atw - International Journal for Nuclear Power | 05.2020

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

www.nucmag.com

nucmag.com

2020

5

ISSN · 1431-5254

24.– €

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

GNS Gesellschaft für Nuklear-Service mbH

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


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.


atw Vol. 65 (2020) | Issue 5 ı May

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

RA Dr. Christian Raetzke

Dr. Stefan Kirsch

RA Dr. Christian Raetzke

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.


atw Vol. 65 (2020) | Issue 5 ı May

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


atw Vol. 65 (2020) | Issue 5 ı May

247

Feature

Research and Innovation

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

Research and Innovation

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

Environment and Safety

276 Physical and Chemical Effects of Containment Debris

on the Emergency Coolant Recirculation

Jisu Kim and Jong Woon Park

Decommissioning and Waste Management

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


atw Vol. 65 (2020) | Issue 5 ı May

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

Inside Nuclear with NucNet

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


atw Vol. 65 (2020) | Issue 5 ı May

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

Inside Nuclear with NucNet

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


atw Vol. 65 (2020) | Issue 5 ı May

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


atw Vol. 65 (2020) | Issue 5 ı May

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


atw Vol. 65 (2020) | Issue 5 ı May

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

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


atw Vol. 65 (2020) | Issue 5 ı May

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 | RESEARCH AND INNOVATION 253

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


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FEATURE | RESEARCH AND INNOVATION 254

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

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


atw Vol. 65 (2020) | Issue 5 ı May

| 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.

FEATURE | RESEARCH AND INNOVATION 255

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

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


atw Vol. 65 (2020) | Issue 5 ı May

FEATURE | RESEARCH AND INNOVATION 256

| 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

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


atw Vol. 65 (2020) | Issue 5 ı May

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...?


atw Vol. 65 (2020) | Issue 5 ı May

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


atw Vol. 65 (2020) | Issue 5 ı May

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.

Research and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

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 and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

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

RESEARCH AND INNOVATION 261

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.

Research and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

RESEARCH AND INNOVATION 262

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.

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


atw Vol. 65 (2020) | Issue 5 ı May

| 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

RESEARCH AND INNOVATION 263

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


atw Vol. 65 (2020) | Issue 5 ı May

RESEARCH AND INNOVATION 264

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.

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


atw Vol. 65 (2020) | Issue 5 ı May

| 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.

RESEARCH AND INNOVATION 265

Research and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

RESEARCH AND INNOVATION 266

| 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.

Research and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

| 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


atw Vol. 65 (2020) | Issue 5 ı May

RESEARCH AND INNOVATION 268

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.

1 Introduction

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


atw Vol. 65 (2020) | Issue 5 ı May

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

RESEARCH AND INNOVATION 269

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atw Vol. 65 (2020) | Issue 5 ı May

RESEARCH AND INNOVATION 270

| 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.

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atw Vol. 65 (2020) | Issue 5 ı May

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.

RESEARCH AND INNOVATION 271

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atw Vol. 65 (2020) | Issue 5 ı May

RESEARCH AND INNOVATION 272

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.

Research and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

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.

RESEARCH AND INNOVATION 273

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.

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


atw Vol. 65 (2020) | Issue 5 ı May

RESEARCH AND INNOVATION 274

| 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

Goodness of fit:

SSE: 156.3

R-square: 0.9708

Adjusted R-square: 0.9683

RMSE: 1.642

Goodness of fit:

SSE: 3139

R-square: 0.9134

Adjusted R-square: 0.9059

RMSE: 7.357

Goodness of fit:

SSE: 342.5

R-square: 0.9832

Adjusted R-square: 0.9818

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

Research and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

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

RESEARCH AND INNOVATION 275

Research and Innovation

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


atw Vol. 65 (2020) | Issue 5 ı May

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.

1 Introduction

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.

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

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).

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

ENVIRONMENT AND SAFETY 278

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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Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

| 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

ENVIRONMENT AND SAFETY 279

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

ENVIRONMENT AND SAFETY 280

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.

1 Introduction

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

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

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

0.40 MPa 109 kW 0.045 MPa Ambient temperature

0.45 MPa 109 kW 0.045 MPa Ambient temperature

0.52 MPa 109 kW 0.045 MPa Ambient temperature

1) Unless otherwise

specified, pressure in

this paper refers to

absolute pressure.

ENVIRONMENT AND SAFETY 281

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

ENVIRONMENT AND SAFETY 282

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

0.045 MPa 0.30 MPa ~ 0.52MPa Saturated temperature

0.065 MPa 0.30 MPa ~ 0.52MPa Saturated temperature

0.100 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

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

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

0.100 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).

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

(a) Experimental results

| Fig. 4.

Containment pressure in start-up conditions.

(b) GOTHIC simulation results

ENVIRONMENT AND SAFETY 283

(a) Experiment results

| Fig. 5.

Flow rate of the loop in start-up conditions.

(b) GOTHIC simulation results

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

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

ENVIRONMENT AND SAFETY 284

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.

(a) Experiment results

(b) GOTHIC simulation results

(a) Experiment results

| Fig. 7.

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

(b) GOTHIC simulation results

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

(a) Experiment results

| Fig. 8.

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

(b) GOTHIC simulation results

ENVIRONMENT AND SAFETY 285

(a) Experiment results

| Fig. 9.

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

(b) GOTHIC simulation results

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.

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

ENVIRONMENT AND SAFETY 286

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

Environment and Safety

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


atw Vol. 65 (2020) | Issue 5 ı May

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

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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

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atw Vol. 65 (2020) | Issue 5 ı May

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].

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DECOMMISSIONING AND WASTE MANAGEMENT 290

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.

Decommissioning and Waste Management

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atw Vol. 65 (2020) | Issue 5 ı May

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

9. International Organization for Standardization ISO 13849

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

12. International Organization for Standardization ISO 10218

“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

DECOMMISSIONING AND WASTE MANAGEMENT 291

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


atw Vol. 65 (2020) | Issue 5 ı May

292

Inside

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Sie dies bitte der KTG-

Geschäftsstelle mit.

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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)

Robert-Koch-Platz 4

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

Geburtstagsnennungen:

März 2020

80 Jahre | 1940

7. Dr. Volker Klix, Gehrden

Mai 2020

79 Jahre | 1941

16. Dr. Jürgen Baier, Höchberg

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


atw Vol. 65 (2020) | Issue 5 ı May

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


atw Vol. 65 (2020) | Issue 5 ı May

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


atw Vol. 65 (2020) | Issue 5 ı May

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 B2 PWR CZ 500 473 744 371 137 371 137 111 414 455 100.00 100.00 99.81 99.81 99.77 99.77

Dukovany B3 PWR CZ 500 473 580 284 866 284 866 110 536 602 77.96 77.96 76.86 76.86 76.58 76.58

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


atw Vol. 65 (2020) | Issue 5 ı May

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


atw Vol. 65 (2020) | Issue 5 ı May

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

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

Jan. 2015

Jan. 2016

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020

Jan. 2021

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

Jan. 2009

Jan. 2010

Jan. 2011

Jan. 2012

Jan. 2013

Jan. 2014

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

Jan. 2015

Jan. 2016

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020

Jan. 2021

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

| 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

p Conversion: 6.00–14.50

p Separative work: 34.00–42.00

2019

January to June 2019

p Uranium: 23.90–29.10

p Conversion: 13.50–18.00

p Separative work: 41.00–49.00

July to December 2019

p Uranium: 24.50–26.25

p Conversion: 18.00–23.00

p Separative work: 47.00–52.00

2020

January 20202

p Uranium: 24.10–24.90

p Conversion: 22.00–23.00

p Separative work: 48.00–51.00

February 20202

p Uranium: 24.25–25.00

p Conversion: 22.00–23.00

p Separative work: 45.00–53.00

| Source: Energy Intelligence

www.energyintel.com

News


atw Vol. 65 (2020) | Issue 5 ı May

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


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