and Natural Circulation in 

Water Cooled Nuclear Power Plants, IAEA-TECDOC-1624. IAEA, 

Austria. 

[2] Westinghouse Electric Company LLC, 2010. AP1000 Design 

Control Document, Revision 19. Westinghouse Electric 

Company LLC, America. 

[3] Li Jingya, Zhang Xiaoying, 2017. Simulations for cooling effect 

of PCCS in hot leg SB-LOCA of 1000 MW PWR, Nuclear 

Engineering and Design, 320, 222-234. 

[4] Chenlong Wang, Zhangpeng Guo, 2013. Transient behavior of 

the sodium-potassium alloy heat pipe in passive residual heat 

removal system of molten salt reactor, Progress in Nuclear 

Energy, 68 , 142-152. 

[5] Jean-Michel P. Tournier, Mohamed S. EL-Genk, 2005. Liquid 

Metal Heat Pipes Radiator for Space Nuclear Reactor Power 

Systems, 3rd International Energy Conversion Engineering 

Conference, San Francisco, California, August 15-18. 

[6] EPRI. 2014. GOTHIC Thermal Hydraulic Analysis Package 

User Manual, Version 8.1, Electric Power Research Institute, 

America. 

[7] Hui-Un Ha, Han-Gon Kim, 2013. GOTHIC Simulation of Passive 

Containment Cooling System, Transactions of the Korean 

Nuclear Society Spring Meeting, Gwangju, Korea, May 30-31, 

2013. 

[8] Philipp Broxtermann, Hans-Josef Allelein, 2011. Simulation of 

AP1000’s Passive Containment Cooling with the German 

Containment Code System COCOSYS, Nuclear Energy for New 

Europe 2011, 20th International Conference, Bovec, Slovenia, 

September 12-15. 

Authors 

Lu Changdong 

Cai Wei 

China Nuclear Power Technology 

Research Institute Shanghai Branch 

Shanghai 

China, 200241 

Ji Wenying 

Yang Jiang 

Wang Ting 

Cheng Cheng 

China Nuclear Power Technology 

Research Institute 

Shenzhen, Guangdong 

China, 518031 

Xiao Hong 

Nuclear and Safety Center 

Ministry of Environmental 

Protection 

Beijing 

China, 100082 




Safety Case Considerations for the Use 

of Robots in Nuclear Decommissioning 

Howard Chapman, John-Patrick Richardson, Colin Fairbairn, Darren Potter, Stephen Shackleford and 

Jon Nolan 

Decommissioning activities in the nuclear industry can often require personnel to undertake tasks manipulating 

plant, equipment and deploying tooling in close proximity to contaminated materials. 

The predominant risk associated with 

such work is exposure to radiological 

dose uptake from direct radiation, 

internal dose due to inhalation, or 

from wounds. 

There is an aspiration within the 

nuclear industry to remove the need 

for operators to undertake manual 

decommissioning activities by using 

‘robotic systems’ which offer the 

benefit of overall risk reduction safer, 

sooner and cheaper. 

A vital part of the UK Nuclear 

Decommissioning Authority (NDA) 

mission is to help drive innovation to 

address the wide-ranging complex 

challenges across their sites and 

businesses. The NDA’s ‘Grand 

Challenges’ for technical innovation 

aims to remotely decommission gloveboxes 

by 2025 and provide a 50 % 

reduction in decommissioning activities 

carried out by humans in hazardous 

environments by 2030 [1]. 

It is known that: 

“nuclear sites with their background in 

radiological substances and hazards 

have created the need for extensive safety 

measures involving the requirement 

for high integrity instrumentation and 

control measures for protection to stringent 

nuclear standards” [2]. 

This paper examines the underpinning 

Regulations, Standards and 

Technical Assessment Guides necessary 

for the deployment of ‘robotic 

systems’ to remove the need for operators 

to undertake manual nuclear 

decommissioning activities. It also 

investigates the information currently 

available to produce a safety case, 

together with commentary on work 

being undertaken by the UK National 

Nuclear Laboratory (NNL) who are 

currently reviewing technology and 

proof of concept trials to help future 

development in this area. 

Introduction 

The civil nuclear industry worldwide 

is regulated to ensure that activities 

related to nuclear energy and 

ionising radiation are conducted in a 

manner which adequately protects 

people, property and the environment. 

In the UK, the Office for Nuclear 

Regulation (ONR) is the agency 

responsible for the licensing and 

regulation of nuclear installations and 

the legal framework for the nuclear 

industry is based around the Health 

and Safety at Work Act (HSWA) [3], 

the Energy Act [4] and the Nuclear 

Installations Act (NIA) [5]. 

A fundamental requirement cited 

in the legislation is that risks be 

reduced to As Low As Reasonably 

Practicable (ALARP). This principle 

provides a requirement to implement 

proportionate measures to reduce risk 

where doing so is reasonable. The 

ALARP principle is applied by adhering 

to established good practice, or in 

cases where this is unavailable, it is 

applied to demonstrate that measures 

have been implemented up to the 

point where the cost of additional risk 

reduction is disproportionate to the 

benefit gained [6]. 

The aspiration to use robots in the 

nuclear industry requires hazards to 

be safely managed and the risks 

demonstrated to be ALARP. This paper 

investigates how this might be 

achieved to ensure all potential 

hazards are identified and prevented, 

with key safety measures recognised, 

implemented and maintained in an 

appropriate and pragmatic manner, 

benefitting from experience gained 

from wider industry. 

Outside of the nuclear industry 

industrial robots are found increasingly 

in the workplace where it is 

widely acknowledged that robot 

movements can have the potential to 

cause human physical harm and 

damage to other equipment. Deployment 

of robots in the nuclear industry 

also raises further concern that impact 

events may have the potential to result 

in loss of containment of nuclear 

material, and cause damage to nuclear 

safety significant equipment and 

instrumentation. 

Operators and equipment must 

therefore be protected against the 

robot. The strict segregation of man 

and robot has previously been 

employed in wider industry as a key 

Hazard Management Strategy (HMS) 

to protect workers. The robot 

remained enclosed in a controlled 

area while it performed its tasks. In 

the present day, thanks to a new 

generation of robots and technologies 

segregation may no longer be necessary 

if the potential for collision is not 

perceived as being hazardous [14]. 

Assessment of hazards 

Robot systems regulation 

and legal requirements 

The European Union (EU) formulates 

general safety objectives via a large 

number of directives, (circa 30 active 

directives currently available). However, 

only a small selection of directives 

are relevant to a typical machine 

builder and the safety objectives are 

more precisely specified through 

standards [14]. 

The standards have no direct legal 

status on their own until they are 

referenced in domestic laws and 

regulations. In practice manufacturers 

of robotic Commercial off the 

Shelf (COTS) equipment use the 

“Conformité Européenne” (CE) mark 

to document the fact that all relevant 

European directives have been applied 

and appropriate conformity to all 

assessment procedures achieved [14]. 

Based on the European Parliament 

and Council of the European Union 

Machinery Directive 2006/42/EC [7], 

a robot system is considered to be 

partly completed machinery. This 

means that robot systems require CE 

marking. The person placing the 

machine into a specific application is 

known as the ‘integrator’ and must 

perform the conformity assessment 

procedure to conclude a Declaration 

of Conformity [14]. 

Other useful documents include 

the International Organization for 

Standardisation (ISO) 12100 [8] for 

risk assessment; ISO 13849 part 1 

[9]; or International Electrotechnical 

287 

DECOMMISSIONING AND WASTE MANAGEMENT 


Safety Case Considerations for the Use of Robots in Nuclear Decommissioning ı Howard Chapman, John-Patrick Richardson, Colin Fairbairn, Darren Potter, Stephen Shackleford and Jon Nolan


DECOMMISSIONING AND WASTE MANAGEMENT 288 

| Fig. 1. 

Overview of robot systems regulation and standards. 

SIL PFD RRF 

1 1 in 10 – 1 in 100 10 to 100 

2 1 in 100 – 1 in 1,000 100 to 1,000 

3 1 in 1,000 – 1 in 10,000 1,000 to 10,000 

4 1 in 10,000 – 1 in 100,000 10,000 to 100,000 

| Tab. 1. 

Relationship between SIL, PFD and RRF [18]. 

Commission (IEC) 62061 [10] for the 

functional safety requirements. 

Two standards from the ISO 10218 

“Safety of Industrial Robots” Part 1 

[11]: “Robots” and Part 2 [12]: “Robot 

systems and integration” are listed 

under the Machinery Directive 

2006/42/EC [5] to specify detailed 

safety requirements. ISO 10218-1 is 

solely concerned with the actual robot 

system, whilst in contrast to this, ISO 

10218-2 expands to the entire robot 

application [14]. 

In practice the standards above 

have proved to be insufficient in their 

own right when it comes to safely 

implementing an actual Human and 

Robot Collaboration (HRC). Protective 

measures for HRC are therefore 

currently identified through ISO/ 

TS15066 [13] in order to help production 

technicians and safety experts 

in the development of safe shared 

workspaces and the risk assessment 

process. This describes four types of 

collaboration reproduced below [14] 

as protection principles to ensure human 

safety is guaranteed at all times 

during collaborative operation [14], 

as shown in Figure 1: 

1: Safety-rated monitored stop 

Here, the human only has access to 

the robot once stopped and the 

robot system must not start up again 

automatically and unexpectedly. 

2: Hand guiding 

In this case the human only has access 

to a stationary robot. The hand 

guiding of the robot system can only 

be enabled by manually operating an 

enabling device. 

3: Speed and separation monitoring 

With this method, the distance 

between human and robot is permanently 

monitored by a sensor. The 

robot system moves with correspondingly 

safely reduced speed. The closer 

the human gets to the robot, the 

slower the robot becomes. If the 

distance is too short, a safety stop is 

triggered. 

Safety is guaranteed in the first 

three methods by maintaining the 

distance between human and robot, to 

avoid collision. When implementing 

one of these three methods, no special 

HRC robots are necessary. Standard 

industrial robots can be used that are 

equipped with corresponding safety 

packages for speed monitoring, or 

workspace monitoring by the manufacturer. 

4: Power and force limiting 

In contrast to methods one to three, 

contact between human and robot is 

possible under certain circumstances 

in the case of method four. However, 

the manufacturer of the application 

must guarantee that the collision 

between human and robot is not 

hazardous. The manufacturer of the 

application confirms this with a 

signature on the declaration of conformity. 

Risk assessment 

To ensure robot safety, manufacturers 

and users normally apply a threestage 

risk assessment approach 

detailed in ISO 12100 reproduced 

below [8] as follows; 

(i) Inherent safe design measures 

(hazard elimination); 

(ii) Safeguarding and complementary 

protective measures (fixed guards, 

movable guards with interlocks, 

safety devices); and 

(iii) Information for use (safe working 

practices for the use of the machinery, 

warning of residual risks, recommended 

Personal Protective 

Equipment (PPE)). Residual risk is 

then managed by the user. 

The performance requirement of 

safety measures is set out in ISO 10218, 

which also mentions com pliance with 

Safety Integrity Levels (SILs) which 

comes from voluntary International 

Electrotechnical Commission standards 

used by plant owners/operators 

to quantify safety performance requirements 

for hazardous operations 

[15]; including IEC 61508: Functional 

Safety of Elec trical/Electronic/Programmable 

Electronic Safety-Related 

Systems [16]. 

Four SILs are defined in these 

standards, with SIL 4 the most 

dependable and SIL 1 the least. The 

applicable SIL is determined based on 

a number of factors and is an exercise 

in risk analysis, where the risk associated 

with a specific hazard is 

calculated without beneficial risk 

reduction. The unmitigated risk is 

then compared against a tolerable risk 

target [17]. 

The amount of risk reduction 

required to achieve a tolerable risk is 

known as a Risk Reduction Factor 

(RRF) and can be correlated to a SIL 

number and Probability of Failure on 

Demand (PFD) for protection systems 

(the relationship between each is 

outlined in Table 1). Each order of 

magnitude of risk reduction that is 

required essentially correlates with an 

increase in one of the required SIL 

numbers [18] as shown in Figure 2. 

| Fig. 2. 

SIL as a function of hazard frequency and severity. 

Assessment 

of radiological hazards 

Radiological safety assessments 

follow a rigorous process and are required 

as part of Nuclear Instal lations 

Site Licence Conditions. 

The fundamental requirement in 

any nuclear decommissioning safety 




case involving robot systems will be to 

demonstrate that hazards presenting 

radiological exposure; loss of containment 

of nuclear material; and 

damage to nuclear safety significant 

equipment and instrumentation can 

all be safely managed, and also that 

the identified risks are deemed 

ALARP. 

A clear link of how the assessment 

will be implemented is known as the 

‘Golden Thread’. This can be achieved 

through a Claims Arguments Evidence 

(CAE) approach, as illustrated in 

Figure 3. From a robotic CAE perspective, 

there is top level claim requirement 

to ensure all robot systems can 

be safely managed and the risks are 

ALARP. This is supported by a series of 

sub-claims listed below: 

p All robot system hazards can be 

identified, and potential hazards 

understood. 

p All robot system hazards can be 

adequately prevented or managed, 

by determining the unmitigated 

consequences such that appropriate 

safety measures can be 

identified and the risks can be 

shown to be ALARP. 

p All key operational and engineering 

measures can be identified, 

implemented and maintained. 

The foundation of a HMS in a nuclear 

robot system safety case will be 

based upon a standard hierarchical 

approach to safety. This starts with 

elimination of the hazard wherever 

possible, followed by substitution to 

replace the hazard, isolation of people 

from the hazard, administrative control, 

with reliance upon PPE being 

the weakest and therefore least 

favour able HMS as shown in Figure 4. 

It is argued that in the context of 

nuclear robot systems which operate 

re motely, the use of PPE is not 

necessarily relevant unless it relates to 

the need for human intervention, for 

example during repair or main tenance 

work. 

The approach for developing a 

robot system safety case is summarised 

as: 

p Identification of hazards; 

p Assessment of hazards and 

identification of suitable safety 

measures; 

p Substantiation of safety measures; 

and 

p Implementation of safety 

measures. 

A structured and systematic examination 

of robot systems will be undertaken 

using HAZard and OPerability 

(HAZOP) studies to identify potential 

problems that may represent risks to 

| Fig. 3. 

Claims arguments evidence approach. 

personnel, or equipment, or prevent 

efficient operation [20]. 

Hazards are then assessed, and 

safety measures are identified in the 

safety case. 

The HMS developed for the robot 

system will be used to identify safety 

measures which are proportional to 

hazard severity and demonstrate 

there is sufficient strength in depth 

and the risk is ALARP. 

The individual hazards identified 

by HAZOP will be presented in the 

form of a number of fault sequences. 

Each fault sequence starts with an 

initiating event that could lead to 

unwanted consequences and place a 

demand on a set of safety measures. 

The assessment of the fault sequence 

included failure of some or all of these 

safety measures. 

Radiological safety assessments 

specify the Engineering and/or 

Operational Safety Measures that 

need to be in place to minimise the 

risks to acceptable levels, i.e. ALARP 

and ensure the adequacy of safety. 

The concept of defence in depth is 

fundamental to radiological safety to 

prevent accidents and if prevention 

fails, to limit potential consequences. 

For significant faults Design Basis 

Analysis (DBA) requires the designation 

of a passive safety measure, 

(such as an enclosure wall), or two 

key independent safety measures, 

(such as high integrity Control, Electrical 

and Instrumentation Equipment 

(CE&I)) with predefined action on 

failure and substitution arrangements. 

Alternatively, it is possible in 

some instances for Operational Safety 

| Fig. 4. 

Hierarchy of controls, (by the National Institute of Occupational Safety and Health) [19]. 






Measures to be claimed, which must 

be carried out to prevent possible 

harm /dose uptake. 

For lesser significant faults, DBA 

requires the designation of one safety 

measure, which can either be passive, 

or an item of CE&I equipment that 

does not need to have any predefined 

action on outage or substitution 

arrangements. Alternatively, it is 

possible to in some instances for 

Operational Safety Measures, about 

operator actions, or plant conditions 

to be claimed which support the safety 

case. 

The various engineering safety 

measures in the safety case are 

uniquely identified as a Structure, 

System, or Component (SSC), and the 

safety function and performance 

requirement of each is recorded in an 

Engineering Schedule and substantiated 

against their required Safety 

Function, Performance Requirement 

and PFD. The operational safety 

measures and compliance arrangements 

are defined within a Clearance 

Certificate. 

However, a fault sequence is initiated, 

it is also important to identify 

the involvement of any Programmable 

Electronic Systems (PES) in protection/mitigation 

as the system may not 

be capable of substantiation, ultimately 

requiring a different safety 

measure to be defined. 

PES contain both hardware and 

software. Software is different from 

hardwired systems in that it has a 

greater potential for a number of 

systematic failures (as opposed to 

random failures) which may remain 

| Fig. 5. 

Future deployment of robot systems for decommissioning activities 

operated within a virtual enclosure. 

unrevealed for many years. Knowledge 

of the failure of a PES is usually only 

identified when the system fails in 

operation, because they employ 

hierarchal coding and identification of 

sequential coding errors are usually 

difficult. 

Where the PES controls a process, 

the liability to initiate fault sequences 

must be recognised in the safety 

assessment, and an ‘initiator type’ 

safety function defined. Where a PES 

initiates a fault sequence, no credit 

may be claimed for protection by the 

same PES in the same fault sequence. 

Therefore, dependency upon PES 

for protection/mitigation should be 

minimised wherever possible. 

PES should be distinguished from 

SMART Instruments – although the 

latter include some software (sometimes 

referred to as ‘firmware’), they 

are arguably very little different from 

the hardwired (‘dumb’) instruments. 

Unlike a PES, SMART instrument 

software can be simulated, run 

inactively or actively with real-time 

communication between execution 

and operation limit. SMART instrument 

software may only be altered 

using configured operator parameters, 

allowing the opportunity to 

remove any potential coding error 

identified and the opportunity for 

multiple level recovery. Hence SMART 

instrumentation is not prone to the 

same level of systematic failure. 

There is currently little specific 

data for PES/SMART reliability available 

for the purposes of making a 

nuclear decommissioning safety case. 

This results in some frequency estimates 

(for comparison with criteria) 

that are over-estimated in comparison 

to reality, but this drawback is not as 

significant as using reliability figures 

that cannot easily be justified. 

Any risk reduction benefit claimed 

for PES/SMART is currently generally 

limited. For example, a PES would 

normally be claimed within a possible 

PFD range of unity to 1 in 30. 

For nuclear decommissioning 

purposes, substantiation of PES and 

SMART systems is achieved through 

interpretation of the relationship 

between PFD and SIL requirements 

contained in IEC61508 [16]. 

Assessment of robot 

systems for decommissioning 

activities 

There appears to be an understanding 

in wider industry that stringent standards 

for nuclear decommissioning 

places a requirement for CE&I safety 

measures to be substantiated to SIL 3, 

or even 4 to meet the designation of 

high integrity protection systems. The 

dilemma in the nuclear industry is 

often a choice of placing reliance upon 

a single but complex safety measure, 

versus multiple layers of safety 

measures. Complex systems typically 

demand significant effort, and therefore 

cost more, to substantiate and 

maintain, compared with systems 

involving multiple layers. 

For the majority of nuclear decommissioning 

cases the integrity level 

designated to each individual hardwired 

‘dumb’ CE&I layer of protection 

is usually no more than SIL 1 in practice, 

which provides a PFD of 1 in 100 

and a risk reduction of 100 for each 

layer. Only in rare cases have claims 

been made on SIL 2 CE&I safety protection 

systems. It is argued that the 

substantiation process would prove 

far too onerous to achieve SIL 3, or SIL 

4 level of integrity. 

One common mis-understanding 

appears to be in the interpretation of 

safety integrity claims made upon 

multiple layer protection systems. 

An example multiple layer protection 

system arbitrarily consisting of 3 

layers of protection is used to exemplify 

the mis-understanding. Architectures 

with 3 layers of CE&I protection 

are not the same as a SIL 3 system and 

should be substantiated as a series of 

3 x SIL 1 separate systems. It is argued 

that the safety integrity level of such 

circumstances should default to the 

lowest common denominator, i.e. SIL 

1, or possibly SIL 1 + 1 in rare circumstances. 

A robot system recently deployed 

by NNL at its Preston Laboratory 

included the use of a robot controlled 

5kW laser which enabled selective, 

semi-autonomous controlled laser 

cutting for disassembly in confined 

spaces [20]. This capability consisted 

of a KUKA KR series robot which 

operated in an enclosure with a SIL 1 

rated hardwired door interlock 

system, which disallowed laser activation 

and robotic movement if anyone 

attempted to access the enclosure 

during usage. 

Multiple safety systems focused on 

limiting the robot’s movement to a 

controlled safe working area. This 

provided additional laser firing safety 

inputs and reducing the amount of 

human intervention required in order 

to reduce rig downtime. The KUKA 

robot included physical hard-stops 

installed in each robot joint which 

helped reduce potential damage to 

the enclosure, as well as limiting its 

working area. 




Based on the methods described 

earlier for HRC, the NNL robot system 

safety case at Preston Laboratory 

ultimately relied primarily upon 

claims on ‘dumb’ hardwired door 

interlock systems and physical end 

stops, rather than claims on robot 

SMART systems. 

It is recognised that future deployment 

of robot systems for decommissioing 

activities may not benefit 

from physical enclosures, and will 

require hazard management strategies 

moving towards methods 

described previously under HRC 3 or 4 

to prevent potential collisions. 

NNL are currently reviewing 

available industry wide SMART technology 

together with proof of concept 

non-active commissioning trials, to 

support the necessary substantiation 

to achieve a SIL 1 rating for individual 

layers within a diverse multi layer 

protection system. It is argued that 

such an approach could prove useful 

to create vitual enclosures (as shown 

in Figure 5), allowing HRC 3 or 4 for 

nuclear decommissioing. 

Historically most of the ISO standards 

defined for robot systems have 

been developed singularly for the 

automotive industry with the opportunity 

for human intervention for 

teach and repeat. Future deployment 

of SMART robot systems for decommissioning 

activities enable the 

opportunity for the review and monitoring 

of sequences with constant 

communication to the robot prior, 

during and after the execution of 

operations. 

Path forwards 

This paper has examined the 

underpinning Regulations, Standards 

and Technical Assessment Guides 

necessary for the deployment of 

‘ robotic systems’ to remove the 

need for operators to undertake 

manual nuclear decommissioning 

activities. 

It is NNL’s view that consideration 

of the approach taken for the 

robot systems outside of traditional 

industrial settings, for example their 

use in medical applications, may have 

useful applicability for safety in harsh 

nuclear decommissioning environments 

and HRC 3 or 4 interaction. 

NNL believes the adoption of HRC 

3 or 4 methods for decommissioning 

purposes will require a change in the 

way the nuclear industry views the 

reliability of SMART protective layers. 

This will be achieved by striking a 

balance between risk versus the 

benefits gained from using robot 

systems. A challenge to the current 

position of high risk and low confidence 

in SMART protective layers 

will offer the potential for decommissioning 

risk reduction safer, sooner 

and cheaper. 

The forthcoming NNL review of 

wider industry SMART instrument 

applications will make reference to 

any guidance currently in the process 

of being established by the International 

Atomic Energy Agency 

(IAEA), due for publication later in 

2020. It is expected that the IAEA 

guidance will provide a common 

technical basis of how to design, 

select and evaluate candidate SMART 

devices for their safe use in nuclear 

safety systems, including instrumentation 

and control, electrical, 

mechanical and other areas [21]. 

NNL aims to improve on the 

current position by establishing a 

higher degree of confidence in SMART 

protection systems, which can provide 

a safety function to prevent impact 

causing harm to humans and equipment 

resulting in loss of containment 

of nuclear material. This will be 

supported by a safety performance 

requirement to operate within 

specified distances within a virtual 

enclosure to ensure the risk of 

generating a hazardous collision 

between robot, human and equipment 

is reduced to ALARP. 

The intention is to ensure the 

science becomes a robust, safe and 

efficient engineered solution for 

nuclear industry decommissioning 

activities and achieve UK NDA’s 

‘Grand Challenges’. 

References 

1. https://nda.blog.gov.uk/2020/01/31/the-ndas-grandchallenges-for-technical-innovation/ 

2. National Nuclear Laboratory “A Pragmatic Approach to 

Chemotoxic Safety in the Nuclear Industry”, H Chapman, 

Marc Thomas, Stephen Lawton, ATW-International Journal for 

Nuclear Power, Issue 8/9/2019 

3. United Kingdom Government, “Health and Safety at Work Act,” 

1974 

4. United Kingdom Government, “Energy Act,” 2013 

5. United Kingdom Government, “Nuclear Installations Act,” 1965 

6. https://www.hse.gov.uk/risk/theory/alarpglance.htm 

7. European Parliament and Council of the European Union 

Machinery Directive 2006/42/EC 

8. International Organization for Standardization ISO 12100 

“Safety of Machinery General Principles for Design – Risk 

assessment and Risk Reduction”, 2010 


part 1 “Safety of Machinery – Safety Related Parts of Control 

Systems” Part 1 General Principles of Design, 2015 

10. International Electrotechnical Commission IEC 62061 “Safety 

of Machinery – Functional Safety of Safety Related Electrical – 

Electronic and Programmable Electronic Control Systems”, 

2015 

11. International Organization for Standardization; ISO 10218 

“Safety Requirements for Robot System in an Industrial 

Environment” Part 1, Robot, 2011 


“Safety Requirements for Industrial Robots” Part 2, Robot 

Systems Integration, 2011 

13. International Organization for Standardization ISO/TS15066 

“Robots and Robot Devices – Collaborative Robots”, 2016 

14. https://www.pilz.com › TechBo_Pilz_safety_compendium_ 

1004669-EN-01, 5 th Edition March 2018 

15. https://www.crossco.com/resources/articles/determiningsafety-integrity-levels-for-your-process-application/ 

16. International Electrotechnical Commission IEC 61508 

“ Standard for Functional Safety of Electrical/Electronic/ 

Programmable Electronic Safety Related Systems”, 2010 

17. Petroleum Refining Design and Applications Handbook 

Volume 1. A. Kayode Coker. © 2018 Scrivener Publishing LLC. 

Published 2018 by John Wiley & Sons, Inc.17 on line 

library.wiley.com 

18. Honeywell Plant and Personnel Safety Control Engineering 

2019 eBook Series 

19. “‘Hierarchy of Controls’. U.S. National Institute for 

Occu pational Safety and Health. Retrieved 2017-01-31.,” 

[Online] 

20. National Nuclear Laboratory “Laser Cutting for Nuclear 

Decommissioning An Integrated Safety Approach”, 

H Chapman, Stephen Lawton, Joshua Fitzpatrick, 

ATW – International Journal for Nuclear Power, 63 Issue 10 

2018 

21. https://www.world-nuclear-news.org/Articles/ 

IAEA-addresses-safety-of-smart-devices-in-nuclear 

Authors 

Howard Chapman 

John-Patrick Richardson 

Colin Fairbairn 

Darren Potter 

Stephen Shackleford 

Jon Nolan 

National Nuclear Laboratory 

Limited 

5 th Floor, Chadwick House, 

Birchwood Park, Warrington, 

WA3 6AE 

United Kingdom 





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und wünscht ihnen weiterhin alles Gute! 

Juni 2020 

55 Jahre | 1965 

04. Dipl.-Phys. Jan-Christian Lewitz, 

Dresden 

60 Jahre | 1960 

14. Dipl.-Ing. Hermann Altendorfer, 

Essenbach 

81 Jahre | 1939 

06. Dr. Peter Drehmann, Kornwestheim 

07. Dr. Peter Antony-Spies, Liederbach 

10. Dipl.-Ing. Reinhard Seepolt, Hamburg 

14. Dr. Gustav Meyer-Kretschmer, Jülich 

23. Dr. Rolf Krieg, Karlsruhe 

82 Jahre | 1938 

25. Dipl.-Ing. Horst Roepenack, Bruchköbel 

87 Jahre | 1933 

12. Prof. Dr. Carsten Salander, Bad Sachsa 

88 Jahre | 1932 

28. Hans Schuster, Aachen 

94 Jahre | 1926 

27. Dipl.-Ing. Heinz-Arnold Leising, 

Bergisch Gladbach 

KTG Inside 

Verantwortlich 

für den Inhalt: 

Die Autoren. 

Lektorat: 

Natalija Cobanov, 

Kerntechnische 

Gesellschaft e. V. 

(KTG) 


10115 Berlin 

T: +49 30 498555-50 

F: +49 30 498555-51 

E-Mail: 

natalija.cobanov@ 

ktg.org 

www.ktg.org 

70 Jahre | 1950 

21. Dr. Sieghard Hellmann, Grossenseebach 

76 Jahre | 1944 

08. Jürgen Fabian, Büsingen am Hochrhein 

24. Hans-Jürgen Schlesinger, Essen 

78 Jahre | 1942 

10. Ing. Wolfgang Feltes, 

Bergisch Gladbach 

79 Jahre | 1941 

15. Dr. Frank Depisch, Erlangen 

80 Jahre | 1940 

04. Dipl.-Phys. Hans-Peter Dyck, Forchheim 

13. Dr. Heinz Hoffmann, Einhausen 

83 Jahre | 1937 

10. Dipl.-Phys. Reinhard Wolf, 

Grosskrotzenburg 

84 Jahre | 1936 

12. Dipl.-Ing. Heinz Malmström, Ahaus 

24. Dipl.-Ing. Christian-Theodor Körner, 

Breitenbronn 

30. Kai-Michael Pülschen, Erlangen 

85 Jahre | 1935 

08. Dr. Ing. Heinrich Löffler, Wennigsen/CH 

08. Ing. Karl Rudolph, Wettingen 

17. Dipl.-Ing. Peter Gottlob, 

Stutensee-Friedrichstal 

23. Dipl.-Ing. Werner Schultz, Hirschberg 

22. Dipl.-Ing. Johann Pisecker, Tulln 

Nachträgliche 

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NEWS 

Top 

Nuclear power supports clean 

energy transition with secure 

and flexible electricity supply 

(iaea) With a transition underway in 

the global energy industry to reduce 

greenhouse gas emissions and stem 

climate change, countries are looking 

at ways to ensure a continuous 24/7 

supply of clean electricity while avoiding 

power blackouts and disruptions 

to other critical facilities, such as 

public transport and medical care. 

Nuclear power is one solution, as 

the International Energy Agency 

noted this week in a commentary on 

how the Covid-19 crisis also highlights 

the need for a secure and flexible 

electricity supply. 

As countries increasingly turn to 

solar and wind to generate electricity, 

flexibly operated nuclear power plants 

(NPPs) can provide a reliable stream 

of low carbon power as well as fill the 

output gaps left when variable 

renewable sources (VREs) lack 

sunshine or wind. Likewise, NPPs can 

adapt their power production when 

renewable generation varies. This 

balancing act, known as non-baseload 

operation, can ensure the supply of 

power and limit the risk of disruptions 

by enhancing the reliability of the 

electrical grid. 

But this flexibility comes at a cost. 

Most existing NPPs are best run at 

full or “baseload power” because 

with high upfront costs but very low 

operating costs, their economics 

depend on running close to capacity 

over many years. 

“Flexible operation results in 

higher operation and maintenance 

costs, and the magnitude of those 

costs will depend on the grid system’s 

flexibility needs,” said Nikhil Kumar, a 

contributor to a forthcoming IAEA 

report on the economics of flexible 

operation and Managing Director at 

U.K.-based Intertek, an assurance, 

inspection, product testing and certification 

company. “These costs increase 

as the depth and periodicity of load 

following increases.” 

France, where NPPs provide 

almost three-quarters of the country's 

electricity, has years of operational 

experience adjusting output based 

on electricity demand. Around twothirds 

of France's NPPs utilize load 

following and frequency control on a 

regular basis, which helps minimize 

the days per year in which electricity 

generation exceeds demand. 

Germany also uses load following 

and frequency control to respond to 

market demand and ensure grid 

stability. Load following NPPs have 

KTG Inside


helped integrate greater shares of 

variable renewable sources, which 

produced almost half of Germany’s 

electricity last year and are expected 

to further expand in years to come. 

“The outstanding issue in many 

power markets is what kind of value to 

assign to these services for the grid,” 

said Victoria Alexeeva, an energy 

economist at the IAEA. “In the absence 

of an adequate valuation for such 

services, nuclear power’s economic 

competitiveness is reduced,” added 

Nesimi A. Kilic, an IAEA nuclear 

engineer. 

Amid the clean energy transition, 

electrical grids may face different 

challenges. 

Last August, for example, the UK 

suffered its most severe power outage 

in more than a decade – a blackout 

of between 15 and 50 minutes for 

more than a million customers that 

disrupted some passenger trains and 

caused a temporary loss of power at 

one hospital and airport. In a report 

last month, Germany’s grid operators 

said the country may need to import 

electricity at times in the coming 

years as firm sources such as coal and 

nuclear are retired. 

The IAEA supports countries in 

understanding all relevant aspects 

of flexible NPP operation through 

publications, workshops and technical 

meetings, including one held in 

Phoenix in the U.S. state of Arizona in 

December 2019. Around 60 plant 

operators, regulatory officials and 

policymakers from 10 countries 

discussed “future energy needs and 

proactive actions that would ensure 

nuclear power plants continue to 

provide clean, affordable and reliable 

power to people around the world,” 

said Robert Bement, Executive Vice 

President and Advisor to the Chief 

Executive Officer at Arizona Public 

Service, which hosted the meeting. 

The IAEA is also working with 

governmental and non-governmental 

bodies, including the Flexible Nuclear 

Campaign for Nuclear-Renewables 

Integration. The campaign – a project 

of the Clean Energy Ministerial led by 

the Nuclear Innovation: Clean Energy 

Future (NICE Future) initiative – seeks 

to model revenue for flexible 

NPPs, including costs and technical 

requirements. 

“A clear understanding of how 

flexible integrated energy systems – 

that include both nuclear and renewable 

energy – can meet our future 

energy needs must be developed and 

communicated,” said Kelly Lefler, a 

Senior Advisor at the U.S. Department 

of Energy’s Office of Nuclear Energy. 

“Technical meetings and other 

initiatives by the IAEA and the Clean 

Energy Ministerial bring together 

governments, research institutions, 

non-governmental organizations and 

industry to explore innovative clean 

energy solutions with nuclear power.” 

| Jeffrey Donovan and Matt Fisher, 

IAEA Department of Nuclear Energy 

www.iaea.org (201121401) 

World 

The Versatile Test Reactor 

can help unlock the future 

of carbon-free energy 

(nei) The 2020s will be the decade of 

innovations in nuclear energy. The 

technologies and tools that will enable 

advanced nuclear reactors to become 

a reality are being developed now. 

The U.S. Department of Energy’s 

Versatile Test Reactor (VTR) is one of 

those cutting-edge, specialized tools. 

Just getting under way, the VTR is 

intended to mimic the conditions that 

would exist in a category of advanced 

reactors now under development: 

fast reactors, which include sodiumcooled 

fast reactors, molten salt 

reactors and high-temperature gas 

reactors. 

With a pressing need to reduce 

carbon emissions and a growing 

worldwide demand for electricity, it is 

urgent to commercialize advanced 

reactor technologies, many of which 

use molten salt, sodium or helium gas 

(instead of water, as current plants 

do). 

Fast reactors are quite different 

than the reactors currently operating 

in the United States. When they run, 

the neutrons – subatomic particles 

that sustain the chain reaction – are 

moving with vastly more energy than 

in today’s reactors, in some cases with 

100,000 times more energy. 

Those more energetic neutrons 

have many advantages. They can split 

a much wider variety of atoms to make 

energy, including many atoms that 

were produced in today’s reactors and 

would otherwise be considered waste. 

They can run reactors that operate at 

much higher temperatures than are 

common today, which would produce 

steam that can be used for many more 

purposes. And many of those designs 

would run at far lower pressures, 

making them easier and less expensive 

to build. 

There is a catch, though. No one 

is completely sure how all of the 

components of these new reactors 

would behave after a few decades in 

the stew of high-energy neutrons. And 

engineers don’t want to wait to find 

out. 

With a simulated environment, 

engineers can bathe the components 

in neutrons at a pace three or four 

times faster than they would see in an 

actual power reactor, pull the parts 

out for evaluation, and if necessary, 

make changes and try again. This is 

exactly what the VTR would provide. 

“We want to do a quick screening 

of these technologies,” said Kemal 

Pasamehmetoglu, executive director 

of the VTR project. 

In fact, the reactor could also be 

used to test materials for other 

industries and for materials that could 

be useful in today’s reactors. 

To prosper, experts say the U.S. 

needs its own high-tech test facility for 

fast neutrons. 

“The nuclear leadership that we 

had in the world derived from our 

technical leadership,” said Irfan Ali, 

who is on the board of directors 

of Advanced Reactor Concepts, a 

sodium -cooled reactor developer. “For 

us to maintain that, we have to keep 

moving forward with the technology.” 

Because of the lack of testing 

facilities in the United States, Terra- 

Power LLC, the company backed by 

Bill Gates, has had to use a reactor in 

Russia, the BOR-60. But access to 

that reactor, and problems moving 

irradiated materials across international 

borders, make that a cumbersome 

route. 

Congress gave initial approval to a 

versatile neutron source in the Nuclear 

Energy Innovation Capabilities Act, 

signed into law in September 2018. 

Two companies have already submitted 

a proposal to develop the 

reactor. 

Once completed, the Versatile Test 

Reactor would enable the development 

of these fast reactors. Along with 

other types of advanced reactors, the 

next generation of nuclear will power 

our way of life into the future, without 

carbon emissions. 

| www.nei.org (201121455) 

Research 

Wendelstein 7-X fusion device 

at Greifswald to be upgraded 

(ipp-mpg) The next round of the 

stepwise expansion of the Wendelstein 

7-X fusion device at Max Planck 

Institute for Plasma Physics (IPP) at 

293 

NEWS 

News


294 

NEWS 

| Inside the plasma vessel: The previous cladding with carbon tiles has been 

abandoned; the vessel is ready for installation of the new water-cooled 

wall protection. 

Photo: IPP, Torsten Bräuer 

Greifswald is in full swing. Watercooled 

inner cladding of the plasma 

vessel will make the facility suitable 

for higher heating power and longer 

plasma pulses. Production of the new 

cladding’s centrepiece, the so-called 

divertor, was taken over by the 

institute’s Garching branch. For 

tomorrow, final delivery to Greifswald 

is scheduled, where the preparations 

for installation of the components 

have been completed. The installation 

work will last until well into next 

year. Wendelstein 7-X, the world's 

largest stellarator fusion device, is to 

investigate the suitability of such 

devices for power plants. 

At the end of 2018, the experiments 

on Wendelstein 7-X were 

temporarily terminated after two successful 

work phases (see PI 11/18). 

Upgrading of the plasma vessel has 

been ongoing since then. “First of all, 

most of the old components had to be 

taken out. Installation of the new ones 

can now begin,” says Prof. Dr. Hans- 

Stephan Bosch, whose division is 

responsible for technical operation of 

the device. Whereas most of the 

wall protection components were 

previously operated uncooled, large 

sections of the wall will be watercooled 

starting with the next round of 

experiments: “This will then enable 

Wendelstein 7-X to generate plasma 

pulses lasting up to 30 minutes”, states 

Professor Bosch. 

Centrepiece of the new wall 

cladding is the so-called divertor, the 

most heavily loaded component of the 

plasma vessel. In ten double strips on 

the inner wall of the plasma vessel, 

the divertor tiles follow the curved 

contour of the plasma edge. They 

protect those wall areas to which 

particles from the edge of the plasma 

are magnetically directed. A pump 

behind a gap in the middle of each 

double strip removes the impinging 

plasma and impurity particles. In this 

way, the divertor can be used to 

control the purity and density of the 

plasma. 

Demanding manifacture 

In the high-performance experiments 

planned, the new water-cooled 

divertor plates, which replace the 

previous uncooled ones, are designed 

to withstand a load of up to ten 

megawatts per square metre – like a 

space shuttle re-entering the Earth’s 

atmosphere. Without water cooling, 

however, the heat-resistant divertor 

tiles made of carbon-fibre-reinforced 

carbon could not withstand this load 

for the intended 30-minute plasma 

pulses. They are therefore welded 

onto water-cooled plates made of 

a copper-chromium-zirconium alloy. 

The coolant, supplied by small steel 

tubes, ensures that the heat energy is 

removed. 

Each of the ten curved divertor 

strips consists of twelve of these 

plates, which in turn are composed of 

individual elements. In total, these 

890 elements comprise almost half a 

million individual parts, from the 

heat-resistant surfaces to the special 

screws. 

The high-performance components 

are the result of a long 

development, manufacturing and 

testing process carried out by the 

Integrated Technical Centre (ITZ) and 

the “Components in the Plasma 

Vessel” work group at IPP in Garching 

in cooperation with industrial 

com panies. “The complex geometry of 

the components was particularly 

challenging, given the high level of 

accuracy and reliability required,” 

explains IPP engineer Dr. Jean 

Boscary, who headed production and 

assembly of the “big puzzle”: “There 

should be no water leakage later in 

Wendelstein 7-X”. 

Accordingly, already the preparatory 

work was extensive: In 

2003, the development and production 

contract for the divertor 

elements was concluded with an 

industrial company. After four preseries 

and more than 60 prototypes, 

five years of series production could 

start in 2009. 

To complete a divertor element, 82 

manufacturing steps and 44 tests 

were necessary. The surface of each of 

the 16,000 carbon tiles had to be 

milled three-dimensionally into shape 

– with tolerances of sometimes only 

0.1 millimetres to avoid any overheating 

of protruding edges. The 

joining technique between carbon and 

copper alloy was specially developed 

for Wendelstein 7-X. 

At IPP in Garching, the divertor 

elements were then joined together 

on steel frames to form plates. Cooling 

pipes and cooling water distributors 

were joined by means of a special 

welding technique developed at the 

ITZ: “Among the 2,000 welded joints, 

the subsequent tests were only able to 

detect two leaks,” says Dr. Boscary. In 

other respects, too, there were always 

quality assurance tests between the 

individual work steps. For production 

control, for example, the load capacity 

of the parts was examined in 

Garching's GLADIS heat test rig. 

The experience gained in this “largest 

heat protection project in fusion 

research to date” is unique worldwide, 

Jean Boscary emphasizes. All ten 

divertor strips have now been completed. 

A major part has already 

been delivered to IPP at Greifswald; 

the last transport is scheduled for 

tomorrow. 

Challenging assembly 

At Greifswald, everything is prepared 

for installation of the high- performance 

components: In particular, 

the water pipes are installed in the 

plasma vessel, a total of 4.5 kilometres. 

“In the meantime, we have 

started laying the complex water 

pipes that bridge the last 40 centimetres 

between the vessel wall 

and the divertor plates,” explains 

assembly head Dr. Lutz Wegener. 

Later on, the plates must fit exactly 

to these connections. Although the 

extremely tricky work had previously 

been practised in a one-to-one 

model – “virtually a double assembly,” 

says Dr. Wegener – there are surprises 

when installing the 240 fitting 

pipes. The great tightness between 

the components makes welding 

a challenge, for which a special 

precision technique is used anyway. 

Often new designs and new manufacturing 

are necessary. In the 

narrow space also many screws are 

News


Operating Results January 2020 

Plant name Country Nominal 

capacity 

Type 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 

Energy generated, gross 

[MWh] 

Month Year Since 

commissioning 

Time availability 

[%] 

Energy availability 

[%] *) Energy utilisation 

[%] *) 

Month Year Month Year Month Year 

295 

OL1 Olkiluoto BWR FI 910 880 744 690 053 690 053 270 155 522 100.00 100.00 100.00 100.00 100.81 100.81 

OL2 Olkiluoto BWR FI 910 880 744 689 116 689 116 260 053 202 100.00 100.00 99.93 99.93 100.68 100.68 

KCB Borssele PWR NL 512 484 744 380 431 380 431 168 361 865 99.58 99.58 99.57 99.57 100.16 100.16 

KKB 1 Beznau 7) PWR CH 380 365 744 286 929 286 929 130 595 749 100.00 100.00 100.00 100.00 101.56 101.56 

KKB 2 Beznau 7) PWR CH 380 365 744 285 111 285 111 137 581 894 100.00 100.00 100.00 100.00 100.94 100.94 

KKG Gösgen 7) PWR CH 1060 1010 744 792 734 792 734 322 908 969 100.00 100.00 99.80 99.80 100.52 100.52 

CNT-I Trillo PWR ES 1066 1003 744 786 250 786 250 256 534 276 100.00 100.00 100.00 100.00 98.69 98.69 

Dukovany B1 PWR CZ 500 473 744 374 399 374 399 116 258 583 100.00 100.00 100.00 100.00 100.64 100.64 



Dukovany B4 2) PWR CZ 500 473 0 0 0 110 706 957 0 0 0 0 0 0 

Temelin B1 4) PWR CZ 1080 1030 664 712 309 712 309 122 627 122 89.25 89.25 88.38 88.38 88.48 88.48 

Temelin B2 PWR CZ 1080 1030 744 815 233 815 233 118 297 851 100.00 100.00 100.00 100.00 101.27 101.27 

Doel 1 2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0 

Doel 2 2) PWR BE 454 433 0 0 0 136 335 470 0 0 0 0 0 0 

Doel 3 PWR BE 1056 1006 744 805 963 805 963 263 917 613 100.00 100.00 100.00 100.00 102.00 102.00 

Doel 4 PWR BE 1084 1033 744 817 807 817 807 270 456 082 100.00 100.00 100.00 100.00 99.92 99.92 

Tihange 1 2) PWR BE 1009 962 0 0 0 307 547 424 0 0 0 0 0 0 

Tihange 2 PWR BE 1055 1008 744 781 306 781 306 258 835 824 100.00 100.00 99.99 99.99 100.50 100.50 

Tihange 3 PWR BE 1089 1038 744 806 745 806 745 281 369 321 100.00 100.00 100.00 100.00 100.18 100.18 

NEWS 

Plant name 

Type 

Nominal 

capacity 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 

Energy generated, gross 

[MWh] 

Time availability 

[%] 

Energy availability 

[%] *) Energy utilisation 

[%] *) 

Month Year Since Month Year Month Year Month Year 

commissioning 

KBR Brokdorf DWR 1480 1410 744 947 448 947 448 361 668 470 100.00 100.00 94.14 94.14 85.65 85.65 

KKE Emsland DWR 1406 1335 744 1 046 514 1 046 514 358 646 715 100.00 100.00 100.00 100.00 100.13 100.13 

KWG Grohnde DWR 1430 1360 744 994 849 994 849 389 269 695 100.00 100.00 99.97 99.97 92.91 92.90 

KRB C Gundremmingen SWR 1344 1288 744 1 004 133 1 004 133 342 327 685 100.00 100.00 100.00 100.00 99.94 99.94 

KKI-2 Isar DWR 1485 1410 744 1 102 078 1 102 078 366 864 547 100.00 100.00 99.98 99.98 99.52 99.52 

GKN-II Neckarwestheim DWR 1400 1310 744 1 042 000 1 042 000 341 280 244 100.00 100.00 100.00 100.00 100.28 100.28 

difficult to access for tools and a 

solution has to be found on a case- bycase 

basis: “Welded or screwed – the 

connections should remain tight for 

the next twenty years”. 

Compared with these tasks, subsequent 

installation of the divertor 

plates should be easier. “We have 

already developed special tools for this 

purpose – for example to lift and move 

the 70-kilogram plates,” says Lutz 

Wegener. Even the kick plate, on which 

the technicians in the vessel walk over 

the sensitive divertor and wall protection 

tiles, was a separate development 

project: it had to guarantee safe standing 

in a very confined space and be 

adapted to the unusual shape of the 

plasma vessel. On the other hand, it 

must not damage the wall structures or 

lead to any impurities that could later 

perturb the plasma. 

Plasma operation is expected to 

resume at the end of 2021. It is 

planned to begin with low water 

cooling, low heating power and short 

plasma pulses in order to allow testing 

of all installations in operation after 

the long break in experiments. With 

full cooling, longer pulses with plasma 

energies of up to one gigajoule should 

be possible – a target that will be 

slowly approached. Instead of the 

previous hundred-second pulses with 

heating powers of two megawatts and 

plasma energies of 200 megajoules, 

the cooled high-performance divertor 

should later allow pulses lasting up to 

30 minutes at full heating power. 

Wendelstein 7-X will then be able to 

demonstrate the essential advantage 

of stellarators, namely their ability to 

operate continuously. 

| www.ipp.mpg.de (201121501) 

Reactors 

Finland, Hanhikivi-1: 

Fennovoima announces 

details of progress with basic 

design review 

(fennovoima, nucnet) The company 

building the Hanhikivi-1 nuclear 

power plant in Finland has received 

222 documents from the Russian 

plant supplier, of which 134 have been 

conditionally accepted, 54 rejected 

and 34 remain under review. 

At the end of 2019, Fennovoima 

was still waiting for the delivery from 

Raos Project Oy, a subsidiary of 

Russian state nuclear corporation 

Rosatom, of the basic design packages 

of the turbine island and buildings, of 

which the basic design documents for 

buildings are – except for the control 

room building – almost complete. 

*) 

Net-based values 

(Czech and Swiss 

nuclear power 

plants gross-based) 

1) 

Refueling 

2) 

Inspection 

3) 

Repair 

4) 

Stretch-outoperation 

5) 

Stretch-inoperation 

6) 

Hereof traction supply 

7) 

Incl. steam supply 

BWR: Boiling 

Water Reactor 

PWR: Pressurised 

Water Reactor 

Source: VGB 

News


296 

NEWS 

All approved design documentation 

is available to regulator Stuk, but 

does not require Stuk’s approval. The 

preliminary safety assessment, which 

is a condition for the construction 

licence, is a separate documentation 

package, and must be submitted to 

Stuk for approval. 

The documents are part of the basic 

design review for the plant, which 

takes place in two stages. In the first 

stage, Fennovoima evaluates the safety 

of the plant, availability and main tenance 

aspects. At this stage, Fennovoima 

only issues conditional approvals 

for the documentation, meaning there 

are no technical obstacles in the design 

documentation that would prevent its 

final approval at a later stage. 

Fennovoima did not say how many 

documents it is still waiting for from 

Raos Project Oy or when the preliminary 

safety assessment would be 

ready. Statistics in the company’s 2019 

annual report, published on 25 March, 

suggested that as of 15 January 2020 

almost 50 % of the documents had 

been submitted. 

Project engineering director Petri 

Jyrälä said once the first review stage is 

complete, the documents will already 

clearly determine what the physical 

plant will look like. “We do not expect 

to see any major modifications of the 

plant after that stage, and we can proceed 

to finalising the documentation,” 

he said. 

Preparatory construction work on 

the Hanhikivi headland has reached a 

point where “we are ready for the construction 

of the nuclear power plant as 

soon as the construction licence is 

granted”, the report said. However, 

before beginning the construction of 

the plant, some 700,000 cubic metres of 

rock must be extracted from the excavation 

pit and the levelling concrete for 

the plant foundation must be poured. 

The plant’s projected startup date 

has been pushed back to 2028, four 

years behind the original schedule 

and eight years later than the proposed 

start when Finland’s government 

approved the project in 2010. 

Fennovoima, a consortium of 

Finnish industrial and energy companies, 

had warned in 2017 of 

potential delays. The aim is to receive 

the construction licence and to start 

construction in 2021. 

Hanhkivi-1 will be a 1,200-MW 

VVER pressurised water reactor. The 

reference plant for the unit Leningrad 

2 in Sosnovy Bor, Russia. 

According to Fennovoima’s website, 

the total investment cost for 

Hanhikivi-1 will be between € 6.5 and 

€ 7 bn, which includes initial plant 

costs, financing and waste management. 

This estimate has remained the 

same since spring 2014, when the 

original investment decision was 

made, Fennovoima said. 

| www.fennovoima.fi (201121516) 

Company News 

Framatome earns high safety 

marks from US Nuclear 

Regulatory Commission 

(framatome) Framatome’s fuel manufacturing 

facility in Richland, 

Washington, received a positive report 

from the U.S. Nuclear Regulatory Commission 

(NRC) following its recent biennial 

license performance review 

(LPR). The NRC concluded that no program 

areas require improvement – an 

accomplishment the site has achieved 

for seven con secutive reviews. 

“We hold our manufacturing facilities 

around the world to the highest 

standards of excellence for safety, quality, 

performance and delivery,” said 

Lionel Gaiffe, senior executive vice 

president, Framatome Fuel Business 

Unit. “This outstanding report by the 

NRC is recognition of our commitment 

to continuous improvement.” 

The NRC review takes place every 

two years, and examines four major 

categories for fuel manufacturing: 

Safety Operations, Radiological Controls, 

Facility Support and Other 

Areas. This latest review confirmed 

that the Richland facility continues to 

conduct activities safely and securely, 

while protecting public health and 

the environment during the 2018-19 

review period. 

“Our workforce manufactures the 

most advanced nuclear fuel designs 

with an uncompromising focus on 

safety and operational excellence,” 

said Ron Land, Richland site manager 

at Framatome. “This review confirms 

our commitment to our customers and 

our community.” 

In 2019, Framatome’s Richland 

facility celebrated its 50th anniversary. 

After receiving the industry’s 

first 40-year nuclear fuel fabrication 

license renewal from the NRC in 

2009, Framatome’s Richland facility is 

licensed to operate to 2049. 

| www.framatome.com (201121444) 

Decommissioning of the GNS 

plant in Duisburg-Wanheim is 

completed 

(gns) On 31 March 2020, GNS 

Gesellschaft für Nuklear-Service mbH 

vacated its former premises in 

Duisburg- Wanheim, returned the 

buildings and the premises to the 

lessor and thus terminated its activities 

at the site after 35 years. Since 1985, 

GNS had been processing low to intermediate-level 

radioactive waste from 

the operation and decommissioning of 

German nuclear power plants in three 

rented halls of the former Thyssen precision 

forge and packing it for subsequent 

interim storage or final disposal. 

In the course of the decommissioning, 

all facilities and installations 

for waste treatment and packaging 

were completely removed by GNS and, 

with the involvement of independent 

experts, the freedom from contamination 

of the entire site was demonstrated 

to the supervisory authority. 

This was the prerequisite for GNS to 

return the radiation pro tection handling 

permit required for operation by 

then as early as mid-March. Thus, the 

site can be put to conventional use 

again in the future. 

The employees who were last 

employed at the Duisburg plant will in 

future be deployed at other GNS 

locations. 

Background 

In Duisburg-Wanheim, GNS has 

operated a facility for the packaging 

of low- to intermediate-level radioactive 

waste from the operation and 

decommissioning of German nuclear 

power plants since 1985. For this 

purpose, the waste was generally 

compacted, dried and packed in containers 

suitable for interim storage 

and final disposal. With the gradual 

shutdown of the German nuclear 

power plants, the amount of operational 

waste as processed at the 

Duisburg facility of GNS is decreasing. 

At the same time, new capacities for 

processing local decommissioning 

waste have been created at the 

power plant sites. Therefore, GNS already 

announced the decision to close 

the Duisburg plant in December 2013. 

| www.gns.de (201121405) 

ROSATOM presents 

new type of SMR 

(rosatom) ROSATOM participated in 

Africa Energy Indaba Forum, which 

was hosted in Cape Town, South 

Africa. Ryan Collyer, acting CEO of 

Rosatom Central and Southern Africa 

highlighted the global shift towards 

nuclear, not only in the energy sector 

but also to address a myriad of other 

issues. 

His speech was focused on the 

possible use of nuclear technologies 

News


Uranium 

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

140.00 

) 1 

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

140.00 

) 1 

120.00 

120.00 

297 

100.00 

100.00 

80.00 

80.00 

60.00 

40.00 

20.00 

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

60.00 

40.00 

20.00 

NEWS 

0.00 

1980 

1985 

1990 

1995 

2000 

2005 

2010 

2015 

2020 

Year 

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

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020 

* Actual nominal USD prices, not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020 

| Uranium spot market prices from 1980 to 2020 and from 2009 to 2020. The price range is shown. 

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

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

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

180.00 

26.00 

) 1 ) 1 

160.00 

140.00 

0.00 

24.00 

22.00 

20.00 

Jan. 2009 

Jan. 2010 

Jan. 2011 

Jan. 2012 

Jan. 2013 

Jan. 2014 

Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

Jan. 2021 

120.00 

18.00 

16.00 

100.00 

14.00 

80.00 

12.00 

10.00 

60.00 

8.00 

40.00 

6.00 

20.00 

4.00 

2.00 

0.00 

0.00 

Jan. 2009 

Jan. 2010 

Jan. 2011 

Jan. 2012 

Jan. 2013 

Jan. 2014 


Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

Jan. 2021 


Jan. 2009 

Jan. 2010 

Jan. 2011 

Jan. 2012 

Jan. 2013 

Jan. 2014 


Jan. 2015 

Jan. 2016 

Jan. 2017 

Jan. 2018 

Jan. 2019 

Jan. 2020 

Jan. 2021 


| Separative work and conversion market price ranges from 2009 to 2020. The price range is shown. 

)1 

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

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

Sources: Energy Intelligence, Nukem; Bilder/Figures: atw 2020 

for desalination purposes apart from 

heat and electricity supply and the 

latest developments of ROSATOM in 

the area of SMRs featuring RITM-200 

reactor technology. ROSATOM SMRs 

can be a good alternative to diesel 

generators providing reliable power 

supply and preventing harmful 

emissions at a competitive price. 

Speaking at Energy Indaba, Ryan 

Collyer put a special emphasis on 

ROSATOM’s current developments in 

the field of small modular reactors. In 

particular, he presented RITM-200, an 

advanced pressurized-water reactor 

that incorporates all the best features 

from its predecessors – ship reactors. 

R. Collyer added that the main advantages 

of RITM-200 reactor are costefficiency, 

small size and safety. 

RITM-200 is designed for nuclear 

icebreakers, land-based small NPPs, 

and floating nuclear power plants. 

He also pointed out that RITM-200 

is a reference reactor. ROSATOM has 

already constructed six RITM-200 

reactors by now. Two reactors onboard 

Arktika icebreaker have already 

attained criticality. 

The speaker also outlined the 

features of the floating nuclear power 

plant that was connected to the grid at 

the end of 2019 and started supplying 

electricity to the grid. At present, 

ROSATOM is working on the next 

generation of the offshore nuclear 

power plants – an optimized floating 

power unit (OFPU). 

“We are working hard to do our 

part in delivering the great stories 

from our industry, to highlight its true 

potential to become a catalyst for 

sustainable development in Africa. We 

all understand that nuclear will play a 

vital role in achieving the United 

Nations sustainability goals not only 

in Africa but across the globe,” noted 

Ryan Collyer. 

| www.rosatom.ru (201121448) 

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)]. 

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 to June 2019 

p Uranium: 23.90–29.10 



July to December 2019 

p Uranium: 24.50–26.25 



2020 

January 20202 

p Uranium: 24.10–24.90 



February 20202 

p Uranium: 24.25–25.00 



| Source: Energy Intelligence 

www.energyintel.com 

News


298 

NUCLEAR TODAY 

John Shepherd is a 

freelance journalist 

and communications 

consultant. 

Sources: 

NEI’s response 

to Covid-19 

https://bit.ly/2JEJiDL 

Rosatom 

announcement 

https://bit.ly/2JUscC1 

Dr Fatih Birol 

statement 

https://bit.ly/2ReYKup 

Energy Providers Deserve Our Gratitude 

Now More Than Ever 

What a strange and unnerving time we live in at the moment. As I write this article, tens of thousands of people have 

lost their lives as Covid-19 sweeps across the world. 

According to the director-general of the World Health 

Organization, Dr Tedros Adhanom Ghebreyesus, world 

leaders are confronting “the defining health crisis of our 

time… at war with a virus that threatens to tear us apart – 

if we let it”. 

The world is indeed at war with a common enemy and 

the ‘soldiers’ confronting the virus on the front line are 

undoubtedly health service workers. 

Tributes have been paid to healthcare professionals in 

many countries by members of the general public who 

have emerged from lockdowns and isolation only to show 

their appreciation by applauding, singing and waving 

national flags from their windows and balconies. 

Beyond the medics, there are many others who are 

rightly identified as essential workers, those whose day- today 

jobs in supporting services and infrastructure take on 

far greater significance at this unsettling time. 

I suspect many of us now depend more than ever on our 

internet-connected devices to work from home, to stay in 

contact with family and friends, or to order shopping and 

to pass the time with films and games. And for that, we 

owe a debt of gratitude to some of the unsung essential 

workers of this crisis – the energy sector employees who 

ensure electricity continues to reach our homes, hospitals 

and other services. 

Electricity, whether derived from nuclear, fossil fuels, 

wind or solar, is always taken for granted in the developed 

world, but it should not be so during the Covid-19 pandemic. 

The multifaceted benefits of peaceful nuclear power activities 

deserve particular recognition and praise at this time. 

For example, the International Atomic Energy Agency 

(IAEA) is currently part of the United Nations’ Crisis 

Management Team on Covid-19. The agency recently 

announced that it had dispatched the first batch of 

equipment to more than 40 countries to enable them to use 

a nuclear-derived technique to rapidly detect the 

coronavirus that causes Covid-19. 

The IAEA said dozens of laboratories in Africa, Asia, 

Europe, Latin America and the Caribbean will receive 

diagnostic machines and kits, reagents and laboratory 

consumables to speed up national testing, which is crucial in 

containing the outbreak. They will also receive biosafety 

supplies, such as personal protection equipment and 

laboratory cabinets for the safe analysis of collected samples. 

The first batch of supplies, worth around €4 million, 

will help countries use the technique known as real time 

reverse transcription–polymerase chain reaction (real 

time RT-PCR). The IAEA said that this is the most sensitive 

technique for detecting viruses currently available. 

The nuclear-derived DNA amplification method 

originally used radioactive isotope markers to detect 

genetic material from a virus in a sample, the IAEA said. 

Subsequent refining of the technique has led to the more 

common use today of fluorescent markers instead. 

Meanwhile, the nuclear industry continues to keep the 

power flowing to essential services while also increasing 

already stringent safeguards for employees. The Nuclear 

Energy Institute (NEI) said measures in the US included 

setting up screening points before people can enter nuclear 

plants, to identify those who have symptoms. 

The NEI said it was committed to maintaining “safe, 

reliable operations in times of challenging national 

circumstances”. “Our industry has had pandemic guidelines 

since 2006, and these were updated early this year,” 

the NEI said. “They include keeping masks on hand and 

exercising social distancing.” 

In the UK, the Nuclear Industry Association said there 

was a “strong focus on protecting personnel by keeping 

only essential workforce on-site and adapting working 

practices to make sure social distancing is possible – 

including the approach to worker transport and 

accommodation and increased temperature monitoring at 

construction sites like Hinkley Point C”. 

The UK’s Nuclear Advanced Manufacturing Research 

Centre is also supporting a national effort to step up 

production of vital medical equipment such as ventilators. 

In common with other nuclear sites worldwide, Sellafield 

Ltd in the UK said it had donated items including disposable 

respirators and protective clothing to health care workers. 

In Russia, a subsidiary of Rosatom’s nuclear fuel company 

TVEL, Rusatom-Additive Technologies, said it had started 

producing prototypes and was set to start 3D printing valves 

for Venturi oxygen masks – a component of ventilators. 

Rosatom said the need for the valves had increased 

substantially as a result of the pandemic. Rosatom said 

production facilities had the capacity to produce about 

300 valves per week using a biocompatible polymer “that 

does not require additional processing”. 

The China National Nuclear Corporation announced 

that it was sending tonnes of personal protective equipment 

to a number of countries to support the fight against the 

virus. 

Covid-19 has focused minds as never before in modern 

times on the importance of maintaining and expanding 

critical power systems in the developed world, while 

helping less-developed regions acquire the infrastructure 

needed to connect to electricity grids. 

The executive director of the International Energy 

Agency (IEA), Dr Fatih Birol, has said in response to the 

pandemic that baseload electricity generating capacity 

such as that provided by nuclear is “a crucial element in 

ensuring a secure electricity supply”. 

Birol has urged policymakers to already be thinking 

and preparing for beyond the crisis and to “design markets 

that reward different sources for their contributions to 

electricity security, which can enable them to establish 

viable business models”. 

The IEA chief’s analysis is spot on. Covid-19 can and 

will be beaten, but its legacy should include a commitment 

to ensure that more and not less is invested in clean 

nuclear-generated electricity systems to help protect and 

prepare the world for whatever future challenges we may 

face. 

Author 

John Shepherd 

Nuclear Today 

Energy Providers Deserve Our Gratitude Now More Than Ever ı John Shepherd

Kommunikation und 

Training für Kerntechnik 

Rückbau kerntechnischer Anlagen 

In Kooperation mit 

TÜV SÜD Energietechnik GmbH 

Baden-Württemberg 

Seminar: 

Stilllegung und Rückbau in Recht und Praxis 

Seminarinhalte 

Genehmigungen für die Stilllegung und den Rückbau 

ı 

ı 

Gestaltung der Übergänge zwischen der Betriebs- und der Stilllegungsgenehmigung 

Gestaltung der Übergänge zwischen Genehmigungs- und Aufsichtsphase 

Rechtliche Gestaltung des Genehmigungsverfahrens 

ı 

ı 

ı 

ı 

Elemente des Umweltbereiches 

UVP und Erörterungstermin 

Anfechtung von Genehmigungen 

Stellungnahmen der EU-Kommission 

Reststoffe und Abfälle 

ı 

ı 

ı 

ı 

Vorstellung von Reststoffkonzepten 

Bewertung der Reststoffkonzepte 

Neue Regelungen zum Übergang der Entsorgungsverantwortung 

Entsorgungsfragen rund um Abfälle aus dem Rückbau 

Zielgruppe 

Die 2-tägige Schulung wendet sich an Fach- und Führungskräfte, Mitarbeiterinnen und Mitarbeiter 

von Betreibern, Industrie und Dienstleistern, die sich mit der Thematik aktuell bereits beschäftigen 

oder sich künftig damit auseinander setzen werden. 

Maximale Teilnehmerzahl: 12 Personen 

Referenten 

Dr. Stefan Kirsch 

Dr. Christian Raetzke 

Wir freuen uns auf Ihre Teilnahme! 

ı Abteilung Stilllegung, Entsorgung, Reaktorphysik, TÜV SÜD Energietechnik 

GmbH Baden-Württemberg 

ı Rechtsanwalt, Leipzig 

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

Termin 

2 Tage 

24. bis 25. September 2020 

Tag 1: 11:00 bis 18:00 Uhr 

Tag 2: 09:00 bis 16:30 Uhr 

Veranstaltungsort 

Kaiserin Friedrich-Haus 


10115 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 


10115 Berlin 

Petra Dinter-Tumtzak 

Fon +49 30 498555-30 

Fax +49 30 498555-18 

seminare@KernD.de

#51KT 

www.kerntechnik.com 

Medienpartner 

ABSAGE: KERNTECHNIK 2020 

Auf Grund der aktuellen Entwicklungen der Corona-Pandemie 

ist die Jahrestagung KERNTECHNIK 2020 in Berlin abgesagt. 

Bitte informieren Sie sich auch auf unserer Website: www.kerntechnik.com 

51. KERNTECHNIK 

2020 

5. – 6. Mai 2020 

Estrel Convention Center 

Berlin

atw - International Journal for Nuclear Power | 05.2020

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