atw 2018-05v6

inforum

nucmag.com

2018

5

285

Security Controls

for Nuclear Safety

299 ı Operation and New Build

Safety Enhancement in Operation of Czech VVER Units

305 ı Operation and New Build

Underwater-Robotics in Nuclear Power Plants

ISSN · 1431-5254

24.– €

312 ı Decommissioning and Waste Management

The New CASTOR® geo

354 ı Nuclear Today

Nuclear Newcomer Turkey and Japan

Show the Way Ahead

AMNT

Edition

Join us in Berlin

29–30 May


atw Vol. 63 (2018) | Issue 5 ı May

Welcome Addresses for the 49 th Annual

Meeting on Nuclear Technology (AMNT 2018)

29 –30 May 2018, Berlin

KTG (German Nuclear Society)

The Chairman

On behalf of the Kerntechnische Gesell schaft e.V.

I would like to warmly welcome you to the 49 th

Annual Meeting on Nuclear Technology. The members

of the KTG once again prepared an excellent

and highly topical programme and special thanks

should be expressed to the members of the Programme

Committee, who contributed personally

and with great engagement in the preparation of this conference.

We also like to thank all our speakers and exhibitors, who formed our

AMNT 2018 as well into a magnet for a national and international

audience.

In the past few weeks important settings, with regards to the final

stage of the life cycle of German nuclear power plants – decommissioning

and disposal of radioactive waste – were made:

• The in Germany largely and politically imposed decommissioning

continues: in February this year Unterweser as 5 th plant in the

“Post-Fukushima era” and Grafenrheinfeld as 6 th plant just a

couple of days ago received the first decommissioning and

dismantling license;

• In many decommissioned plants spent fuel elements were and

still are loaded from the spent fuel pools into Castor® casks and

transferred to the on-site storage facilities.

For us, as nuclear technicians, it is also worth noticing:

• That the German jurisdiction „ideology-free“ e.g. passed judgement

on the “Enforcement of Castor®­ Transportation on the River

Neckar“ or the “Illegality of the Fuel Element Tax“.

• That in our neighbour country Switzerland the power plant

Beznau’s unit 1, after a three-year-interruption of operation, may

start operating again after the operator Axpo was able to prove,

that the reactor pressure vessel does not contain any flaws that

could lead to a failure. Swiss experts and authorities, ideology-free

and supported by the opinion of international experts, verified

and accepted the technical evidence.

• That german nuclear power plants, which are connected to the

grid, generate after all in secured power operation environmentally

friendly electricity, a responsibility, which they will meet

until the last day of the year 2022 and beyond.

During the different formats of our annual meeting we will reflect on

these and further focal points and place emphasis on the professional

dialogue between our national and international knowledge carriers.

This year’s AMNT will also continue with successful formats of young

talents such as Nuclear Energy Campus or Young Scientists’ Workshop.

I am particularly pleased that this year we will not only award the

Karl-Wirtz Prize to one of our young scientists but additionally also

award the honorary membership of the KTG for long term commitment

and merits around nuclear technology, “Innovation Made in

Germany“ and supporting youngsters.

For our annual meeting, I wish all participants’ new insights as

well as interesting meetings, contacts and exchange. Once more also

our very impressive industrial exhibition provides a suitable platform

for this purpose.

And please do not forget: what unites the individual members of

the KTG is the “fascination of nuclear energy”. Get yourself (once

again) infected by it…

DAtF (German Atomic Forum)

The President

It is that time of the year again: Our joint annual

meeting from the DAtF and KTG, the Annual

Meeting on Nuclear Technology is approaching.

On the 29 and 30 May the 49 th edition of our

industry’s key meeting in Germany and one of

the most recognised nuclear expert conference

worldwide takes place. Already today I would

like to bring you attention to the very considerable anniversary we

will celebrate next year: our 50 th Annual Meeting. Please already

save the date today for the 7 and 8 May 2019. Many attractive

additional programme elements will await you!

However, first of all – as usual – an outlook on our once more

highly topical and professionally excellent programme of this year’s

meeting. The list of our speakers within our plenary programme

promises a maximum of interesting and entertaining lectures:

Karsten Möhring MdB (CDU), Dr. Dr. Jan Backmann (Nuclear

Regulatory Schleswig-Holstein), David Peattie (NDA), Dr. Willibald

Kohlpaintner ( Axpo), Dr. Dirk Stenkamp (TÜV Nord Group), Carsten

Haferkamp (Framatome), Jacopo Buongiorno (MIT), Ursula

Heinen-Esser (BGE), Dr. Maria J. Rodriguez (MPI).

Also, outside the plenary session exciting expert lectures are

awaiting you, for example on the implementation of the new EU BSS

in Germany or on the role of radiation protection during the

dismantling, release and disposal. The paradigm shift in nuclear

waste management in Germany and the skills and recourses needed

for safe decommissioning will be discussed. It will be spoken about

the safety of advanced reactor concepts and practical protective

measures in nuclear power plants as well as about aspects of the

safety culture. In addition to many other topics, research projects

such as DEMO or NUGENIA will also be reported. Last but not least

– the Young Scientists’ Workshop demonstrates the strong engagement

of the younger generation and the support by the involved

institutions.

An extraordinary exhibition of companies and organizations

awaits you in the industrial exhibition. While we are very excited to

welcome again country pavilions from the United Kingdom and the

Czech Republic also many new national and international exhibitors

announced their participation at the AMNT. Use the breaks to give

you an idea of the development of nuclear products and services.

And above all: the industrial exhibition offers in this spectrum the

unique opportunity in Germany to establish or deepen contacts

with high-level representatives from politics, public authorities, the

industry, operators, expert organisations and research institutes.

We are very much looking forward to your registration at

www.nucleartech-meeting.com if you are not a registered user yet

to the AMNT 2018.

My especially warm thanks go to all persons involved in the programme

design, speakers, participants, exhibitors and sponsors.

Your involvement and your professional expertise are the decisive

factors for the success of our annual meeting.

Dr. Ralf Güldner

WELCOME TO AMNT 2018 275

Frank Apel

Editorial

Welcome Addresses for the 49 th Annual Meeting on Nuclear Technology (AMNT 2018)


atw Vol. 63 (2018) | Issue 5 ı May

276

Grußworte zum 49. Annual Meeting on

Nuclear Technology (AMNT 2018)

WELCOME TO AMNT 2018

29. und 30. Mai 2018, Berlin

KTG (Kerntechnische Gesellschaft e.V.)

Der Vorsitzende

Zur 49. Jahrestagung Kerntechnik möchte ich Sie

im Namen der Kerntechnischen Gesellschaft e.V.

herzlich willkommen heißen. Die Mitglieder

der KTG haben erneut ein exzellentes und

hoch aktuelles Programm mit vorbereitet, und

besonderer Dank gebührt den Mitgliedern des

Programm­ Ausschusses, die sich persönlich mit

großem Engagement in die Vorbereitung dieser

Konferenz eingebracht haben. Unser Dank geht auch an alle Referenten

und Aussteller, die auch unsere 49. Jahrestagung Kerntechnik zu

einem Magnet für nationales und internationales Fachpublikum

geformt haben.

In den vergangenen Wochen erfolgten hinsichtlich der abschließenden

Lebenszyklen der deutschen Kernkraftwerke – dem Rückbau und

der Entsorgung radioaktiver Abfälle – wichtige Weichenstellungen:

• der in Deutschland weitestgehend politisch verordnete Rückbau

geht voran: im Februar dieses Jahres haben Unterweser als

5. Anlage in der „Post-Fukushima-Ära“ und vor wenigen Tagen

Grafenrheinfeld als 6. Anlage die 1. Stilllegungs- und Abbaugenehmigung

erhalten;

• in vielen stillgelegten Anlagen wurden und werden die abgebrannten

Brennelemente aus den Brennelement-Lagerbecken in

Castoren geladen und in die standortnahen Zwischenlager

verbracht.

Ferner ist für uns Kerntechniker erwähnenswert, dass:

• im letzten Jahr die deutsche Rechtsprechung „ideologiefrei“ z. B.

Urteile zur „Durchsetzung von Castor®-Transporten auf dem

Neckar“ oder die „Nichtrechtmäßigkeit der Brennelement-Steuer“

gefällt hat,

• in unserem Nachbarland Schweiz das Kraftwerk Beznau den

Block 1 nach einer dreijährigen Betriebsunterbrechung wieder

den Betrieb aufnehmen kann, nachdem der Betreiber Axpo nachweisen

konnte, dass die Einschlüsse im Stahl des Reaktordruckbehälters

keinen negativen Einfluss auf die Sicherheit haben. Die

Schweizer Gutachter und Behörden haben – ideologiefrei und

gestützt auf die Meinung internationaler Experten – die entsprechenden

technischen Nachweise geprüft und akzeptiert.

• die am Netz befindlichen deutschen Kernkraftwerke nach wie

vor im sicheren Leistungsbetrieb umweltfreundlichen Strom

erzeugen, eine Verantwortung, der sie bis zum letzten Tag des

Jahres 2022 und darüber hinaus nachkommen werden.

Diese und weitere Schwerpunkte werden wir in den unterschiedlichen

Formaten unserer Jahrestagung reflektieren und setzen dabei

wieder auf den fachlichen Dialog zwischen nationalen und internationalen

Wissensträgern. Auch unser diesjähriges AMNT wird

erfolgreiche Formate der Nachwuchsarbeit wie Nuclear Energy

Campus oder Young Scientists‘ Workshop fortführen.

Besonders freue ich mich, dass wir in diesem Jahr nicht nur den

Karl-Wirtz-Preis an einen Nachwuchswissenschaftler verleihen

werden, sondern auch noch die Ehrenmitgliedschaft der KTG für

langjähriges Engagement und Verdienste rund um Kerntechnik,

„ Innovation Made in Germany“ und Nachwuchsförderung.

Für unsere Jahrestagung wünsche ich allen Teilnehmern neue

Erkenntnisse, interessante Begegnungen, Kontakte und Gespräche.

Auch unsere erneut sehr eindrucksvolle Industrieausstellung bietet

dafür eine geeignete Plattform.

Und vergessen Sie bitte nicht: was die Mitglieder der KTG verbindet,

ist die „Faszination Kerntechnik“. Lassen auch Sie sich davon

(erneut) anstecken …

DAtF (Deutsches Atomforum e.V.)

Der Präsident

Es ist wieder soweit. Unsere gemeinsame

Jahrestagung von DAtF und KTG, das Annual

Meeting on Nuclear Technology steht vor der

Tür. Am 29. und 30. Mai findet bereits die

49. Auflage des zentralen Treffens unserer

Branche in Deutschland und eine der anerkanntesten

kerntechnischen Fachtagungen

weltweit statt. Bereits heute möchte ich darauf

hinweisen, dass wir im nächsten Jahr ein beachtliches Jubiläum

feiern werden: unsere 50. Jahrestagung. Bitte merken Sie sich schon

heute das Datum, den 7. und 8. Mai 2019, vor. Viele attraktive

zusätzliche Programm-Elemente werden Sie erwarten.

Zunächst aber – wie gewohnt – ein Ausblick auf das hochaktuelle

und fachlich exzellente Programm unserer diesjährigen Tagung.

Die Rednerliste im Plenarprogramm verspricht ein Höchstmaß an

interessanten und kurzweiligen Vorträgen: Karsten Möhring MdB

(CDU), Dr. Dr. Jan Backmann (Atomaufsicht Schleswig-Holstein),

David Peattie (NDA), Dr. Willibald Kohlpaintner (Axpo), Dr. Dirk

Stenkamp (TÜV Nord Group), Carsten Haferkamp (Framatome),

Jacopo Buongiorno (MIT), Ursula Heinen-Esser (BGE), Dr. Maria J.

Rodriguez (MPI).

Auch außerhalb der Plenarsitzung erwarten Sie spannende

Fachvorträge, so zum Beispiel zu Implementierung der neuen

Strahlenschutzgrundnormen der EU oder etwa zur Rolle des

Strahlen schutzes beim Rückbau, Freigabe und Entsorgung. Es wird

über den Paradigmenwechsel in der Entsorgung schwach- und

mittelradioaktiver Abfälle diskutiert, und die benötigten

Kompetenzen und Ressourcen für einen sicheren Rückbau werden

erörtert. Die Sicherheit bei fortschrittlichen Reaktorkonzepten

sowie praktische Schutzmaßnahmen in Kernkraftwerken werden

ebenso thematisiert wie Aspekte der Sicherheitskultur. Neben

vielen anderen Themen wird auch über Forschungsprojekte wie

DEMO oder NUGENIA berichtet. Nicht zu letzt demonstriert der

Young Scientists’ Workshop großes Engagement junger Wissenschaftler

in unserer Branche sowie die Unterstützung der beteiligten

Institutionen.

Eine außergewöhnliche Leistungsschau der Unternehmen und

Organisationen wartet in der Industrieausstellung auf Sie. Während

wir uns sehr freuen, erneut Länderpavillons aus dem Vereinigten

Königreich und Tschechien begrüßen zu dürfen, haben sich für das

AMNT auch viele neue nationale und internationale Aussteller

angekündigt. Es wird also im besten Sinne „voll“ in unserer

Ausstellung. Nutzen Sie die Pausen, um sich über die Entwicklung

der kerntechnischen Produkte und Dienstleistungen selbst ein Bild

zu machen. Und vor allem: Die Industrie-Ausstellung bietet die in

Deutschland in dieser Bandbreite einzigartige Gelegenheit, Kontakte

mit hochrangigen Vertretern aus Politik, Behörde, Industrie,

Betreibern, Gutachtern und Forschungseinrichtungen zu knüpfen

oder zu vertiefen.

Sofern Sie sich noch nicht als Teilnehmer zum AMNT 2018

registriert haben, freuen wir uns über Ihre Anmeldung unter

www.nucleartech-meeting.com.

Allen an der Programmgestaltung Beteiligten, allen Rednern,

Teilnehmern, Ausstellern und Sponsoren danke ich ganz herzlich.

Ihr Engagement und Ihre fachliche Expertise sind die ent scheidenden

Faktoren für den Erfolg unserer Jahrestagung.

Dr. Ralf Güldner

Frank Apel

Editorial

Welcome Addresses for the 49 th Annual Meeting on Nuclear Technology (AMNT 2018)


Unsere Jahrestagung – die gemeinsame Fachkonferenz von KTG und DAtF

29. und 30. Mai 2018 | Estrel Convention Center Berlin ı Deutschland

Kompetenz

& Innovation

Unsere Austeller, Sponsoren

und Medien-Partner

Sicherheitsstandards &

Betriebsexzellenz

3 Silber Sponsor

Rückbauerfahrung &

Entsorgungslösungen

Registrieren Sie sich für die

Jahrestagung Kerntechnik unter:

http://www.nucleartech-meeting.com/

registration/online-registration.html

3 Tschechischer Pavillon

3 Medien-Partner

3 Britischer Pavillon

Department for

International Trade

Weitere Unternehmen angekündigt.

und ANTECH, Arup, Jacobs, National Nuclear Laboratory,

Porvair Filtration Group, Radiation Protection Advice Ltd

www.unserejahrestagung.de

Unsere Jahrestagung Kerntechnik – das Original seit fast 50 Jahren. Hier trifft sich die Branche.


atw Vol. 63 (2018) | Issue 5 ı May

278

Issue 5

May

CONTENTS

285

Security Controls

for Nuclear Safety

| | View of the Isar nuclear power plant site. On the left the Isar-2 plant, a 1,485 MW (gross electricity) pressurised water reactor in

commercial operation since 1988. The plant took the first place in gross electricity production worldwide “Top Ten” in 1994, 1999, 2000

to 2004, 2006, 2011 and 2013 and generated about 345 TWh (345 * 109 kWh), adequate to power Germanys electricity consumers for

7 months. On the right the 912 MW (gross) Isar-1 boiling water reactor. The plant is under decommissioning and has stopped electricity

production in 2011 respectively to the revision of the German Atomic Energy Act after the Fukushima accident in Japan

(Courtesy: PreussenElektra)

Welcome To AMNT 2018

Welcome Addresses for the 49 th Annual Meeting

on Nuclear Technology (AMNT 2018) 275

Grußworte zum 49. Annual Meeting on Nuclear

Technology (AMNT 2018) 276

Abstracts | English 280

Abstracts | German 281

Inside Nuclear with NucNet

Yucca Mountain: Can the US Finally

End the $12 Billion Impasse? 282

NucNet

DAtF Notes. . . . . . . . . . . . . . . . . . . . . .283

Calendar . . . . . . . . . . . . . . . . . . . . . . . 284

Environment and Safety

Detective Application Security Controls

for Nuclear Safety 285

Deeksha Gupta, Karl Waedt and Yuan Gao

Energy Policy, Economy and Law

General Data Protection Regulation (GDPR)

of the European Union – What Had to Be

Considered until 25 May 2018 289

EU-Datenschutzgrundverordnung – Was bis

zum 25.5.2018 beachtet sein muss(te) 289

Stefan Loubichi

285

| | Secure Centralized Logging via Data Diode.

Spotlight on Nuclear Law

The New Radiation Protection Law and the

Approval: May Makes Everything New? 296

Das neue Strahlenschutzrecht und die Freigabe:

Alles neu macht der Mai? 296

Ulrike Feldmann

Contents


atw Vol. 63 (2018) | Issue 5 ı May

Operation and New Build

Continuous Process of Safety Enhancement

in Operation of Czech VVER Units 299

The New CASTOR® geo – A Compre hensive

Solution For Transport and Storage of

Spent Nuclear Fuel, MOX and Damaged Fuel 312

279

J. Duspiva, E. Hofmann, J. Holy, P. Kral and M. Patrik

Linus Bettermann and Roland Hüggenberg

Optimal Holistic Disposal Planning

– Development of a Calculation Tool – 316

Johannes Schubert, Anton Philipp Anthofer and Max Schreier

299

CONTENTS

| | Nodalization scheme of VVER-1000 for RELAP5.

316

Applications of Underwater-Robotics

in Nuclear Power Plants 305

Gunnar Fenzel, Dr. Dietmar Nieder and Alexandra Sykora

| | Workflow of calculation.

Scope for Thermal Dimensioning

of Disposal Facilities for High-level

Radioactive Waste and Spent Fuel 319

Joachim Heierli, Helmut Hirsch, Bruno Baltes

|305

312

| | GUI with integrated simulation surroundings.

360 Degree Area Atlas

in the Biblis Nuclear Power Plant 308

360° Raumatlas im Kraftwerk Biblis 308

Jürgen Kircher

Decommissioning and Waste Management

ELINDER – European Learning Initiatives

for Nuclear Decommissioning and

Environmental Remediation 309

Pierre Kockerols, Hans Günther Schneider and Daniela Santopolo

| A CASTOR® geo24B.

Research and Innovation

Heavy Ions Irradiation as a Tool

to Minimize the Number of In-Pile Tests

in UMo Fuel Development 325

H. Breitkreutz, J. Shi, R. Jungwirth, T. Zweifel, H.-Y. Chiang and W. Petry

325

| | Irradiation set-up for heavy ions.

KTG Inside . . . . . . . . . . . . . . . . . . . . . . 331

News . . . . . . . . . . . . . . . . . . . . . . . . . 333

Report

Operating results 2017 – Part I* 336

Nuclear Today

Nuclear Newcomer Turkey and

‘Comeback Kid’ Japan Show the Way Ahead 354

Imprint 332

AMNT 2018: Registration Form . . . . . . . . . . . Insert

Contents


atw Vol. 63 (2018) | Issue 5 ı May

280

ABSTRACTS | ENGLISH

Yucca Mountain: Can the US Finally End

the $12 Billion Impasse?

NucNet | Page 282

A US federal advisory panel recently took a step in

what could be a lengthy process to determine if a

deep geological nuclear waste repository should

finally be built at Yucca Mountain, a project that has

been on the drawing board since the 1970s at a cost

of around $ 12 bn (€ 9.7 bn).

Detective Application Security Controls

for Nuclear Safety

Deeksha Gupta, Karl Waedt and Yuan Gao | Page 285

The current Draft Nuclear IEC 63096 New Work

Item Proposal (NWIP), a new downstream standard

of IEC 62645, distinguishes between preventive,

detective and corrective security controls. The focus

of this paper is on resilient detective cybersecurity

controls that are needed especially for high security

degrees in the context of Advanced Persistent

Threats (APTs). The approach is fully in line with

Nuclear IEC 62859 that provides requirements

on coordinating safety and cybersecurity. The

recommendations on separating selected detective

security controls from the process control software

can be achieved by avoiding an increased complexity

and the possibility of retroactions of security

measures on safety related functionality.

General Data Protection Regulation (GDPR)

of the European Union – What Had to Be

Considered until 25 May 2018

Stefan Loubichi | Page 289

With the General Data Protection Regulation

( GDPR) of the European Union there will be the

beginning of a new chapter in history of data protection.

With the beginning of May 25, 2018 we will

have harmonized regulations in the European

Union. With penalties up to 20 million euros and

imprisonment up to 3 years, the data protection will

have a high priority in future. In this essay we

present the subject-matter and objectives, material

and territorial scope and the principles relating to

processing of personal data. The GDPR presents a

sustainable change in data protection. For years to

come, this will lay the foundation for trust in data

protection in Europe.

The New Radiation Protection Law:

May Makes Everything New?

Ulrike Feldmann | Page 296

Last summer, a radiation protection law was

launched for the first time in Germany. The Federal

Government was obliged to implement the revised

version of the basic European radiation protection

Directive and emphasised the importance of

radiation protection by ranking it into a law. This

law now has to be filled with “life” at the level of

ordinances, so that it can be applied in practice. The

deadline for implementing the Directive always

expired on 6 February 2018.

Continuous Process of Safety Enhancement

in Operation of Czech VVER Units

J. Duspiva, E. Hofmann, J. Holy, P. Kral

and M. Patrik | Page 299

A continuous process of a safety enhancement of

VVER units in the Czech Republic is briefly described

including a presentation of important milestones

and examples of particular safety measures already

implemented. A special attention is given to the

evaluation and implementation of safety measures

following stress tests recommendations and R&D

activities supporting this process. As examples an

implementation of the “design extension condition

without core melt” concept and various activities

related to severe accident mitigation strategies are

presented in the more detailed way.

Applications of Underwater-Robotics

in Nuclear Power Plants

Gunnar Fenzel, Dr. Dietmar Nieder

and Alexandra Sykora | Page 305

Cutting and packing of the reactor pressure

vessel (RPV) is one important step during decommissioning

of nuclear power plants. Therefore, it is

the objective of the research project Automated

Cutting of Reactor Pressure Vessels Internals Using

Underwater-Robotics (AZURo) to (semi-) automate

frequently repeated activities by an underwater

robot.

This joint research project was sponsored by the

German Federal Ministry of Education and Research

(BMBF). It was executed together with Fraunhofer­

Einrichtung für Gießerei-, Composite- und Verarbeitungstechnik

IGCV. The project AZURo started in

2012 and was finished in 2016.

360 Degree Area Atlas

in the Biblis Nuclear Power Plant

Jürgen Kircher | Page 308

The operation and the dismantling of decommissioned

nuclear power plants is a technical

challenge. It must be fully documented. A helpful

tool for operation and dismantling is the so-called

spatial atlas. The atlas provides the rooms in the

nuclear power plants in high-resolution 360° HDR

images and technical circumstances resulting

therefrom may be deduced.

ELINDER – European Learning Initiatives

for Nuclear Decommissioning and

Environmental Remediation

Pierre Kockerols, Hans Günther Schneider

and Daniela Santopolo | Page 309

The decommissioning of nuclear facilities is an

industrial activity that is expected to grow worldwide,

creating many attractive career opportunities.

European industry has acquired know-how and

today Europe can position itself at the top level in the

world decommissioning market. However, in view of

the expected expansion of the activities, efforts are

necessary to share and enhance the underpinning

knowledge, skills and competences and to ensure the

availability of the necessary workforce in the future.

JRC and partners in the EU decommissioning field

have launched a project to consolidate and improve

existing training programmes. The joint training

programme project is called ‘ELINDER’ (European

Learning Initiatives for Nuclear Decommissioning

and Environmental Remediation) and is implemented

from 2018 onwards.

The New CASTOR® geo – A Comprehensive

Solution For Transport and Storage of Spent

Nuclear Fuel, MOX and Damaged Fuel

Linus Bettermann and

Roland Hüggenberg | Page 312

Dry interim storage has become a common solution

for the disposal of spent fuel in recent years worldwide.

However, in particular the complete defueling

of NPP prior to decommissioning and dismantling

will dramatically increase the demand especially for

non-standard fuel. Here the new dry storage system

by GNS is presented for international markets with

its capability to also store MOX and damaged spent

fuel. The new CASTOR® geo cask system is a product

line based on standardized modules and components

featuring different cask dimensions and

basket designs.

Optimal Holistic Disposal Planning –

Development of a Calculation Tool

Johannes Schubert, Anton Philipp Anthofer

and Max Schreier | Page 316

The expected volume of radioactive waste from

dismantling of nuclear facilities in the forthcoming

scope and the opening of the Konrad disposal

requires an optimised planning of the removal of

radioactive waste. For the treatment of radioactive

raw waste, with negligible heat generation, different

conditioning processes are available. Thereby different

waste volumes and masses with different properties

can result even from the same raw waste. An

optimisation can be realised. The complex process

can be carried out by a calculation tool.

Scope for Thermal Dimensioning

of Disposal Facilities for High-level

Radioactive Waste and Spent Fuel

Joachim Heierli, Helmut Hirsch

and Bruno Baltes | Page 319

The objective of final disposal of high-level radioactive

waste in deep geological formations is to

isolate the radionuclides from the accessible

biosphere for a sufficient period of time. To achieve

this, both the functionality and the integrity of the

disposal system must be assured under ambient

conditions that

depend both on the geological environment and

on engineering choices taken in the planning

of the facility. In particular, the amplitude of

the transient temperature increase caused by the

release of nuclear decay heat in the disposal area

is scalable through design strategies and thermal

dimensioning.

Heavy Ions Irradiation as a Tool

to Minimize the Number of In-Pile Tests

in UMo Fuel Development

H. Breitkreutz, J. Shi, R. Jungwirth,

T. Zweifel, H.-Y. Chiang and W. Petry | Page 325

Irradiation with heavy ions from an accelerator

source is an increasingly often used tool to quickly

reproduce and simulate certain effects of in-pile

irradiation tests, avoiding the complexity and high

costs of handling highly radioactive samples.

At the Maier-Leibnitz Laboratorium (MLL) of the

Technische Universität München (TUM), swift

heavy ions have been applied in the development of

Uranium-Molybdenum (UMo) based research

reactor fuels for more than 10 years. Since then, the

technique has been advanced from feasibility

over qualitative analysis to quantitative prediction,

including fission gas implantation.

Nuclear Newcomer Turkey and ‘Comeback

Kid’ Japan Show the Way Ahead

John Shepherd | Page 354

Around 20 years ago there was one story that

cropped up again and again: “Forget that – it will

never happen, they’ve been talking about it for

years.” The subject was Turkey and its desire to

build the country’s first nuclear power plant. But

today, first safety-related concrete was poured and

finally marked the start of construction of Turkey´s

first nuclear power plant in Akkuyu, with four units

planned. There was progress too on the international

nuclear front from Japan, where KEPCO

confirmed the restart of the third unit of its Ohi

nuclear power plant.

Abstracts | English


atw Vol. 63 (2018) | Issue 5 ı May

Yucca Mountain, USA: Ein Ende der

12-Milliarden-Dollar-Sackgasse?

NucNet | Seite 282

Ein US-Bundesbeirat hat kürzlich einen ersten

Schritt in einem möglicherweise langwierigen

Prozess eingeleitet, um zu entscheiden, ob das Endlagerprojekt

Yucca Mountain, in das seit den 1970er

Jahren rund 12 Mrd. $ geflossen sind, endlich als

geologisches Tiefenlager realisiert werden soll.

Anwendungen für Sicherheitskontrollen

zur Gewährleistung nuklearer Sicherheit

Deeksha Gupta, Karl Waedt und Yuan Gao | Seite 285

Der aktuelle Entwurf der IEC 63096 New Work Item

Proposal (NWIP), eine neue nachgelagerte Norm

der IEC 62645, unterscheidet zwischen präventiven,

detektivischen und korrektiven Sicherheitskontrollen.

Der Schwerpunkt dieses Beitrags liegt

auf belastbaren Cybersicherheitskontrollen, die insbesondere

für hohe Sicherheitsstufen im Kontext

von Advanced Persistent Threats (APTs) erforderlich

sind. Der Ansatz entspricht voll und ganz der

Nuclear IEC 62859, die Anforderungen an die

Koordination von Sicherheit und Cybersicherheit

stellt.

EU-Datenschutzgrundverordnung – Was bis

zum 25.5.2018 beachtet sein muss(te)

Stefan Loubichi | Seite 289

Mit der Datenschutzgrundverordnung (DSGVO)

der Europäischen Union beginnt in ein neues

Kapitel in der Geschichte des Datenschutzes. Zum

25. Mai 2018 werden wir in der Europäischen Union

eine Harmonisierung der Datenschutzbestimmung

vorfinden. Mit Geldbußen von bis zu 20 Millionen

Euro und Freiheitsstrafen von bis zu 3 Jahren

werden die Datenschutzbestimmungen in Zukunft

einen hohen Stellenwert haben.

Das neue Strahlenschutzrecht:

Alles neu macht der Mai?

Ulrike Feldmann | Seite 290

Erstmalig wurde im vergangenen Sommer ein

Strahlenschutzgesetz in Deutschland aus der Taufe

gehoben. Die Bundesregierung hatte die Verpflichtung

zur Umsetzung der europäischen Strahlenschutzgrundnormen

zum Anlass genommen, die

Wichtigkeit des Strahlenschutzrechts durch

Hochzonen in den Gesetzesrang zu unterstreichen.

Dieses Gesetz gilt es nun auf Verordnungsebene

mit „Leben“ zu erfüllen, um es für die Praxis

anwendbar zu machen. Die Frist zur Umsetzung der

EU-Richtlinie war bereits am 6. Februar 2018

bgelaufen.

Kontinuierlicher Prozess der

Sicherheitsoptimierung im Betrieb der

tschechischen WWER-Kernkraftwerke

J. Duspiva, E. Hofmann, J. Holy, P. Kral

und Herr Patrik | Seite 299

Der kontinuierliche Prozess zur weiteren Optimierung

der Sicherheit von WWER-Kernkraftwerken in

der Tschechischen Republik wird vorgestellt mit

wichtigen Meilensteinen und Beispielen für bereits

durchgeführte Maßnahmen. Ein besonderer Fokus

wird auf die Bewertung und Umsetzung von Sicherheitsmaßnahmen

im Anschluss an die „EU-Stresstests“

und F&E-Aktivitäten zur Unterstützung dieses

Prozesses gelegt. Als Beispiele werden eine Umsetzung

des Konzepts „Auslegungserweiterung

ohne Kernschmelze“ und verschiedene Aktivitäten

im Zusammenhang mit Strategien zur Vermeidung

schwerer Unfälle näher diskutiert.

Unterwasser-Robotik in Kernkraftwerken

Gunnar Fenzel, Dr. Dietmar Nieder

und Alexandra Sykora | Seite 305

Das Trennen und Verpacken des Reaktordruckbehälters

(RPV) ist ein wichtiger Schritt bei

der Still legung von Kernkraftwerken. Ziel des

Forschungsprojektes Automated Cutting of Reactor

Pressure Vessels Internals Using Underwater­

Robotics ( AZURo) ist es, häufig wiederkehrende

Tätigkeiten mit einem Unterwasserroboter zu automatisieren.

Das Verbundprojekt wurde vom Bundesministerium

für Bildung und Forschung (BMBF)

gefördert. Es wurde gemeinsam mit der Fraunhofer-Einrichtung

für Gießerei-, Composite- und Verarbeitungstechnik

IGCV durchgeführt. Das Projekt

AZURo begann 2012 und wurde 2016 abgeschlossen.

360° Raumatlas im Kraftwerk Biblis

Jürgen Kircher | Seite 308

Betrieb und Rückbau von Kernkraftwerken stellen

Betreiber, Behörden, Gutachter und letztendlich

die Ingenieure vor vielfältige Herausforderungen.

Unter anderem sind Betrieb und Abbau der

still gelegten Kernkraftwerke zu dokumentieren.

Ein hilfreiches Werkzeug ist der Raumatlas. Er

ermöglicht es, die Räume in Kernkraftwerken in

hochauflösenden 360° HDR Bildern darzustellen

und daraus technische Folgerungen abzuleiten.

ELINDER – Europäische Know-how- Initiativen

für Stilllegung und Umweltsanierung

Pierre Kockerols, Hans Günther Schneider

und Daniela Santopolo | Seite 309

Die Stilllegung kerntechnischer Anlagen ist eine

industrielle Aufgabe, die weltweit viele attraktive

Karrieremöglichkeiten schaffen wird. Die europäische

Industrie hat dazu wichtiges Know-how

erarbeitet und kann sich heute auf dem weltweiten

Stilllegungsmarkt an der Spitze positionieren.

Angesichts der zu erwartenden Ausweitung der

Arbeiten sind Anstrengungen erforderlich, um die

zugrunde liegenden Kenntnisse, Fähigkeiten und

Kompetenzen zu teilen und zu verbessern und die

Verfügbarkeit der erforderlichen Arbeitskräfte für

die Zukunft sicherzustellen. Das JRC und ihre

Partner im Bereich der Stilllegung haben dazu ein

Projekt zur Konsolidierung und Verbesserung

bestehender Ausbildungsprogramme gestartet. Das

gemeinsame Schulungsprojekt heißt „ELINDER“

(European Learning Initiatives for Nuclear Decommissioning

and Environmental Remediation) und

wird ab 2018 angeboten.

Der neue CASTOR® geo – eine umfassende

Lösung für den Transport und die Lagerung

von abgebranntem Kernbrennstoff, MOX

und beschädigten Brennelementen

Linus Bettermann und Roland Hüggenberg | Seite 312

Die trockene Zwischenlagerung hat sich in den

letzten Jahren weltweit zur gängigen Lösung für

abgebrannte Brennelemente entwickelt. Ins besondere

die vollständige erforderliche Brennstofffreiheit

der Kernkraftwerke vor Stilllegung und

Rückbau wird die Nachfrage vor allem für

Nicht-Standard-Brennelemente erhöhen. Dazu wird

das neue Trockenlagersystem der GNS für internationale

Märkte vorgestellt, in das auch MOX­

Elemente und beschädigte abgebrannte Brennelemente

eingelagert werden können. Das neue

CASTOR® geo-System ist eine Produktlinie, die auf

standardisierten Modulen und Kompo nenten mit

unterschiedlichen Behälterab messungen und Korbausführungen

basiert.

Optimale ganzheitliche Entsorgungsplanung

– Entwicklung eines Berechnungstools

Johannes Schubert, Anton Philipp Anthofer

und Max Schreier | Seite 316

Die zu erwartende Menge an radioaktiven Abfällen

aus dem Rückbau kerntechnischer Anlagen im

bevorstehenden Umfang und der Einlagerung der

Abfälle im Endlager Konrad erfordern eine

optimierte Planung. Für die Behandlung radioaktiver

Rohabfälle mit vernachlässigbarer Wärmeentwicklung

stehen verschiedene Konditionierungsverfahren

zur Auswahl. Dabei können auch bei

gleichem Rohabfall unterschiedliche Abfallmengen

und -massen mit unterschiedlichen Eigenschaften

anfallen. Eine Optimierung kann hierbei den Prozess

begleiten. Der komplexe Prozess kann mit einem

Berechnungstool durchgeführt werden, um ein Ergebnis

mit geringstmöglichen Volumina zu erzielen.

Bandbreite der thermischen Dimensionierung

von Endlagern für hoch radioaktive Abfälle

und abgebrannten Kernbrennstoff

Joachim Heierli, Helmut Hirsch

und Bruno Baltes | Seite 319

Ein wichtiger Schritt bei der Planung eines End lagers

für hoch radioaktive Abfälle und abgebrannten Kernbrennstoff

im tiefen Untergrund ist die thermische

Dimensionierung. Die durch Zerfallswärme verursachte

Temperaturzunahme kann zu negativen

Effekten auf Funktionalität und Langlebigkeit

barriere wirksamer Materialien führen. Die vorgestellten

Untersuchungen zeigen, dass frühzeitige Entscheidungen

bei der Standortauswahl den späteren

Handlungsspielraum für die definitive thermische

Dimensionierung von Endlagern erheblich einschränken

können. Sie zeigen auch, dass technische

Lösungen mit deutlich tieferen Temperaturen möglich

wären, wenn der Zielsetzung der raumsparenden

Auslegung eine geringere Priorität zugeordnet kann,

als dies in der Vergangenheit der Fall war.

Schwerionenbestrahlung als Werkzeug zur

Minimierung der Anzahl von In-Pile-Tests in

der UMo-Brennstoffentwicklung

H. Breitkreutz, J. Shi, R. Jungwirth,

T. Zweifel, H.-Y. Chiang und W. Petry | Seite 325

Die Bestrahlung mit schweren Ionen aus einer

Beschleunigerquelle ist ein immer häufiger eingesetztes

Werkzeug, um schnell und effizient bestimmte

Effekte von In-Pile-Bestrahlungstests zu simulieren,

wodurch die Komplexität und die hohen Kosten für

die Handhabung hochradioaktiver Proben vermieden

werden. Am Maier-Leibnitz­ Laboratorium (MLL) der

Technischen Universität München (TUM) werden seit

mehr als 10 Jahren schnelle Schwerionen bei der Entwicklung

von Uran-Molybdän (UMo) basierten

Forschungs reaktorbrennstoffen eingesetzt. Seitdem

wurde die Technik von der Machbarkeit über die

qualitative Analyse zur quantitativen Prognose, einschließlich

der Spaltgaseinschlüsse, weiterentwickelt.

Nuclear Newcomer Türkei und

‘Comeback Kid‘ Japan weisen den Weg

John Shepherd | Seite 354

Vor etwa 20 Jahren gab es eine Geschichte, die

immer wieder auftauchte: „Vergiss das – es wird nie

passieren, sie reden schon seit Jahren darüber.“

Thema war die Türkei und ihr Bestreben das erste

Kernkraftwerk des Landes zu bauen. Jetzt wurde der

erste Beton gegossen mit Baubeginn des ersten Kernkraftwerks

in Akkuyu, für das vier Blöcke geplant

sind. Fortschritte gab es auch international; in Japan

hat KEPCO die Wiederinbetriebnahme des dritten

Blocks seines Kernkraftwerks Ohi bestätigt.

281

ABSTRACTS | GERMAN

Abstracts | German


atw Vol. 63 (2018) | Issue 5 ı May

282

INSIDE NUCLEAR WITH NUCNET

Yucca Mountain: Can the US Finally End

the $12 Billion Impasse?

NucNet

A US federal advisory panel recently took a step in what could be a lengthy process to determine if

a deep geological nuclear waste repository should finally be built at Yucca Mountain, a project that has been

on the drawing board since the 1970s at a cost of around $ 12 bn (€ 9.7 bn).

The panel held a meeting to receive input on

reconstructing an electronic library for

documents needed to decide on the US Department of

Energy’s (DOE) Yucca licence application. The meeting, at

the Nuclear Regulatory Commission’s headquarters in

Maryland, came one week after another development: the

White House pledged $ 120 m of funding in its 2019 federal

budget proposal to restart licensing for the Yucca site, north

of Las Vegas in Nevada, and to establish an interim storage

programme to address the growing stockpile of nuclear

waste produced by nuclear plants across the nation.

The panel meeting, and Donald Trump’s budget proposal,

came despite a lack of federal funding for the Yucca

project and repeated vows by Nevada officials to fight the

process.

In the scheme of things, neither development is likely to

result in any significant short-term progress on Yucca. But

they do mark something of a revival. After decades of

wrangling, could the US finally be on course to resolve the

question of what to do with the high-level nuclear waste

from the nation’s 99 commercial nuclear reactors?

Yucca Mountain’s history is complex, and riven by party

politics and state antipathy. According to the National Conference

of State Legislatures (NCSL), the story of Yucca

Mountain is a cautionary tale about what can happen

when the federal government imposes its will on a state

without its cooperation or consent.

The site was designated by Congress in 1987 as the sole

site to permanently store nuclear waste, but the licensing

process was halted when the Obama administration ended

funding in 2010.

There was, and continues to be, significant support for

using Yucca Mountain to store nuclear waste. Support

comes especially from states housing some of the waste

and from small communities near the site hungry for

new jobs. The state of Texas – home to the Comanche Peak

and South Texas nuclear stations – sued several federal

agencies claiming the federal government had violated the

Nuclear Waste Policy Act by failing to complete the licensing

process at Yucca Mountain. The Act codifies the DOE’s

responsibility for developing a geologic repository for used

nuclear fuel.

There is also substantial resistance. Generally speaking,

Nevadans don’t want nuclear waste stored at Yucca

Mountain and they never have. Surveys show that around

58 % oppose and 33 % support full development of the

repository, according to the Nevada Independent. Governor

Brian Sandoval, a Republican, responded to the renewed

federal interest in Yucca by warning that “any attempt to

resurrect this ill-conceived project will be met with

relentless opposition, and maximum resources”. Democrat

Aaron Ford was equally as strident. He said: “I am

disappointed that the Trump administration is arrogantly

choosing to ignore the fact that Nevadans don’t want

dangerous nuclear waste dumped on our state.”

But the waste has to go somewhere. There is consensus

on one inescapable fact: tonnes of nuclear waste need a

permanent home, and more is coming. Nuclear reactors

have generated more than 76,000 tonnes of nuclear waste

since they first began producing electricity in the late

1950s. That’s the equivalent of a football field covered

almost 10 metres deep in spent nuclear fuel, according to

the Nuclear Energy Institute (NEI), a Washington-based

industry group. And every year the US generates between

2,000 and 2,300 tonnes more, primarily from commercial

nuclear reactors.

The US has been seeking a solution to nuclear waste for

decades. With 99 reactors in 30 states producing almost

20 % of the nation’s electricity, concern is building.

Without a central facility to send the high-level radioactive

waste to, energy generators have been storing it on site in

steel canisters, in concrete-lined pools of water or in dry

casks.

The NEI sums up the industry’s view, saying the DOE

shut down the Yucca Mountain project without citing any

technical or safety issues. In contrast, decades of scientific

study had “consistently concluded that the proposed

repository could safely protect future generations”.

The NEI says that at the time of the shutdown, in 2010,

$ 12 bn had already been spent on Yucca Mountain and

65,000 tonnes of spent fuel were in temporary storage

across 33 states. In 2014, a federal court ordered the NRC

to complete safety and environmental reviews of the site.

However, while these reviews have since concluded that

Yucca Mountain complies with all regulations, a final

decision awaits an extensive formal hearing. “That hearing

can’t happen until Congress funds it,” said the NEI.

How Did We Get Here?

The Nuclear Waste Policy Act established a national

programme for the safe and permanent disposal of spent

nuclear fuel and high-level radioactive waste. It included a

small fee utilities passed on to consumers to help pay for it

all. In 1987, Congress designated Yucca Mountain to be the

Inside Nuclear with NucNet

Yucca Mountain: Can the US Finally End the $12 Billion Impasse? ı NucNet


atw Vol. 63 (2018) | Issue 5 ı May

permanent disposal site for the waste. The federal government

carried out geological and environmental impact

studies over several years to prepare the site, promising

utility companies it would transport their stockpiles of

spent fuel to the site by 1998. The problem is, the federal

government never received support for the project from

Nevada, which, ironically, has no commercial nuclear

plants itself.

In 2009, the DOE determined the repository was

unworkable and the Obama administration agreed to cut

funding. It has been a costly decision. Over the past

20 years, the federal government has paid more than

$ 4.5 bn in damages to utilities for not taking ownership

of the spent fuel as promised.

Since then, the DOE has been using a different tactic

to identify communities that may be willing to host a

repository. This “consent-based approach” encourages

input from the public and from state, local and tribal

officials. Proponents say it is designed to be a transparent

process that considers the public as partners in managing

nuclear waste. The DOE has hosted eight public meetings

around the country to structure the process and determine

what issues should be included. The DOE’s next step is to

design a framework to educate communities about the

pros and cons of siting a facility, including the increased

national security of having all the waste in one location

and the potential economic benefits to the host community.

The DOE said it hopes that by bringing states together,

it can finally find a willing and informed community to

host a storage site with a publicly acceptable system for

transporting waste to it.

According to the NCSL, finding communities that will

take the nuclear waste is not the most difficult part of the

problem. The jobs, federal money and other economic

benefits that follow a nuclear waste site make it attractive

to many.

The most significant hurdle can be convincing others

in the state that the benefits of accepting nuclear waste

outweigh the potential risks. “Finding a consenting

community is merely a first step,” wrote William Alley, the

former chief of the Yucca Mountain waste storage site, in

an opinion piece in New Scientist. “The harder part is

getting everyone else to sign on.”

Meanwhile, NRC commissioners have directed staff to

start gathering information aimed at preparing for the

resumption of Yucca Mountain’s licensing. The NEI has

underlined that consent-based siting should not take

precedence over the government’s legal obligations to find

a repository site.

Whether that site will be Nevada or somewhere else,

seems no closer to a resolution than it was when Yucca

Mountain was approved by President and Congress in 2002.

Author

NucNet

The Independent Global Nuclear News Agency

Editor responsible for this story: David Dalton

Editor in Chief, NucNet

Avenue des Arts 56

1000 Brussels, Belgium

www.nucnet.org

DATF EDITORIAL NOTES

283

Notes

Gross electricity production

in Germany 2017

The eight nuclear power plants in Germany produced about

76 billion kWh of electricity in 2017 which accounts for 11.7 percent

of all gross electricity production in Germany. 50.7 percent of

electricity produced in Germany came from fossil energy carriers.

4.3

Other

sources

11.7

Nuclear

energy

22.5

Lignite

Gross electricity production

(654,8 billion kWh) 2017 in percent

14.1

Hard coal

33.3

Renewable

energy

among:

3.1 Hydro power

13.5 Wind power onshore

2.7 Wind power offshore

6.9 Biomass

6.1 Photovoltaics

0.9 Garbage

Quelle: AG Energiebilanzen; Stand: 2. Februar 2018

13.2

Gas

0.9

Petroleum products

For further details

please contact:

Nicolas Wendler

DAtF

Robert-Koch-Platz 4

10115 Berlin

Germany

E-mail: presse@

kernenergie.de

www.kernenergie.de

DAtF Notes


atw Vol. 63 (2018) | Issue 5 ı May

284

CALENDAR

Calendar

2018

08.05.-10.05.2018

29 th Conference of the Nuclear Societies in Israel.

Herzliya, Israel. Israel Nuclear Society and Israel

Society for Radiation Protection, ins-conference.com

13.05.-19.05.2018

BEPU-2018 – ANS International Conference on

Best-Estimate Plus Uncertainties Methods. Lucca,

Italy, NINE – Nuclear and INdustrial Engineering S.r.l.,

ANS, IAEA, NEA, www.nineeng.com/bepu/

13.05.-18.05.2018

RadChem 2018 – 18 th Radiochemical Conference.

Marianske Lazne, Czech Republic, www.radchem.cz

14.05.-16.05.2018

ATOMEXPO 2018. Sochi, Russia,

atomexpo.ru

15.05.-17.05.2018

11 th International Conference on the Transport,

Storage, and Disposal of Radioactive Materials.

London, United Kingdom, Nuclear Institute,

www.nuclearinst.com

20.05.-23.05.2018

5 th Asian and Oceanic IRPA Regional Congress

on Radiation Protection – AOCRP5. Melbourne,

Australia, Australian Radiation Protection Society

(ARPS) and International Radiation Protection

Association (IRPA), www.aocrp-5.org

23.05.-25.05.2018

15 th International Conference of Young Scientists

on Energy Issues. Kaunas, Lithuania, Lithuanian

Energy Institute in partnership with Nuclear Society

of Lithuania, cyseni.com/

29.05.-30.05.2018

49 th Annual Meeting on Nuclear Technology

AMNT 2018 | 49. Jahrestagung Kerntechnik.

Berlin, Germany, DAtF and KTG,

www.nucleartech-meeting.com – See you there!

03.06.-07.06.2018

38 th CNS Annual Conference and 42nd CNS-CNA

Student Conference. Saskotoon, SK, Canada,

Canadian Nuclear Society CNS, www.cns-snc.ca

03.06.-06.06.2018

HND2018 12 th International Conference of the

Croatian Nuclear Society. Zadar, Croatia, Croatian

Nuclear Society, www.nuklearno-drustvo.hr

04.06.-05.06.2018

13 th European Nuclear Energy Forum. Bratislava,

Slova Republic, European Commission, ec.europa.eu

04.06.-07.06.2018

10 th Symposium on CBRNE Threats. Rovaniemi,

Finland, Finnish Nuclear Society, ats-fns.fi

04.06.-08.06.2018

5 th European IRPA Congress – Encouraging

Sustainability in Radiation Protection.

The Hague, The Netherlands, Dutch Society for

Radiation Protection (NVS), local organiser,

irpa2018europe.com

04.06.-05.06.2018

eurelectric Power Summit 2018 – “Watt’s Next?

New Players, Products and Peers”. Ljubljana,

Slovenia, eurelectric, www.eurelectric.org

06.06.-08.06.2018

2 nd Workshop on Safety of Extended Dry Storage

of Spent Nuclear Fuel. Garching near Munich,

Germany, GRS, www.grs.de

06.06.-07.06.2018

2018 Decommissioning Strategy Forum. Nashville,

TN, U.S.A. go.exchangemonitor.com

14.06.-15.06.2018

Kernkraftwerk Rheinsberg – Zukunft eines

kulturellen Erbes. Kernkraftwerk Rheinsberg,

Germany, EWN GmbH, Stadtgeschichte Rheinsberg

e.V., stadtgeschichte.rheinsberg@gmail.com,

www.ewn-gmbh.net

18.06.-19.06.2018

Decom2018. London, United Kingdom, Nuclear

Industry Association (NIA), decom2018.co.uk

25.06.-26.06.2018

index2018 – International Nuclear Digital

Experience. Paris, France,

Société Française d’Energie Nucléaire,

www.sfen.org, www.sfen-index2018.org

27.06.-29.06.2018

EEM – 2018 15 th International Conference on the

European Energy Market. Lodz, Poland, Lodz

University of Technology, Institute of Electrical

Power Engineering, Association of Polish Electrical

Engineers (SEP), www.eem18.eu

24.06.-30.06.2018

ANNETTE Summer School on Nuclear Technology,

Nuclear Waste Management and Radiation

Protection. Turku, Finland, Advanced Networking

for Nuclear Education, Training and Transfer of

Expertise, annettesummerschool.org, www.enen.eu

29.07.-02.08.2018

International Nuclear Physics Conference 2019.

Glasgow, United Kingdom, www.iop.org

22.08.-31.08.2018

Frédéric Joliot/Otto Hahn (FJOH) Summer School

FJOH-2018 – Maximizing the Benefits of

Experiments for the Simulation, Design and

Analysis of Reactors. Aix-en-Provence, France,

Nuclear Energy Division of Commissariat à l’énergie

atomique et aux énergies alternatives (CEA)

and Karlsruher Institut für Technologie (KIT),

www.fjohss.eu

28.08.-31.08.2018

TINCE 2018 – Technological Innovations in

Nuclear Civil Engineering. Paris Saclay, France,

Société Française d’Energie Nucléaire, www.sfen.org,

www.sfen-tince2018.org

03.09.-06.09.2018

Jahrestagung des Fachverbandes Strahlenschutz.

Dresden, Germany, Fachverband für Strahlenschutz

e.V., www.fs-ev.org

05.09.-07.09.2018

World Nuclear Association Symposium 2018.

London, United Kingdom, World Nuclear Association

(WNA), www.world-nuclear.org

09.09.-14.09.2018

21 st International Conference on Water

Chemistry in Nuclear Reactor Systems.

San Francisco, CA, USA, EPRI – Electric Power

Research Institute, www.epri.com

17.09.-21.09.2018

62 nd IAEA General Conference. Vienna, Austria.

International Atomic Energy Agency (IAEA),

www.iaea.org

17.09.-20.09.2018

FONTEVRAUD 9. Avignon, France,

Société Française d’Energie Nucléaire (SFEN),

www.sfen-fontevraud9.org

17.09.-19.09.2018

4 th International Conference on Physics and

Technology of Reactors and Applications –

PHYTRA4. Marrakech, Morocco, Moroccan

Association for Nuclear Engineering and Reactor

Technology (GMTR), National Center for Energy,

Sciences and Nuclear Techniques (CNESTEN) and

Moroccan Agency for Nuclear and Radiological

Safety and Security (AMSSNuR), phytra4.gmtr.ma

26.09.-28.09.2018

44 th Annual Meeting of the Spanish Nuclear

Society. Avila, Spain, Sociedad Nuclear Española,

www.sne.es

30,.09.-05.10.2018

14 th Pacific Basin Nuclear Conference (PBNC).

San Francisco, CA, USA, pbnc.ans.org/

30.09.-04.10.2018

TopFuel 2018. Prague, Czech Republic, European

Nuclear Society (ENS), American Nuclear Society

(ANS). Atomic Energy Society of Japan, Chinese

Nuclear Society and Korean Nuclear Society,

www.euronuclear.org

01.10.-05.10.2018

3 rd European Radiological Protection Research

Week ERPW. Rovinj, Croatia, ALLIANCE, EURADOS,

EURAMED, MELODI and NERIS, www.erpw2018.com

02.10.-04.10.2018

7 th EU Nuclear Power Plant Simulation ENPPS

Forum. Birmingham, United Kingdom, Nuclear

Training & Simulation Group, www.enpps.tech

14.10.-18.10.2018

12 th International Topical Meeting on Nuclear

Reactor Thermal-Hydraulics, Operation and

Safety – NUTHOS-12. Qingdao, China, Elsevier,

www.nuthos-12.org

14.10.-18.10.2018

NuMat 2018. Seattle, United States,

www.elsevier.com

16.10.-17.10.2018

4 th GIF Symposium at the 8th edition of Atoms

for the Future. Paris, France, www.gen-4.org

22.10.-24.10.2018

DEM 2018 Dismantling Challenges: Industrial

Reality, Prospects and Feedback Experience. Paris

Saclay, France, Société Française d’Energie Nucléaire,

www.sfen.org, www.sfen-dem2018.org

24.10.-26.10.2018

NUWCEM 2018 Cement-based Materials for

Nuclear Waste. Avignon, France, French

Commission for Atomic and Alternative Energies

and Société Française d’Energie Nucléaire,

www.sfen-nuwcem2018.org

24.10.-25.10.2018

Chemistry in Power Plant. Magdeburg, Germany,

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

05.11.-08.11.2018

International Conference on Nuclear

Decom missioning – ICOND 2018. Aachen,

Eurogress, Germany, achen Institute for Nuclear

Training GmbH, www.icond.de

06.11-08.11.2018

G4SR-1 1 st International Conference on

Generation IV and Small Reactors. Ottawa, Ontario,

Canada. Canadian Nuclear Society (CNS), and

Canadian Nuclear Laboratories (CNL), www.g4sr.org

03.12.-14.12.2018

United Nations, Conference of the Parties –

COP24. Katowice, Poland, United Nations

Framework Convention on Climate Change –

UNFCCC, www.cop24.katowice.eu

2019

07.05.-08.05.2019

50 th Annual Meeting on Nuclear Technology

AMNT 2019 | 50. Jahrestagung Kerntechnik.

Berlin, Germany, DAtF and KTG,

www.nucleartech-meeting.com

Calendar


atw Vol. 63 (2018) | Issue 5 ı May

Detective Application Security Controls

for Nuclear Safety

Deeksha Gupta, Karl Waedt and Yuan Gao

The current Draft Nuclear IEC 63096 New Work Item Proposal (NWIP), a new downstream standard of IEC

62645, distinguishes between preventive, detective and corrective security controls. The focus of this paper is on

resilient detective cybersecurity controls that are needed especially for high security degrees in the context of Advanced

Persistent Threats (APTs). The Stuxnet malware demonstrates that sophisticated attacks on physical processes

can make use of both, the manipulation of output signals that control the automation equipment in parallel with

manipulations of the graphical process feedback information displayed to users. In the international IAEA coordinated

research proposal CRP J02008 several project partners investigate (as one of the objectives) the development of

detective security controls that do not rely on the process control software itself. Thus, a manipulation of the process

control software could still be detected by the detective security controls implemented by diverse means.

Implementing detective security controls at this conceptual

level requires knowledge about selected analog and binary

variables corresponding to the current state of physical

processes. This knowledge (expressed as modeled specifications

of expected safe transitions and value ranges) is

needed in order to detect potential manipulations. In

the case of Stuxnet this corresponds to detecting the

high frequency speed variations of centrifuges (with the

aim of physically destroying them) without regard to

the ( manipulated, maliciously reassuring) information

displayed to operators.

This paper will address the selection of process

variables, the different points at which these process

variables can be acquired, the equipment used for the

acquisition (e.g. additional networking equipment or

enhanced embedded software), the aggregation of the

variables, the detective logic and the reporting towards a

Security Information and Event Management System

( SIEM). The paper will also explain how existing nuclear

safety impact analyses can be leveraged by replacing the

typical safety events (e.g. due to ageing or random failures)

by graded targeted security attack events.

The semi-formal description of the detective security

controls will make use of the Application Security Control

(ASC) concepts as introduced by ISO/IEC 27034-x.

This approach is fully in line with Nuclear IEC 62859

that provides requirements on coordinating safety

and cybersecurity. The recommendations on separating

selected detective security controls from the process

control software can be achieved by avoiding an increased

complexity and the possibility of retroactions of security

measures on safety related functionality.

to physical property, computer system, information or

other assets.

The detective application security controls provide

alternative or enhanced protection means as part of a

security defense-in-depth concept. Accordingly, for a new

NPP they may serve as security enhancements while for an

existing NPP they may serve as an effective and cost-saving

alternative, in cases where the legacy process control

software cannot be easily revised to include cybersecurity

specific functionality.

In principle the term “Cybersecurity Control” or just

“Security Control” denotes one or a set of Cyber Security

countermeasures. Security Controls are applied during all

lifecycle phases of the I&C-product or project [4]. In an

( initially) non safety related context several Security

Controls are similar to measures that are by default implemented

in safety systems. These include, e.g., the degree

of requirements traceability (e.g., to avoid the intro duction

of requirements on debugging functionality that may later

be misused for manipulations) and requirements on

testing. However, the Security Level [2, 3] of a subsystem

may be increased due to results of a risk assessment [5].

This may mandate the implementation of individual

security controls that have to be met only for higher safety

classes [4].

IEC 62645 [8] is the top level nuclear cybersecurity

standard. It defines security controls for the nuclear

285

ENVIRONMENT AND SAFETY

Introduction

Cybersecurity is considered as a major element of physical

protection plans for nuclear facilities and it should be

implemented from the initial phase of nuclear facility

design [1]. In nuclear power plants (NPPs), I&C systems

have evolved during the last decades from non-digital

equipment and standalone environments to digital

technologies and interconnected systems. Such an

evolution exposes them to risks related to cyberattacks.

The cyberattack can later affect the safety of the NPP.

Cyberattack or digital attack is the attempt by digital

means to destroy, expose, alter, disable, steal or gain

unauthorized access to or make unauthorized use of an

asset [2]. To minimize these risks security controls

provides the countermeasures to avoid, detect, counteract

| | Fig. 1.

Example Plant Operation related to Security Control for a Detective System [6, 7].

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atw Vol. 63 (2018) | Issue 5 ı May

ENVIRONMENT AND SAFETY 286

specific Security Degrees, SD1 (most stringent requirements),

SD2 and SD3 (less stringent requirements) as well

as for security Baseline Requirements. This consider I&C

systems of Safety Classes 1, 2, 3 and non-classified (NC)

I&C systems [9], without requiring direct mapping

between Security Degrees and Safety Classes as shown in

Figure 1.

Advanced Persistent Threat (APT)

Major discussions regarding APTs started after the Stuxnet

exploited several zero-day vulnerabilities (that were not yet

known to the equipment vendors) [10, 11]. Experts raised

concerns on how to protect critical infrastructure against

exploitation of unidentified day-zero vulner abilities [10,

11]. The vulnerabilities are not only considered in software

or firmware, but also in the lifecycles of technical, operational

and management of cybersecurity controls. Zero-day

vulnerabilities may exist e.g. in commercial-of-the-shelf

(COTS) software or firmware, custom-based I&C, and

system designs [11]. Also, different versions of Stuxnet

were able to upgrade to the newest version in the same

network. As cyber protection systems cannot immediately

recognize zero-day vulnerabilities, they can stay undetected

for long time [10, 11]. The above concerns regarding

zero­ day vulnerabilities emphasize importance of using

detective security controls in the systems of a NPP for

cybersecurity.

Security Controls

Security controls are the countermeasures to avoid,

detect, counteract, or minimize security risks to physical

property, computer system, information or other assets

[8]. According to IEC 62645 [8] security controls can be

divided in the following three categories:

• Technical Controls: hardware and/or software solutions

for the protection, detection and mitigation of and

recovery from intrusion or other malicious acts.

• Physical controls: physical barriers for the protection

of computer and supporting assets from physical

damage and unauthorized physical access. The physical

controls include barriers such as locks, physical

encasements, smart electrical cabinets; tamper seals,

isolation rooms, gates and guards.

• Administrative controls: policies, procedures and

practices designed to protect computer systems by

controlling personnel actions and behaviors. The

administrative controls are directive in nature,

specifying what employees and third party personnel

should and should not do. In the nuclear environment,

administrative controls are understood to include

operational and management controls.

Stuxnet

Stuxnet – a powerful and malicious piece of code, is a

500-kilobyte computer worm that affected the software of

minimum 14 industrial locations in Iran [10, 12]. One of

the affected locations was a uranium-enrichment plant.

While a computer virus depends on an individual person to

perform installation, a worm spreads on its own, frequently

throughout computer networks [12]. The worm attacked

in three phases. First, Microsoft Windows® machines and

networks were under the attack. The worm repeatedly

replicated itself. Then, it tried to find Windows-based

Siemens S7 software and used to program industrial

control systems that control equipment, e.g. centrifuges

[10, 12]. Lastly, it compromised the programmable logic

controllers. Two things are important to notice in the case

of Stuxnet; first, results were hidden consequently the

adversary could spy and infiltrate the industrial systems

and force the fast-spinning centrifuges to split themselves

apart without being recognized by operator (e.g. by

displaying valid graphical charts that originated from past

safe plant states); and second – input and output of the

system both were manipulated at the same time. It is

interesting to mention that the development of Stuxnet

started in 2007. Figure 2 provides a cybersecurity threat

landscape timeline related to critical infrastructure,

including NPPs.

| | Fig. 2.

Critical Infrastructure related Cybersecurity Threat Landscape [10].

| | Fig. 3.

Categorization of Security Control.

All three of these elements are critical to the creation of

an effective control environment. Cybersecurity program

shall involve the use of the above mentioned three types

of cybersecurity controls. Cybersecurity controls may

contribute in different manners, mostly by contributing to

– the prevention of cybersecurity events; their detection;

correction, reaction and response [8]. Security controls

could be used to solve the problem of Advanced Persistent

threat like Stuxnet. Figure 3 illustrates the characterization

of security controls.

Preventive: These are controls that prevent the loss or

harm from occurring.

Detective: These controls monitor activity to identify

instances where practices or procedures were

not followed.

Corrective: Corrective controls restore the system or process

back to the state prior to a harmful event.

Security defense-in-depth

Security defense-in-depth is an approach to security

in which multiple and independent security controls,

covering organizational, technical and operational aspects,

are deployed in an architecture, as no individual security

control can provide the expected security [8]. In such

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atw Vol. 63 (2018) | Issue 5 ı May

(a)

(b)

| | Fig. 4.

(a) Data Diode (High Sec. to Low Sec. Zone),

(b) Data Diode (Low Sec. to Higher Sec.).

approach, it is the set of diversified and independent

security controls which is able to bring the needed

prevention, detection and response capabilities. The

security defense-in-depth principle shall guide the

selection of security controls [8].

Detective Security Controls

The one most frequently used security control is detective

controls – identifying events after they have happened.

Depending on how soon the detective control is invoked

after an event, a system may uncover a loss long after there

is any opportunity to limit the amount of damages.

Following detective security controls could be used in

NPPs to protect control system of a NPP from emerging

and sophisticate cyber-attacks:

1. Data diodes

Data diodes operate on a simple rule – data moves just in

one direction between networks [14, 15]. For example, in

the nuclear domain, data diodes are used as possible

segregation enforcement between safety and control

networks as well as isolated engineering support networks

and enterprise networks (intranets).

Data diodes can provide enhanced security by e.g.

allowing either incoming or outgoing data. As presented in

Figure 4(a), Data diode just allows data transfer from

High Security Zone to Low Security Zone. In this way,

no manipulation to data can be done and therefore,

manipulated data cannot be sent to I&C system.

Figure 4(b) indicates Data diode data transmission

from a Low Security Zone, e.g. Intranet to Higher Security

Zone. In this way, confidential information will not be

stolen through the Low Security Zone and no control/

command manipulations can be initiated from the Low

Security Zone. Often, lower security zone is composed of

only monitoring systems and in some cases a connection

is only required for time stamping [15, 16]. Figure 5

elaborates how Data diodes maintain the employment

of solutions that meet requirements of high security

levels [17].

The advantage of Data diodes is that they are secure as

cyber-attacks mainly depend on bidirectional traffic [15,

16]. Therefore, data diode could be used as a tool of

detective security control. For e.g. data diode would not

allow to delete any data from the system, which could be

| | Fig. 5.

Secure Centralized Logging via Data Diode [6, 7].

intractable if done once. And also notification will be sent

to the main server by data diodes if there any change

would be made in log file.

2. Preventive, detective and corrective security

control

By assigning detective control, it is possible to collect

digital evidence later on. Based on an annotation by

preventive, detective, corrective and mitigating security

controls potential vulnerabilities at the cybersecurity

architecture or design level may be detected by an analysis

of prioritized paths along the branches of the attack trees

[18]. The specification of the Application Security Controls

may optionally indicate their protective strength with

regard to a specific unskilled/trained/sophisticated/

persistent attacker. Figure 6 shows an Attack Tree Analysis

(ATA) for e.g. of a Turbine Island Electrical System (TI ES)

[18], based on an elaboration by preventive, detective and

corrective security controls.

Attack Trees are based on the semi-formal notation,

e.g. with regard to Causeways and their assumed (and

later on to be assessed) properties. The ATA may start at

an Environment (containing staff) where an attacker

is assumed. Based on the Business Connections the

possible targets (Business Domains) can be evaluated

and the attack trees (spanning the Communication

Connections) can be generated [18].

The security controls are structured in line with ISO/

IEC 27002 [19]. In order to reflect the special nuclear I&C

requirements like handling security of legacy topics, an

additional nuclear I&C security specific IEC 63096 [7] is

introduced to extend the SC45A series of documents

addressing cybersecurity. In future, this standard can also

be used as a basis for refurbishment projects.

3. Standard detective controls

Logging is another detective control, e.g. card reader

indication. If a system is operated by ten log files which are

configurable to some extent. Limited status changes should

be allowed in system log file. Logging events are generated

after each security status change and if for e.g., in every

100 events, there are at most 20 changing speed. But if it is

up to 50 now then it may be an attack. A policy setting

ontrols Event Log behavior when the log file reaches its

maximum size. If this policy is enabled then if setting and a

log file reaches its maximum size new events are not

written to the log and are lost. If it is disabled or do not

configure this policy setting and a log file reaches its

maximum size new events overwrite old events. When

logging is disabled, time losses are evident as security

events should be identified immediately as they occur and

ENVIRONMENT AND SAFETY 287

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ENVIRONMENT AND SAFETY 288

| | Fig. 6.

Attack Tree Analysis based on Security Controls.

then associated to a timeline. Therefore log file either

should be on or enabled.

Record 1

(deleted)

Record 2

(new generated)

4. Direct logging without relying on process

control software

Direct and completely independent logging of the data

is another detective security control. The data should

be stored in analog and binary variables before collecting

it together like in Security Information and Event

Management (SIEM) [20]. As shown in the Figure 7, the

values of regular changes in centrifuge pump speed could

be stored in a separate system before sending them to the

central data storage system. This security control will help

to keep the data secured in case of attacker makes any

changes in the central display of the data.

5. Other security controls

Periodical scan of the software should be done to avoid

some abnormal things in the system. And also handling of

new generated file is important to avoid any cyber-attack.

| | Fig. 7.

Direct Logging of Speed of a Centrifuge Pump.

100_Room1_

2016-06-15_12:38

| | Tab. 1.

An Example of Naming Convention.

250_Room1_

2016-06-15_12:38

NEI 08-09 requires log records are reviewed at least every

92 days, or as required by the Physical Security Plan.

In addition, each generated new file should follow the

predefined naming. Individual generated file should have

a source. In addition, new created file should have property

to specify its source as shown in the Table 1 so that no one

can override the existing file with malicious codes.

Conclusion

Non-targeted cybersecurity attacks, e.g. malware that is

not conceived for the manipulation a specific target,

can be sophisticated. Completely preventing these attacks

as well as Advance Persistent Threats cannot be addressed

only by protective security controls. Detective security

controls are necessary to identify attacks within a

reasonable time frame and they are the precondition

for initiating the application of protective controls.

Accordingly, resilient implementations of detective

security controls are needed, in line with the stringency

of the security grading. In this paper the most important

detective controls, like the logging at all relevant levels

were addressed. The benefits of securing the information

generated by the detective controls by using data diodes

were explained together with example architectures

that merge the collected security intelligence as input

of a central Security Information and Event System.

As two important conceptual extensions, the security

controls based on independent data collection, directly

from physical devices and the monitoring and evaluation

of analog and binary signal value transients were

introduced. Implementing appro priate combinations of

these resilient detective security controls will help in

improving the security posture of refurbished and new

power plants.

Environment and Safety

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atw Vol. 63 (2018) | Issue 5 ı May

Acknowledgments

Some of the addressed modelling and cybersecurity

related topics are being elaborated as part of AREVA

GmbH’s (today Framatome GmbH) participation in the

“SMARTEST” R&D (2015-2018) with German University

partners, partially funded by German Ministry BMWi.

References

[1] M. Holt, A. Andrews, Nuclear Power Plant Security and

Vulnerabilities, Congressional Research Service. (n.d.).

[2] IEC 62859:2016, NPPs – I&C Systems – Requirements for

Coordinating Safety and Cybersecurity.

[3] IEC 61513:2011, NPPs – I&C Important to Safety – General

Requirement for Systems.

[4] K. Waedt, Y. Ding: 2015, Safety and Cybersecurity Aspects

in the Safety I&C Design for Nuclear Power Plants, 3 rd China

( International) Conference on Nuclear Power I&C Technology,

Shanghai.

[5] ISO/IEC 27005:2011 – Information Technology – Security

Techniques – Information Security Risk Management.

[6] M. Parekh, K. Waedt, A. Ciriello, Y. Gao: 2016, Cybersecurity

During Plant Operation, 42 nd Annual Meeting of the SNE,

Santander.

[7] IEC 63096 (Draft): 2016, Security Controls.

[8] IEC 62645:2014, NPPs – I&C Systems – Requirements for Security

Programmes for Computer-Based Systems.

[9] IEC 61513:2011 – NPPs – I&C important to safety – General req.

for systems.

[10] E. Bajramovic, D.Gupta: 2016, Providing Security Assurance in

Line with National DBT Assumptions, Women in Nuclear (WiN),

Shah Alam, Malaysia.

[11] P. Zavarsky, K. Waedt, A. Kuskov: 2015, High Assurance

Cybersecurity Controls against Persistent and Targeted Attacks

on Instrumentation and Control Systems in Nuclear Facilities,

9 th International Conference on Nuclear Plant Instrumentation,

Control & Human-Machine Interface Technologies (NPIC & HMIT

2015), Charlotte, USA.

[12] D. Kushner: 2013, The Real Story of Stuxnet, IEEE Spectrum.

[13] IAEA NSS 8:2008, Nuclear Security Series No. 8, Technical

Guidance, Preventive and Protective Measures Against Insider

Threats.

[14] A. Scott, 2015, Tactical Data Diodes in IACS, SANS Institute.

[15] E. Bajramovic, J. Bochtler, I. B. Zid, A. Lainer, 2016, Planning the

Selection and Assignment of Security Forensics Countermeasures,

ICONE; Shanghai.

[16] J. Li, E. Bajramovic, Y. Gao, M. Parekh, 2016, Graded Security

Forensics Readiness for SCADA Systems, GI 2016, Klagenfurt.

[17] E. Knapp, J. Langill: 2014, Security Monitoring of Industrial

Control Systems, In Industrial Network Security. 2 nd Edition,

Syngress Publishing.

[18] K. Waedt, A. Kuskov, P. Zavarsky: 2015, Domain Specific

Cybersecurity Applied to I&C, IAEA, Vienna.

[19] ISO/IEC 27002:2013 Information Technology – Security

techniques – Code of Practice for Information Security Controls.

[20] Y. Gao, X. Xie, M. Parekh, E. Bajramovic: 2016, SIEM: Policy-Based

Monitoring of SCADA System, GI 2016, Klagenfurt.

Authors

Deeksha Gupta

Karl Waedt

Framatome GmbH

ICPGOP

Henri-Dunant-Straße 50

91058 Erlangen, Germany

Yuan Gao

Framatome GmbH and

Otto-von-Guericke-Universität Magdeburg,

Germany

289

ENERGY POLICY, ECONOMY AND LAW

EU-Datenschutzgrundverordnung –

Was bis zum 25.5.2018 beachtet

sein muss(te)

Stefan Loubichi

Mit der Datenschutzgrundverordnung (DSGVO) der Europäischen Union beginnt in ein neues Kapitel in der

Geschichte des Datenschutzes. Zum 25. Mai 2018 werden wir in der Europäischen Union eine Harmonisierung der

Datenschutzbestimmung vorfinden. Mit Geldbußen von bis zu 20 Millionen Euro und Freiheitsstrafen von bis zu

3 Jahren werden die Datenschutzbestimmungen in Zukunft einen hohen Stellenwert haben.

In diesem Aufsatz werden erst einmal

Gegenstand und Ziele, sach licher und

räumlicher Anwendungsbereich sowie

die Grundsätze für die Verarbeitung

personenbezogener Daten vorgestellt.

In der neuen DSGVO wurden die

Rechte der betroffenen Personen

präzisiert. Dies wird ebenso vorgestellt

wie die zusätzlichen Pflichten

von Verantwortlichen und Auftragsverarbeitern.

Neue Regelungen in Bezug auf:

• den Datenschutzbeauftragten

• die Aufsichtsbehörde

• die Haftung

• die Auftragsdatenverarbeitung

• das Verzeichnis der Verarbeitungstätigkeiten

• die Datenschutzfolgeabschätzung

• die Meldepflicht bei Datenpannen

• technisch-organisatorische

Maßnahmen und

• die Datenübermittlung ins Ausland

werden in diesem Aufsatz ebenso vorgestellt.

Die DSGVO stellt einen nachhaltigen

Wechsel im Datenschutz dar. Für die

kommenden Jahre wird hier der

Grundstock für das Vertrauen in den

Datenschutz in Europa gelegt.

Worum geht es eigentlich

beim Datenschutz im

Unternehmen?

Beim Datenschutz im Unternehmen

unterscheiden wir:

I. den internen Bereich:

Hier geht es um den datenschutzkonformen

Umgang mit den personenbezogenen

Daten der

Beschäftigten.

II. den externen Bereich:

Hier geht es um den datenschutzkonformen

Umgang mit den

personenbezogenen Daten aller

Personen, mit denen das

Energy Policy, Economy and Law

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atw Vol. 63 (2018) | Issue 5 ı May

ENERGY POLICY, ECONOMY AND LAW 290

Unternehmen im Rahmen seiner

Tätigkeit in Berührung kommt.

Zu beachten sind hier:

I. auf europäischer Ebene:

EU Datenschutz-Grundverordnung

(DSGVO)

VO (EU) 679/2016

II. auf nationaler Ebene:

Bundesdatenschutzgesetz (BDSG),

u.a.

Grundrechtlich ist zu verweisen:

I. auf nationaler Ebene:

Grundrecht auf informationelle

Selbstbestimmung (siehe hierzu

das Volkszählungsurteil des

BVerfG vom 15.12.1983 [BVerfGE

65,1] sowie Art. 1 Abs. 1 GG

[Menschenwürde] und Art. 2

Abs. 1 GG [Persönlichkeitsrecht])

II. auf europäischer Ebene:

Grundrecht auf den Schutz

personen bezogener Daten (siehe

hierzu Art. 8 Charta der Grundrechte

der Europäischen Union

[EU-GRCharta]

Die DSGVO gilt gemäß Art. 3 Abs. 1

für Verarbeiter und Auftragsverarbeiter

in der EU als auch gemäß Art. 3

Abs. 2, so dass im Rahmen des Marktortprinzips

die DSGVO auch für

Anbieter aus den USA und anderen

Drittstaaten gilt. Das Marktortprinzip

ist dabei an folgende Bedingungen

geknüpft:

1. Die Betroffenen sind in der EU

ansässig.

2. Den Betroffenen werden Waren/

Dienstleistungen angeboten oder

die Verarbeitung dient zur Beobachtung

des Verhaltens bzw. des/

der Betroffenen.

In der Anwendung geht die DSGVO

nationalen Datenschutzgesetzen stets

vor. Mit Gesetz vom 30.6.2017 wurde

das EU-Recht an das Bundesdatenschutzgesetz

angepasst. Das „neue“

Bundesdatenschutzgesetz tritt zum

25. Mai 2018 in Kraft.

Betrachten wir erst einmal die

Legaldefintionen nach Art. 4 Nr. 1

DSGVO und Art. 9 Abs. 1 DSGVO.

Art. 4 Nr. 1 DSGVO:

Personenbezogene Daten sind alle

Informationen, die sich auf eine

identifizierte oder identifizierbare

natürliche Person beziehen:

Beispiel:

Name, Staatsangehörigkeit, vertragliche

Beziehung zu Dritten, Beruf,

Telefonnummer, E-Mail-Adresse,

Beruf, Vermögen

Art. 9 Abs. 1 DSGVO:

Besondere personenbezogene Daten

sind Angaben über: die rassische

oder ethnische Herkunft, politische

Meinungen, religiöse oder weltanschauliche

Überzeugungen,

Gewerkschaftszugehörigkeit, Daten

zum Sexualleben oder Daten zur

sexuellen Orientierung

Diese Daten werden in besonderem

Umfang geschützt. Auch gelten

gemäß Art. 9 Abs. 2 DSGVO strengere

Maßstäbe für deren Handhabung.

Betriebs- und Geschäftsgeheimnisse

unterliegen per se erst einmal

nicht dem klassischen Datenschutz,

sondern sind anderweitig geschützt.

Hierunter fallen:

• vertraglich vereinbarte Geheimhaltungspflichten

• gesetzliche Geheimhaltungspflichten

(z.B. § 30 AO, § 35 SGB I)

• standesrechtliche Geheimhaltungspflichten

(Arztgeheimnis,

Schweigepflicht der Rechtsanwälte)

Zu berücksichtigen sind hierbei

auch die beiden Fallkonstellationen

pseudonymisierte Daten (gemäß Art.

4 Nr. 5 DSGVO) und anonymisierte

Daten gemäß ErwGr 26 Satz 5.

Bei pseudonymisierten Daten

wird der Personenbezug dadurch erschwert,

indem das Identitätsmerkmal

durch ein Kennzeichnen ersetzt

wird. Es handelt sich hier weiterhin

um personenbezogene Daten mit der

Folge, dass das Datenschutzrecht

anwendbar bleibt.

Bei anonymisierten Daten handelt

es sich um Daten, die einer Person

nicht mehr zugeordnet werden

können, zum Beispiel 12345678 statt

Heinz Becker. Es handelt sich nicht

mehr um personenbezogene Daten

und das Datenschutzrecht ist auch

nicht mehr anwendbar.

Was ist zulässige Verarbeitung

im Sinne der DSGVO?

Gemäß Artikel 2 Nr. 2 DSGVO versteht

man unter Verarbeitung jeden Vorgang

oder jede Vorgangsreihe mit

personenbezogenen Daten, die mit

oder ohne Hilfe automatisierter Verfahren

ausgeführt wird. Unterfälle

des Verarbeitens sind das Erheben,

das Speichern, das Übermitteln, die

Einschränkung und das Löschen.

ERHEBEN ist das aktive Beschaffen

von Daten über die betroffene

Person (z.B. Befragung des Betroffenen

oder Befragung bei Dritten).

Der Zweck der Verarbeitung muss

beim Erheben geklärt sein; Erheben

löst Informationspflichten nach Art.

13 f. DSGVO aus.

Unter SPEICHERN versteht man

das Erfassen, Aufnehmen, Aufbewahren

personenbezogener Daten auf

einem Datenträger zur weiteren Verarbeitung

oder Nutzung, wobei auch

Papier ein Datenträger sein kann. Die

Aufbewahrungsdauer (Speicherdauer),

die Zugriffsrechte müssen festgelegt

sein und es muss gemäß Art. 15

DSGVO Auskunft über gespeicherte

Daten gegeben werden.

ÜBERMITTELN ist das Bekanntgeben

gespeicherter personenbezogener

Daten an Dritte, wobei auch

das Abrufen hierunter zu subsumieren

ist. Die Übermittlung ist dabei

an Zulässigkeitsvoraussetzungen geknüpft,

wobei die Übermittlung in das

Nicht-EU-Ausland besonders abzusichern

ist.

Unter EINSCHRÄNKEN versteht

man das Kennzeichnen gespeicherter

personenbezogener Daten, um ihre

weitere Verarbeitung / Nutzung einzuschränken.

LÖSCHEN ist das Unkenntlichmachen

gespeicherter personenbezogener

Daten. Wenn Dritte mit der

Löschung beauftragt werden, so ist

hierbei auch Art. 28 DSGVO (Auftrags

verarbeitung) zu berücksichtigen.

Des Weiteren wird in der Regel

die DIN 66399 zu berücksichtigen

sein.

Gemäß Artikel 6 DSGVO ist die

Erhebung, Verarbeitung oder Nutzung

personenbezogener Daten verboten,

es sei denn, dass eine Erlaubnis vorliegt.

Die Zulässigkeit der Verarbeitung

nach Art. 6 Abs. 1 DSGVO kann

gegeben sein durch:

• Einwilligung

• Vertrag oder vorvertragliche Maßnahmen

• Rechtliche Verpflichtung des

Verantwortlichen

• Schutz lebenswichtiger Interessen

• Wahrnehmung einer Aufgabe im

öffentlichen Interesse bzw. Ausübung

öffentlicher Gewalt

• Überwiegende Interessen des Verantwortlichen

Die Struktur des neuen

Bundesdatenschutzgesetzes

Das neue Bundesdatenschutzgesetz

gliedert sich wie folgt:

Teil 1: Gemeinsame Bestimmungen

§§ 1 bis 21

Teil 2: Durchführungsbestimmungen

DSGVO, §§ 22 bis 44

Teil 3: Verarbeitungen gemäß JI-RL,

RL (EU) 2016/680, §§ 45 bis

84

Teil 4: Besondere Bestimmungen § 85

Bevor wir uns die Details anschauen,

sei zuerst einmal verwiesen, wo

welche Anforderungen zu finden sind:

Zulässigkeit der Verarbeitung:

§ 3 BDSG n.F.

Videoüberwachung:

§ 4 BDSG n.F.

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atw Vol. 63 (2018) | Issue 5 ı May

Verarbeitung besonderer personenbezoge-ner

Daten:

§ 22 BDSG n.F.

Übermittlung

§ 25 BDSG n.F.

Beschäftigungsdaten

§ 26 BDSG n.F.

Technisch-organisatorische Maßnahmen

Art. 24,32 DSGVO

Datensparsamkeit/Datenverarbeitung

Art. 25 DSGVO

Auftragsverarbeitung

Art. 28 DSGVO

Verarbeitungsverzeichnis

Art. 30 DSGVO

Datenschutz-Folgeabschätzung

(DSFA)

Art. 35 DSGVO

Datenschutzbeauftragte/r

Art. 37-39 DSGVO sowie

§§ 5-7, 38 BDSG n.F.

Die/der Datenschutzbeauftragte/r

Ein Datenschutzbeauftragter ist

gemäß Art. 37 Abs. 1 a-c DSGVO zu

bestellen, wenn einer der nachfolgenden

Voraussetzungen erfüllt ist:

• Behörde oder öffentliche Stelle

(Ausnahme: Gerichte)

• Kerntätigkeit mit umfangreicher

oder systematischer Überwachung

von Personen

• Kerntätigkeit mit umfangreicher

Verarbeitung besonders sensibler

Daten im Sinne von Art. 9 f. DSGVO

Auf die nachfolgenden Aspekte ist im

Sinne der DSGVO zu verweisen:

Art. 37 Abs. 2 DSGVO:

Eine Unternehmensgruppe darf einen

gemeinsamen Datenschutzbeauftragten

benennen, sofern von jeder

Niederlassung aus der Datenschutzbeauftragte

leicht erreicht werden

kann.

Art. 37 Abs. 6 DSGVO:

Der Datenschutzbeauftragte kann

Beschäftigter des Verantwortlichen

oder des Auftragsverarbeiters sein

oder seine Aufgaben auf der Grundlage

eines Dienstleistungsvertrags

erfüllen.

Art 37 Abs. 7 DSGVO:

Der Verantwortliche oder der Auftragsverarbeiter

veröffentlicht die

Kontaktdaten des Datenschutzbeauftragten

und teilt diese Daten der

Aufsichtsbehörde mit.

Wie ein Vergleich mit Art. 13 Abs. 1

a DSGVO zeigt – wo von Name und

Kontaktdaten die Rede ist – setzt die

Angabe der bloßen Kontaktdaten, z.B.

auf der Homepage nicht voraus, dass

auch der Name des Datenschutzbeauftragten

genannt wird. Gemäß

Art. 30 Abs.1 a in Verbindung mit

Art. 4 DSGVO ist die namentliche

Nennung des Beauftragten im Verhältnis

zur Aufsichtsbehörde jedoch

sinnvoll.

Die Bundesrepublik Deutschland

hat in Sachen Datenschutzbeauftragter

von der DSGVO Öffnungsklausel

Gebrauch gemacht und durch

§ 38 BDSG n.F. eine Pflicht zur Bestellung

eines Datenschutzbeauftragten

festgelegt, wenn:

1. mindestens zehn Personen ständig

mit automatisierter Verarbeitung

befasst sind oder

2. der Verantwortliche Verarbeitungen

vornimmt, die der Datenschutzfolgeabschätzung

nach Art. 35

DSGVO unterliegen oder

3. der Verantwortliche Daten gewerbsmäßig

zum Zweck der Übermittlung,

Markt- und Meinungsforschung

verarbeitet.

§ 38 Abs.2 BDSG verweist dabei auf

wichtige Grundregeln:

• Abberufungsschutz, § 6 Abs. 4

BDSG n.F.

• Verschwiegenheitspflicht § 6 Abs.

5 Satz 2 BDSG n.F.

• Zeugnisverweigerungsrecht, § 6

Abs. 6 BDSG n.F.

Der Datenschutzbeauftragte muss ein

Fachwissen mitbringen und persönlich

geeignet sein. Nach Art. 37

Abs. 5 DSGVO muss ein Fachwissen

auf dem Gebiet des Datenschutzrechtes

und der Datenschutzpraxis

ebenso vorhanden sein wie die Fähigkeit

zur Erfüllung der in Art. 39

DSGVO genannten Aufgaben. Dies

Trias aus rechtlichen, technischen

und organisatorischen Kenntnissen ist

jedoch nicht mehr zwingend nachzuweisen.

Die Mindestaufgaben des Datenschutzbeauftragten

sind durch

Artikel 39 Abs. 1 DSGVO wie folgt

definiert:

• Unterrichtung und Beratung des

Verantwortlichen oder des Auftragsverarbeiters

und der Beschäftigten,

die Verarbeitungen durchführen,

hinsichtlich ihrer Pflichten

nach dieser Verordnung und den

sonstigen Datenschutzvorschriften

der Union bzw. der Mitgliedsstaaten

• Überwachung der Einhaltung

dieser Verordnung, anderer Datenschutzvorschriften

der Union bzw.

der Mitgliedsstaaten sowie der

Strategien der Verantwortlichen

oder des Auftragsverarbeiters für

den Schutz personenbezogener

Daten inkl. der Zuweisung von

Zuständigkeiten, der Sensibilisierung

und Schulung der an den Verarbeitungsvorgängen

beteiligten

Mitarbeiter und der diesbezüglichen

Überprüfungen

• Auf Anfrage Beratung im Zusammenhang

mit der Datenschutzfolgeabschätzung

und Überwachung

Ihrer Durchführung gemäß

Artikel 35 DSGVO

• Zusammenarbeit mit der Aufsichtsbehörde

• Tätigkeit als Anlaufstelle für die

Aufsichtsbehörde in mit der Verarbeitung

zusammenhängenden

Fragen, inklusive der vorherigen

Konsultation gemäß Artikel 36

DSGVO sowie gegebenenfalls

Beratung zu allen sonstigen Fragen

Gemäß Artikel 39 Abs. 2 DSGVO hat

der Datenschutzbeauftragte bei der

Erfüllung seiner Aufgaben dem mit

den Verarbeitungsvorgängen verbundenen

Risiken Rechnung zu

tragen. Art, Umfang, Umstände sowie

die Zwecke der Verarbeitung sind im

Rahmen der Pflicht zur risikoorientierten

Tätigkeit zu berücksichtigen.

Auf die Strafbarkeit gemäß § 203

Abs. 4 StGB für Datenschutzbeauftragte

wird explizit verwiesen.

Für die Zusammenarbeit zwischen

Datenschutzbeauftragten und Betriebsrat

gibt es keinerlei gesetzliche

Regelungen. So hat der Betriebsrat

zum Beispiel auch kein Mitspracherecht

bei der Bestellung. Der Betriebsrat

ist selbst auch zur Einhaltung des

Datenschutzes der ihm anvertrauten

Daten verpflichtet. In diesem Zusammenhang

wird auf die Grundsatzentscheidung

des Bundesarbeitsgerichtes

BAG verwiesen, wonach dieser ein

Kontrollrecht des Datenschutzbeauftragten

bei Datenverarbeitung durch

den Betriebs-/Personalrat ablehnt

(NJW 1998, S. 2466). Nach § 75 Abs.

3 Nr. 17 BPersVG, § 87 Abs. 1 Nr. 6

BetrVG hat der Betriebs-/Personalrat

ein Mitbestimmungsrecht bei der Einführung

technischer Einrichtungen,

die dazu bestimmt sind, das Verhalten

oder Leistung der Beschäftigten zu

überwachen (z.B. Zeiterfassung, Protokolldaten,

Telefondatenerfassung).

In Sachen Personalbögen hat der

Betriebs-/Personalrat (§ 94 BetrVG,

§ 75 Abs.3 Nr. 8 BPersVG) nur bei der

Erhebung ein Mitbestimmungsrecht,

für die Speicherung und Verarbeitung

existieren nämlich bereits gesetzliche

Grundlagen. Bezüglich der Durchführung

von Schulungen wird in diesem

Zusammenhang auf § 96 BetrVG

sowie § 75 Abs. 3 Nr. 7, § 76 Abs. 2 Nr.

6 BPersVG verwiesen. Eine Betriebsvereinbarung

zum Datenschutz mag

wünschenswert sein, gleichwohl ist

diese nicht verpflichtend.

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ENERGY POLICY, ECONOMY AND LAW 292

Die Aufsichtsbehörde

Alleine in den Artikel 51 bis 76 der

DSGVO finden sich 26 Vorschriften zu

den Aufsichtsbehörden. Als wichtigste

Bereiche sind zu nennen:

• Aufgaben und Befugnisse,

Art. 57 f. DSGVO

• Zusammenarbeit, Art. 60-62

DSGVO

• Kohärenzverfahren, Art. 63-67

DSGVO

• Europäischer Datenschutzausschuss,

Art. 68ff. DSGVO

• Sanktionen, Art. 83 DSGVO

Als Aufgaben sind gemäß Art. 57ff.

DSGVO zu nennen:

• Kontrolle

• Beratung

• Bearbeitung von Eingaben von

Bürgern

• Information der Bürger

Befugnisse der Aufsichtsbehörde sind:

• Abberufungsrecht:

Der Datenschutzbeauftragte kann

von der Aufsichtsbehörde in

begründeten Fällen abberufen

werden.

• Auskunft zu Fragen sowie in

Einsicht in alle Unterlagen, vor

allem in die gespeicherten Daten

und in die Datenverarbeitungsprogramme

• Jederzeit Zutritt zu allen Räumen

Kontrollarten sind:

• anlassbezogene Kontrollen, z.B.:

im Rahmen der Eingabe von

Bürgern

• themenbezogene Kontrollen, z.B.

Prüfung / Lösung grundsätzlicher

Probleme

• Querschnittskontrolle, z.B. zum

Kennenlernen einer Branche

Die Aufsichtsbehörde kann bei Verstößen

gegen den Datenschutz Geldbußen

verhängen, wobei Verstöße

gegen die DSGVO „wehtun“ sollen.

Gemäß Art. 79 Abs. 3a DSGVO können

die Geldbußen bis zu 20 Millionen

Euro betragen. Berechnungsgrundlage

ist dabei der Umsatz des Vorjahres

(4 Prozent Regel).

Gemäß § 40 Abs. 1 BDSG n.F. überwachen

die nach Landesrecht zuständigen

Behörden die nichtöffentlichen

Stellen. Dies sind in der Regel die Landesdatenschutzbeauftragten.

Betroffenenrechte

Als Betroffenenrechte sind gemäß des

BDSG n.F. zu nennen:

• § 32:

Informationspflicht bei Erhebung

von personenbezogenen Daten bei

der betroffenen Person

• § 33:

Informationspflicht, wenn die

personenbezogenen Daten nicht

bei der betroffenen Person

erhoben wurden

• § 34:

Auskunftsrecht der betroffenen

Person

• § 35:

Recht auf Löschung

• § 36:

Widerspruchsrecht

• § 37:

Automatisierte Entscheidungen im

Einzelfall einschließlich Profiling

Gemäß Art. 15 DSGVO/§ 34 BDSG

besteht ein Recht auf Auskunft, ob

der Verantwortliche personenbezogene

Daten verarbeitet hat oder nicht. Das

Recht umfasst folgende Informa tionen:

• Verarbeitungszwecke

• Kategorien von Daten

• Empfänger der Daten

• Dauer der Speicherung oder

Kriterien zu deren Festlegung

• Bestehen eines Berichtigungs-,

Löschungs-, Widerspruchs-,

Beschwerderechts

• Herkunft der Daten

Auskunftserteilungen müssen nach

Art. 12 Abs. 3 DSGVO unverzüglich,

spätestens aber innerhalb eines

Monates erfolgen. Nur in begründeten

Ausnahmefällen -über die die betroffene

Person aber zu informieren

istdarf die Monatsfrist überschritten

werden. Nach Art. 12 Abs. 5 S. 3

DSGVO muss die Auskunftserteilung

für die Erstauskunft kostenfrei erfolgen,

wobei für weitere Kopien ein

„angemessenes“ Entgelt verlangt

werden kann.

Im Rahmen von Art. 16 DSGVO

haben Betroffene ein Recht auf

Berichtigung oder Vervollständigung

der Daten.

Das Recht auf Vergessenwerden/

Löschen ist durch Art. 17 DSGVO/

§ 35 BDSG n.F. normiert. Ein Verarbeiter

muss personenbezogene

Daten löschen, wenn:

1. sie für den Erhebungszweck nicht

mehr erforderlich sind, oder

2. die Einwilligung zur Verarbeitung

widerrufen wird, oder

3. die Verarbeitung unrechtmäßig ist

Ein Folgenbeseitigungsanspruch besteht,

wenn die Daten durch den

Verantwortlichen veröffentlicht wurden.

Bei der Erfüllung einer rechtlichen

Verpflichtung oder zur Wahrnehmung

einer Aufgabe im öffentlichen

Interesse oder in Ausübung

öffentlicher Gewalt ist das Recht auf

Löschung nicht gegeben.

Haftungsfragen nach dem

neuen Datenschutzrecht

Nach Art. 82 DSGVO haften

Unternehmer und Auftraggeber für

materiellen und immateriellen

Schaden, welcher aufgrund eines

Verstoßes gegen die DSGVO entstanden

ist. Ein Auftragsverarbeiter

haftet (auch), wenn er die Weisung

des Auftraggebers nicht beachtet.

Eine Haftung entfällt in der Regel nur

dann, wenn der für die Verarbeitung

Verantwortliche nachweisen kann,

dass er nicht für den Schaden verantwortlich

ist. Im Rahmen der Haftung

ist im Übrigen eine gesamt-schuldnerische

Haftung gegeben.

Gemäß § 831 BGB (Haftung aus

Deliktrecht) haftet ein Unternehmen

für Handlungen seiner Mitarbeiter,

die es als Verrichtungsgehilfen ausgewählt

hat, es sei denn es liegt eine

Exkulpierung dergestalt vor, dass der

Mitarbeiter sorgfältig ausgewählt und

überwacht wurde. Die Exkulpationsmöglichkeit

greift jedoch nicht bei

Organen eines Unternehmens

(§§ 30 f. BGB).

Eine Haftung aus Deliktrecht

(§ 823 BGB) ist gegeben, wenn ein

Eingriff in das Persönlichkeitsrecht

des Betroffenen vorliegt. Obgleich

sich der Anspruch gegen den Mitarbeiter

wendet, der die Handlung

begangen hat, hat der Mitarbeiter in

der Regel aber nach den Grundsätzen

des innerbetrieblichen Schadensausgleiches

einen Anspruch auf vollständige

oder teilweise Freistellung

durch das Unternehmen.

In Artikel 83 DSGVO finden sich

die Datenschutzverstöße, welche zu

einer Geldbuße führen können, die

von der Aufsichtsbehörde gemäß

Art. 41 BDSG festgesetzt werden.

Die Strafvorschrift ist durch § 42

BDSG gegeben. Hiernach wird mit

einer Freiheitsstrafe bis zu 2 Jahren

oder Geldstrafe bestraft, wer allgemein

nicht zugängliche Daten ohne

Berechtigung verarbeitet oder unter

falschen Angaben erschleicht und dies

gegen Entgelt oder mit Bereicherungsoder

Schädigungsabsicht tut. Die Tat

wird jedoch nur auf Antrag von

Betroffenen, Verantwortlichen oder

der Aufsichtsbehörde verfolgt.

Auch der Datenschutzbeauftragte

kann strafrechtlich belangt werden,

wen er für einen Bereich zuständig

ist, in dem gemäß § 203 StGB Verschwiegenheitspflichten

gelten und

er gleichwohl unbefugt Daten übermittelt,

verarbeitet oder erschleicht.

Neben § 203 StGB ist hier auch der

§ 42 BDSG in Betracht zu ziehen.

Auftragsdatenverarbeitung

Nach Art. 4 Nr. 8 DSGVO ist Auftragsverarbeiter

eine Stelle, welche personenbezogene

Daten im Auftrag des

Energy Policy, Economy and Law

General Data Protection Regulation (GDPR) of the European Union – What Had to Be Considered until 25 May 2018 ı Stefan Loubichi


atw Vol. 63 (2018) | Issue 5 ı May

Verantwortlichen verarbeitet. Verantwortlicher

nach Art. 4 Nr. 7 DSGVO

ist die Stelle, die allein oder gemeinsam

mit Dritten über die Mittel

und die Zwecke der Verarbeitung

personen bezogener Daten entscheidet.

Für die Weitergabe personenbezogener

Daten an den Auftragsverarbeiter

und die Verarbeitung durch

den Auftragsverarbeiter bedarf es in

der Regel keiner weiteren Rechtsgrundlage

gemäß Art. 6 bis 10

DSGVO. Gleichwohl muss wie bisher

zwischen den beiden Vertragsparteien

ein schriftlicher oder in einem

elektronischen Format abgefasster

Vertrag vorliegen, wobei die Gesamtverantwortung

für die Datenverarbeitung

und die Nachweispflicht

gemäß Art. 5 Abs. 2 DSGVO nach wie

vor beim Verantwortlichen verbleibt.

Nur wenn der Auftragsverarbeiter die

zu verarbeitenden Daten vertragswidrig

verarbeitet, gilt er nach Art. 28

Abs. DSGVO selbst als Verantwortlicher

und muss dann alle rechtlichen

Folgen tragen. Nach der neuen

DSGVO muss der Auftragsverarbeiter

selbst ein Verzeichnis von Verarbeitungstätigkeiten

nach Art. 30 Abs. 2

DSGVO führen. Verstoßen Auftragsverarbeiter

gegen datenschutzrechtliche

Bestimmungen, so können

gemäß Art. 28 DSGVO Geldbußen von

bis zu 10.000.000 Euro oder bis zu

2 % des gesamten weltweiten Jahresumsatzes

des vergangenen Jahres

von der Aufsichtsbehörde verhängt

werden.

Gemäß Anhang A Art. 28 DSGVO

liegt eine Auftragsverarbeitung auch

in folgenden Fällen vor:

• DV-technische Arbeiten für die

Lohn- und Gehaltsabrechnung

oder die Finanzbuchhaltung durch

Rechenzentren

• Auslagerung der Backup-Sicherheitsspeicherung

und anderer

Archivierungen

• Datenträgerentsorgung durch

Dienstleister

• Fernwartung, wenn bei diesen

Tätigkeiten ein Zugriff auf personenbezogene

Daten nicht ausgeschlossen

werden kann

• Zentralisierung bestimmter Shared

Service Dienstleistungen

• Outsourcing personenbezogener

Datenverarbeitung im Rahmen

von Cloud Computing

Keine Auftragsverarbeitung liegt

gemäß Anhang B Art. 28 DSGVO bei

Berufsgeheimnisträgern wie Steuerberater,

Rechtsanwälte, externe

Betriebsärzte oder Wirtschaftsprüfern

vor.

Verzeichnis der

Verarbeitungstätigkeiten

Organisation mir weniger als 250

Mitarbeitenden müssen gemäß Art.

30 Abs. 5 DSGVO kein Verzeichnis von

Verarbeitungstätigkeiten führen. Diese

Begrenzung gilt jedoch dann nicht,

wenn eine Verarbeitung personenbezogener

Daten durchgeführt wird,

• welche Risiken für die Rechten und

Pflichten der betroffenen Personen

birgt oder

• die nicht nur gelegentlich erfolgt

oder

• die besondere Datenkategorien

gemäß Art. 9 Abs. 1 DSGVO oder

strafrechtliche Verurteilungen/

Straftaten nach Art. 19 DSGVO

betreffen

Nach dem neuen Datenschutzrecht ist

weder ein öffentliches Verzeichnis

noch eine Meldepflicht mehr gegeben,

wie dies früher war.

Während die Vorgaben für Verzeichnisse

für Verantwortliche durch

Art. 30 Abs. 1 DSGVO bestimmt sind,

sind die Anforderungen für den Inhalt

eines Verzeichnisses für Auftragsverarbeiter

durch Art. 30 Abs. 2 DSGVO

definiert.

Sowohl für Verarbeiter als auch für

Auftragsverarbeiter müssen in den

Verzeichnissen allgemeine Beschreibungen

der technischen und organisatorischen

Maßnahmen zu finden sein,

wobei nicht definiert ist, wie detailliert

die Beschreibung sein muss.

Hierzu wird auf Art. 32 DSGVO verwiesen.

Neben den Verzeichnissen der

Verarbeitungstätigkeiten gibt es

noch weitere Dokumentationspflichten,

welche zu berücksichtigen sind,

so zum Beispiel:

• das Vorhandensein von Einwilligen

(nach Art. 7 Abs. 1 DSGVO)

• die Ordnungsmäßigkeit der gesamten

Verarbeitung (Art. 24 Abs.

1 DSGVO)

• das Ergebnis von Datenschutzfolgeabschätzungen

(Art. 35 Abs. 7

DSGVO)

Datenschutzfolgenabschätzung

Die Datenschutzfolgeabschätzung

(DSFA) nach Art. 35 DSGVO ist

ein Instrument zur Beschreibung,

Bewertung und Eindämmung von

Risiken für die Rechte und Freiheiten

natürlicher Personen bei der Verarbeitung

personenbezogener Daten.

Im Rahmen der DSFA muss der

Verantwortliche Verarbeitungen mit

hohem Risiko überprüfen, bevor

sie beginnen. Hierdurch soll sichergestellt

sein, dass angemessene

Schutzmaßnahmen vor dem hohen

Risiko vorgesehen sind.

Neben den in Art. 35 Abs. 3 DSGVO

genannten Gründen muss eine DSFA

in den folgenden Fällen durchgeführt

werden:

1. nach ErwGr 91, wenn Verfahren

eingesetzt werden,

• die nach Auffassung der zuständigen

Aufsichtsbehörde

wahrscheinlich ein hohes Risiko

für Freiheiten und Rechte der

betroffenen Personen mit sich

bringen

• bei denen den Betroffenen

die Ausübung ihrer Rechte erschwert

wird

2. nach Leitfaden WP 248 der Art. 29

– Gruppe in den folgenden Fällen:

• Bewertung und Scoring inkl.

Prognosen und Profilerstellung

• automatisch erfolgende Entscheidungen

mit rechtlichen

oder vergleichbaren Auswirkungen

für Dritte

• systematisches Monitoring

• sensitive, vor allem personenbezogene

Daten

• umfangreiche Datenmengen

• Vergleich oder Kombination

von Datensätzen

• Daten ungeschützter Betroffener

• Einsatz innovativer Technologien

oder neuartiger organisatorischer

Lösungen

• Datentransfers in Länder außerhalb

der EU/EWR

• Verhinderung, dass die betroffene

Person ein Recht ausüben

kann

Die Aufsichtsbehörden sind gemäß

Art. 35 Abs. 4 DSGVO verpflichtet,

eine Positiv-Liste für diejenigen

Verfahren zu erstellen, bei denen auf

jeden Fall eine DSFA durchzuführen

ist. Sie kann eine Negativ-Liste für

Verarbeitungen erstellen, für die keine

DSFA durchzuführen ist.

Der Mindestinhalt der Prüfung ist

durch Art. 35 Abs. 7 DSGVO normiert:

1. Systematische Beschreibung der

geplanten Verarbeitungen und der

zugehörigen Verarbeitungszwecke

2. Bewertung der Notwendigkeit und

Verhältnismäßigkeit der Verarbeitungsvorgänge

3. Bewertung der Risiken für die

Rechte und Freiheiten der betroffenen

Personen

4. Bewertung der zur Bewältigung

der Risiken geplanten Abhilfemaßnahmen

Kommt die DSFA zu dem Ergebnis,

dass die Verarbeitung unzulässig ist,

kann der Verantwortliche sich zur

„vorherigen Konsultation“ nach Art.

ENERGY POLICY, ECONOMY AND LAW 293

Energy Policy, Economy and Law

General Data Protection Regulation (GDPR) of the European Union – What Had to Be Considered until 25 May 2018 ı Stefan Loubichi


atw Vol. 63 (2018) | Issue 5 ı May

ENERGY POLICY, ECONOMY AND LAW 294

36 an die für ihn zuständige Landesbehörde

wenden.

Diese prüft, berät und entscheidet

innerhalb von 8 Wochen über die

Zulässigkeit. Die Datenschutzfolgeabschätzung

wird vom Verantwortlichen

durchgeführt, d.h. die Leitung

muss die Aufgabe einer Stelle zur

Erledigung zuweisen. Gemäß Art. 39

Abs. 1 c) DSGVO hat der Datenschutzbeauftragte

hierbei „nur“ Beratungsfunktion.

Meldepflicht

bei Datenpannen

Als Datenpanne im Sinne von Art.

33 f. DSGVO erachtet man die Verletzung

des Schutzes personenbezogener

Daten

• aufgrund Vernichtung oder Veränderung

• aufgrund Zugang oder Offenlegung

• oder in sonstiger Weise

Bei Datenpannen ist eine Risikoeinschätzung

vorzunehmen, ob ein

Risiko für die Rechte und Freiheiten

der betroffenen Personen besteht. Es

besteht in der Regel gegenüber der

Aufsichtsbehörde und den Betroffenen

eine zeitlich definierte Informationspflicht.

Technisch-organisatorische

Maßnahmen

Aus dem Prinzip der Integrität und

Vertraulichkeit gemäß Art. 5 Abs. 1

DSGVO werden die technisch-organisatorischen

Maßnahmen (TOM)

abgeleitet die bewirken (sollen), dass

das Ziel des Datenschutzes erreicht

wird.

Die TOM müssen im Rahmen einer

Dokumentation nachgewiesen als

auch überprüft und aktualisiert

werden. Gemäß Artikel 25 DSGVO

müssen die Datenschutzgründe bereits

bei der Festlegung berücksichtigt

werden und es müssen restriktive

Voreinstellungen bzgl. des Umfangs,

der Zugriffsrechte und Speicherdauer

der personenbezogenen Daten vorliegen.

In Art. 32 Abs. 1 a) bis d) DSGVO

werden exemplarisch nachfolgende

Maßnahmen aufgeführt:

• Pseudonymisierung und Verschlüsselung

• Vertraulichkeit, Integrität, Verfüg-barkeit

und Belastbarkeit der

Verarbeitungssysteme

• Wiederherstellung der Verfügbarkeit

und des Zuganges nach einem

Zwischenfall

• Verfahren zur regelmäßigen

Evaluierung der getroffen TOM

Telekommunikationsgeheimnis

Der Schutzbereich des Fernmeldegeheimnisses

ist der Schutz der

unkörperlichen Übermittlung individueller

Kommunikation, unabhängig

von der Übertragungstechnik und

unabhängig vom Inhalt.

Während für staatliche Stellen in

der Regel Art. 10 GG die Normierungsgrundlage

ist, ist die Anspruchsgrundlage

für Dienstanbieter im

nicht-öffentlichen Bereich § 88 TKG.

Diesbezüglicher Dienstanbieter ist

jeder, der ganz oder teilweise geschäftsmäßig

TK-Dienste erbringt

oder bei der Erbringung solcher mitwirkt,

§ 3 Nr. 6 TKG. Dadurch, dass ein

Unternehmen die private Nutzung

von Telekommunikationsdiensten zulässt,

wird es zum Diensteanbieter

im Sinne des Telekommunikationsgesetzes

(TKG).

Hiernach geschützt sind:

1. Inhalt der Telekommunikation

2. Nähere Umstände der Telekommunikation

3. Beteiligte am Telekommunikationsvorgang

4. Erfolglose Verbindungsversuche

und damit zusammenhängende

Verkehrsdaten

In der Regel verboten ist die Kenntnisnahme

des Inhaltes oder der näheren

Umstände der Kommunikation. In

§ 96 TKG ist geregelt, welche Verkehrsdaten

erhoben werden dürfen.

Neben dem TKG sind in diesem

Zusammenhang auch zu beachten:

• § 206 StGB:

Verletzung des Post- oder Fernmeldegeheimnisses

• § 201 StGB:

Verletzung der Vertraulichkeit des

Wortes

Datenübermittlung

ins Ausland

Die Datenschutzgrundverordnung befasst

sich in den Artikeln 44 bis 49 mit

der Datenübermittlung an Länder

außerhalb der EU/des EWR. Diese

Länder werden als so genannte „Drittstaaten“

bezeichnet.

Es erfolgt hier eine Zwei-Stufen­

Prüfung.

Stufe 1:

Halten die Drittländer neben den

in Art. 45 ff. DSGVO spezifischen

Anforderungen alle übrigen Anforderungen

der DSGVO (z.B. auch Art. 9

Abs. 3 DSGVO) ein?

Stufe 2:

Werden die in Artikel 45 ff. genannten

Anforderungen erfüllt?

Für nichtöffentliche Stellen gibt es

folgende Möglichkeiten des Datentransfers

in Drittländer:

• Feststellung der Angemessenheit

des Datenschutzniveaus im Drittland

durch die EU-Kommission

(Art. 45 DSGVO)

• Vorliegen geeigneter Garantien

(Art. 46 DSGVO)

• Ausnahmen für bestimmte Fälle

(Art. 49 DSGVO)

Fazit

Der vorstehende Aufsatz mag hoffentlich

aufgezeigt haben, dass die DSVGO

mehr als nur eine redaktionelle

Anpassung war und dass hier ein

Paradigmenwechsel

hat.

stattgefunden

Hinweis

Ausdrücklich wird in diesem Zusammenhang

darauf verwiesen, dass

dieser wissenschaftliche Artikel keine

Beratung in Rechtsfragen ersetzen

kann und auch nicht als Rechtsberatung

angedacht ist. Dieser Fachartikel

kann auch keine unternehmensspezifischen

Fragen beantworten, sondern

liefert nur allgemeine Aussagen zu

den Neuerungen des Datenschutzes

durch die EU Datenschutzgrundverordnung

und die Umsetzung durch

das ab dem 25.5.2018 gültige (neue)

Bundesdatenschutzgesetz.

Author

Prof. h.c.(IUK) PhDr. Dipl.-Kfm./

Dipl.-Vw. Stefan Loubichi

Loubichi Business Consulting UG

(haftungsbeschränkt)

Associate expert to Kraftwerksschule

Essen and Simulator Centre

Essen (GfS mbH / KSG mbH)

Grafenberger Allee 125

40237 Düsseldorf

Energy Policy, Economy and Law

General Data Protection Regulation (GDPR) of the European Union – What Had to Be Considered until 25 May 2018 ı Stefan Loubichi


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 Atomrecht

Atomrecht – Was Sie wissen müssen RA Dr. Christian Raetzke 12.06.2018 Berlin

Ihr Weg durch Genehmigungs- und Aufsichtsverfahren RA Dr. Christian Raetzke 18.09.2018 Berlin

Das Recht der radioaktiven Abfälle RA Dr. Christian Raetzke 23.10.2018 Berlin

3 Energie, Politik und Kommunikation

Schlüsselfaktor Interkulturelle Kompetenz –

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Public Hearing Workshop –

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Kerntechnik und Energiepolitik im gesellschaftlichen Diskurs

– Themen und Formate

Angela Lloyd 26.09.2018 Berlin

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N.N. 12.11. - 13.11.2018 Gronau/Lingen

3 Kerntechnik, Rückbau und Strahlenschutz

Export kerntechnischer Produkte und Dienstleistungen –

Chancen und Regularien

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

3 Nuclear English

Das neue Strahlenschutzgesetz –

Folgen für Recht und Praxis

Stilllegung, Rückbau und Entsorgung –

Recht und Praxis

RA Kay Höft, M.A.,

RA Olaf Kreuzer,

Dr. Wolfgang Steinwarz

RA Dr. Christian Raetzke,

Maria Poetsch

RA Dr. Christian Raetzke,

Dr. Matthias Bauerfeind

20.06. - 21.06.2018 Berlin

05.06. - 06.06.2018

27.06. - 28.06.2018

05.11. - 06.11.2018

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24.09. - 25.09.2018 Berlin

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3 Wissenstransfer und Veränderungsmanagement

Veränderungsprozesse gestalten – Heraus forderungen

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Erfolgreicher Wissenstransfer in der Kern technik –

Methoden und praktische Anwendung

Dr. Christien Zedler,

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28.11. - 29.11.2018 Berlin

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

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Die INFORUM-Seminare können je nach

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atw Vol. 63 (2018) | Issue 5 ı May

296

Das neue Strahlenschutzrecht und die Freigabe:

Alles neu macht der Mai?

Ulrike Feldmann

SPOTLIGHT ON NUCLEAR LAW

Erstmalig wurde im vergangenen Sommer ein Strahlenschutzgesetz in Deutschland aus der Taufe gehoben. Die

Bundesregierung hatte die Verpflichtung zur Umsetzung der revidierten Fassung der europäischen Strahlenschutzgrundnormen,

der Richtlinie 2013/59/Euratom, zum Anlass genommen, die Wichtigkeit des Strahlenschutzrechts

durch Hochzonen in den Gesetzesrang zu unterstreichen, womit das Strahlenschutzrecht nun gleichrangig neben dem

Atomgesetz steht. Dieses Gesetz gilt es nun auf Verordnungsebene mit „Leben“ zu erfüllen, um es für die Praxis

anwendbar zu machen. Die Frist zur Umsetzung der Richtlinie 2013/59/Euratom ist bereits am 6.02.2018 abgelaufen.

Das Strahlenschutzgesetz

Das Gesetz zur Neuordnung des Rechts zum Schutz vor der

schädlichen Wirkung ionisierender Strahlung vom 27.06.2017

(BGBl I S. 1966) ist als sog. Artikelgesetz konzipiert. Es

enthält in Artikel 1 das Gesetz zum Schutz vor der

schädlichen Wirkung ionisierender Strahlung (Strahlenschutzgesetz/StrlSchG).

Daneben enthält das Artikelgesetz

in 31 Artikeln Änderungen anderer Gesetze oder Verordnungen.

Während insbesondere die Notfallvorschriften

(§§ 92 – 116 StrlSchG) sowie die mehr als 40 Vorordnungsermächtigungen

bereits am 17.10.2017 in Kraft getreten

sind, treten die übrigen Bestimmungen des StrlSchG erst

am 31.12.2018 in Kraft. Zu diesem Zeitpunkt sollen dann

auch die für den Vollzug erforderlichen Vorschriften auf

Verordnungsebene in Kraft treten.

Das StrlSchG umfasst neben den Regelungsinhalten der

bisherigen Strahlenschutzverordnung (StrlSchV) und der

Röntgenverordnung (RöV) nunmehr alle Bereiche des

Schutzes vor ionisierender Strahlung (s. dazu weiter

unten). Darüber hinaus will das StrlSchG aktuellen

wissenschaftlichen Erkenntnissen Rechnung tragen.

Zudem wird der radiologische Notfallschutz vor dem

Hintergrund der Erfahrungen, die im Rahmen des Unfalls

in Fukushima gemacht wurden, insbesondere organisatorisch

durch die Einrichtung eines Notfallmanagementsystems

des Bundes und der Länder, deutlich verstärkt

(§§ 92 ff StrlSchG). Durch diese Erweiterungen wie

auch durch Vorgaben der Richtlinie 2013/59/Euratom

(im folgenden RL 2013/59) wurde eine Neustrukturierung

der Strahlenschutzvorschriften erforderlich, die für

die Praxis etwas gewöhnungsbedürftig sein wird.

Was bringt das StrlSchG ansonsten Neues?

Ohne Anspruch auf Vollständigkeit sei hier auf folgende

Neuerungen hingewiesen:

Zukünftig ist zwischen geplanten, bestehenden und

notfallbedingten Expositionssituationen zu unterscheiden

(§ 1 StrlSchG). Neben den Expositionssituationen hat die

RL 2013/59 den Radiation Protection Expert (RPE) sowie

den Radiation Protection Officer (RPO) neu eingeführt.

Beide Personen unterscheiden sich sowohl hinsichtlich

ihrer Stellung im Unternehmen als auch im Hinblick auf

ihre Qualifikation und Aufgaben. Die RL 2013/59 lässt es

jedoch zu, dass die Aufgaben des RPO auch von einem RPE

übernommen werden können. Da der deutsche Strahlenschutzbeauftragte

(§ 70 StrlSchG) innerhalb des ihm

zugewiesenen Aufgabenbereichs die Anforderungen an

den RPE wie auch den RPO erfüllt, braucht neben dem

Strahlenschutzbeauftragten (SSB) für seinen Aufgabenbereich

insoweit keine weitere Person bestellt zu werden.

Die Stellung des SSB im Unternehmen wird dafür

im StrlSchG gestärkt (s. § 70 Abs. 6 u. § 71 Abs. 2 S. 3

StrlSchG).

In diesem Zusammenhang sei darauf hingewiesen, dass

zukünftig auch für Nukleartransporte ein Strahlenschutzverantwortlicher

und ein SSB bestellt werden müssen

(§ 29 Abs. 1 Nr. 2 und 3 StrlSchG). Denn auch die

Beförderung stellt eine geplante Expositionssituation dar.

Diese ist eine Exposition, die durch Tätigkeiten entsteht

(§ 2 Abs. 2 StrlSchG). Da die Beförderung unter den Begriff

der Tätigkeiten fällt (§ 4 Abs. 1 Nr. 1 iVm. § 5 Nr. 39

StrlSchG), finden die Vorschriften für Tätigkeiten auch auf

die Beförderung Anwendung. Denkbar ist, dass beispielsweise

der Gefahrgutbeauftragte die Aufgabe des SSB

übernehmen kann. Dies setzt jedoch entsprechende

Fachkunde, in der Regel also eine spezielle „Nachqualifizierung“

des Gefahrgutbeauftragten, voraus.

Das neue Strahlenschutzrecht kennt im Übrigen nur

noch den Begriff der Tätigkeiten, wobei – ebenfalls neu –

ein Verfahren zur Prüfung der Rechtfertigung einer Tätigkeitsart

beschrieben wird (§ 7 StrlSchG). Der Begriff der

Arbeiten ist aufgrund der RL 2013/59 weggefallen.

Außerdem waren in das StrlSchG Regelungen zum

Dosisrichtwert aufzunehmen (s. § 79 Abs. 1 Nr. 2 und § 81

Satz 2 Nr. 9 StrlSchG).

Neu im StrlSchG ist zudem, dass bei der uneingeschränkten

Freigabe von Feststoffen im StrlSchG

aufgrund der Vorgaben der RL 2013/59 basierend auf der

IAEO-Empfehlung RS-G-1.7 im Grundsatz die Freigrenzen

identisch mit den Freigabewerten sind. Das Bundesministerium

für Umwelt, Naturschutz und nukleare Sicherheit

(BMU) betont jedoch, dass sich alle bisher genutzten

Freigabepfade bewährt haben und grundsätzlich auch im

Einklang mit der RL 2013/59 stehen.

Die RL 2013/59 erfordert ferner eine Absenkung

der spezifischen Freigrenzen. Es werden in der Praxis

jedoch eher nur geringe Auswirkungen auf bestehende

Genehmigungen erwartet. Die Freigrenzen der Gesamtaktivität

bleiben im Übrigen erhalten. Die Verordnungsermächtigung

zur Regelung, welche Werte der Aktivität

und spezifischen Aktivität radioaktiver Stoffe als Freigrenzen

gelten, findet sich in § 24 Satz 1 Nr. 10 StrlSchG.

Des weiteren findet sich im StrlSchG als neue Regelung

ein Referenzwert zum (natürlich vorkommenden)

Radon an Arbeitsplätzen und in Aufenthaltsräumen von

300 Bq/m 3 für Gebiete mit erhöhtem Radonpotential

(§§ 124 u. 126 StrlSchG).

Ferner enthält das StrlSchG gegenüber der bisherigen

Strahlenschutzverordnung u.a. Regelungen zum Schutz

des raumfahrenden Personals, (§ 1 Abs. 2 Nr. 1, § 2 Abs. 7

Nr. 2, § 4 Abs. 1 Nr. 11 und §§ 52 – 54 StrlSchG), zu Tätigkeiten

im Zusammenhang mit natürlich vorkommender

Radioaktivität (§§ 55 – 66 StrlSchG), zu Radioaktivität in

Bauprodukten (§§ 133 -135 StrlSchG), zu radioaktiven

Altlasten (§§ 136 – 150 StrlSchG), zur Einführung eines

Informations- und Meldesystems im medizinischen

Bereich (die generelle Verordnungsermächtigung enthält

Spotlight on Nuclear Law

The New Radiation Protection Law and the Approval: May Makes Everything New? ı Ulrike Feldmann


atw Vol. 63 (2018) | Issue 5 ı May

§ 90 StrlSchG, s. auch Begründung zu § 90 StrlSchG in

Bundesrats-Drucksache 86/17, S. 401) und zur Stärkung

der Rolle des Medizinphysik-Experten (s. § 14 StrlSchG).

Neu ist ebenfalls, dass der Dosisgrenzwert für Einzelpersonen

der Bevölkerung (1 mSv/a) sich nunmehr auf

die Summe aller Tätigkeiten mit Genehmigung und

Anzeige bezieht (§ 80 Abs. 1 Nr. 1 StrlSchG). Der Dosisgrenzwert

für die Augenlinse wurde für beruflich strahlenexponierte

Personen von 150 mSv/a auf 20 mSv/a (§ 78

Abs. 12 Nr. 1 StrlSchG) deutlich abgesenkt. Für Einzelpersonen

der Bevölkerung wurde der Wert auf 15 mSv/a

(§ 80 Abs. 2 Nr. 1 StrlSchG) festgelegt. Im Übrigen bleiben

die bisherigen Dosisgrenzwerte bestehen.

Erfreulicherweise hat die in der RL 2013/59 angelegte

behördliche Vorabkontrolle nicht, wie noch in Bezug auf

die ersten Entwurfsfassungen der Richtlinie befürchtet,

dazu geführt, dass die bewährte deutsche Anzeigen- und

Genehmigungsstruktur durch die Richtlinie verkompliziert

worden ist. Die Umsetzung gab jedoch Gelegenheit,

die Anforderungen an die Genehmigungen für Betrieb und

Umgang teilweise zu vereinheitlichen (s. § 12 StrlSchG).

Die neue(n) Strahlenschutzverordnung(en)

Da das StrlSchG sich in vielen Bereichen auf eher strukturelle

Regelungen beschränkt, bedarf es zur Vollzugstauglichkeit

eines konkretisierenden untergesetz lichen Regelwerks.

Die Frist zur Umsetzung der RL 2013/59 lief bereits

am 6.02.2018 ab. Daher arbeitet das BMU derzeit mit

Hochdruck an Strahlenschutzregelungen auf Verordnungsebene.

Wie zu hören ist, soll es eine Artikelverordnung

werden, deren Herzstück die eigentliche neue Strahlenschutzverordnung

(StrlSchV neu) sein wird. Daneben soll

es eine Verordnung im Bereich der Notfallschutzmaßnahmen,

eine Verordnung zur Entsorgung radioaktiver

Abfälle sowie eine Verordnung zum Schutz vor schäd lichen

Wirkungen nichtionisierender Strahlung bei der Anwendung

am Menschen geben. Im Übrigen wird es in einer

Reihe von Gesetzen Anpassungen an das neue Strahlenschutzrecht

geben müssen, nicht zuletzt im Atomgesetz.

Freigabe: Was bleibt – was ändert sich?

Zwar ist noch kein Referentenentwurf zur neuen Strahlenschutzverordnung

bekannt. Aber über ein Thema wird

bereits seit mehreren Jahren diskutiert, das sowohl von

juristischem wie gleichermaßen von praktischem Interesse

ist, nämlich die Regelung zur Freigabe (§ 68 StrlSchG), die

zukünftig eine Änderung erfahren soll. Viel mehr als die

Ermächtigungsgrundlage für eine entsprechende Verordnung

enthält § 68 StrlSchG zwar nicht. Jedoch findet sich

in § 68 Abs. 1 S. 2 StrlSchG ein Hinweis auf eine zukünftige

Änderung bei der rechtlichen Behandlung der Freigabe.

In § 68 Abs. 1 S.2 StrlSchG heißt es: „In der Rechtsverordnung

können auch das Verfahren und die Mitteilungspflichten

für alle Fälle geregelt werden, in denen

die Voraussetzungen für die Freigabe nicht mehr

bestehen“.

Bisher können gemäß § 29 StrlSchV (im naturwissenschaftlichen

Sinne) radioaktive Stoffe bei entsprechender

Unterschreitung von näher festgelegten Grenzen als im

Rechtssinne nicht-radioaktive Stoffe verwendet werden,

wenn die Behörde die Freigabe für diese Stoffe erteilt und

die Übereinstimmung mit den im Freigabebescheid

festgelegten Anforderungen festgestellt hat. Die mit der

Novellierung der Strahlenschutzverordnung 2001

eingeführte Freigabe ist definiert als „Verwaltungsakt, der

die Entlassung radioaktiver Stoffe sowie beweglicher

Gegenstände, von Gebäuden, Bodenflächen ..., die

aktiviert oder mit radioaktiven Stoffen kontaminiert sind

und ..., aus dem Regelungsbereich

a) des Atomgesetzes und

b) darauf beruhender Rechtsverordnungen sowie

verwaltungsbehördlicher Entscheidungen

zur Verwendung, Verwertung, Beseitigung, Innehabung

oder zu deren Weitergabe an Dritte als nicht radioaktive

Stoffe bewirkt“.

An der Definition der Freigabe soll sich auch in Zukunft

nichts ändern. Wenn also die Behörde die Freigabe

(schriftlich) erteilt und die Übereinstimmung mit den

vorgeschriebenen Anforderungen festgestellt hat (keine

Überschreitung der effektiven Dosis im Bereich der

„Bagatell grenze“ von 10 Mikrosievert/a für Einzelpersonen

der Bevölkerung, § 29 Abs. 2 S.1 StrlSchV), ist

der betreffende Stoff aus dem Regelungsbereich des Atomgesetzes

und den einschlägigen Verordnungen entlassen

und unterfällt dem konventionellen Abfallrecht (Kreislaufwirtschaftsgesetz).

Dem Freigabeverfahren liegt wie

bisher das De-minimis-Konzept zugrunde: „de minimis

non curat lex“. Wie bisher auch sind im Gegensatz zur

uneingeschränkten Freigabe bei der zweckgerichteten

Freigabe Einschränkungen hinsichtlich der Verwertung

oder Verwendung zu beachten. Eine Möglichkeit der

zweckgerichteten Freigabe ist die Deponierung, die vor

dem Hintergrund des zunehmenden Abrisses von Kernkraftwerken

als Entsorgungsweg z.B. für freigegebene

Baustoffe an Bedeutung gewinnt. Beispielsweise könnten

also freigegeben Baustoffe aus dem Abriss von Kernkraftwerken

in Deponien für konventionellen Anfall eingebracht

werden.

Deponierung: Ein zunehmend problematischer

Entsorgungsweg

Hiergegen wenden sich jedoch immer häufiger am Standort

der jeweiligen Deponien ansässige Bürger und Bürgerinitiativen,

die Bedenken gegen die gefahrlose Einlagerung

freigegebener (geringfügig radioaktiver) Stoffe geltend

machen, obwohl die Genehmigungs behörden, wie Brigitte

Röller (Staatsministerium für Umwelt und Landwirtschaft

Sachsen) anlässlich der 15. Tagung der Deutschen Landesgruppe

der International Nuclear Law Association anschaulich

darstellte, große Anstrengungen unternehmen,

um über die Gering fügigkeit der Radioaktivität in freigegebenen

Stoffen zu informieren, und zusätzliche Kontrollmessungen

der angelieferten freigegebenen Stoffe in

Anwesenheit der Öffentlichkeit durchführen sowie Vergleichsmessungen

mit von Bürgern mitgebrachten Produkten

(z.B. Garten erde) anbieten. Da sich der Widerstand

gegen die Deponierung freigegebener Stoffe bereits seit

einigen Jahren formiert hat und Betreiber privatrechtlich

organisierter Deponien begannen, ihr Interesse an der

Annahme freigegebener Abfälle zu verlieren, wurde 2015

in Baden­ Württemberg federführend durch den Landkreistag

und den Städtetag unter Mitwirkung des Umweltministeriums

eine Handlungsanleitung entwickelt. Diese

Handlungs anleitung sieht zwecks größtmöglicher Transparenz

und Vertrauensbildung in der Öffentlichkeit über

das übliche Freigabeverfahren hinaus Maßnahmen vor, die

die Strahlenexposition noch weiter reduzieren sollen (z.B.

Schutz vor Staubentwicklung bei der Anlieferung,

Konzentration der Anlieferung von Abfalltransporten, statt

Stichprobenmessungen hundertprozentige Nach messung

durch den von der Behörde zugezogenen Sachverständigen,

außerdem Möglichkeit von Stichprobenmessung

durch den vom Deponiebetreiber zugezogenen

Sachverständigen.)

SPOTLIGHT ON NUCLEAR LAW 297

Spotlight on Nuclear Law

The New Radiation Protection Law and the Approval: May Makes Everything New? ı Ulrike Feldmann


atw Vol. 63 (2018) | Issue 5 ı May

SPOTLIGHT ON NUCLEAR LAW 298

Die Erfahrungen beispielsweise aus Sachsen zeigen

jedoch, dass die vielfältigen vertrauensbildenden Maßnahmen

und das Bemühen um Transparenz ins Leere

gehen, da sich die ortsansässigen Bürger nach wie vor gegen

die Deponierung freigegebener Abfälle wenden und

überwiegend an Informationen und Kontrollmessungen

nicht interessiert sind. In Sachsen lehnen soweit ersichtlich

daher von ehemals 5 Deponien inzwischen min destens

3 die Einlagerung freigegebener Abfälle ab.

Freigabe mit Nebenbestimmungen:

Eine Lösung des Problems?

Aufgrund der steigenden Schwierigkeiten bei der

Deponierung wurde im Rahmen der Umsetzung der RL

2013/59 im Länderausschuss für Atomkernenergie eine

Bund-Länder-Arbeitsgruppe „Freigabe“ eingesetzt, um die

„Rücknahmemöglichkeit“ einer erteilten Freigabe zu prüfen.

Im Ergebnis wird vorgeschlagen, die Ausge staltung des

Freigabeverfahrens dergestalt zu ändern, dass über die

Möglichkeiten des § 17 Abs. 1 S. 2-4 AtG hinaus die Freigabe

mit einer Bedingung, einem Vorbehalt des Widerrufs oder

einem Vorbehalt der nachträglichen Aufnahme, Änderung

oder Ergänzung einer Auflage erlassen werden kann.

§ 36 Abs. 1 Verwaltungsverfahrensgesetz (VwVfG)

erlaubt grundsätzlich derartige Nebenbestimmungen

zum Verwaltungsakt, wenn sie durch Rechtsvorschrift

zugelassen sind oder wenn sie sicherstellen sollen, dass die

gesetzlichen Voraussetzungen des Verwaltungsakts erfüllt

werden. Nach § 36 Abs. 2 VwVfG muss die Verbindung

eines Verwaltungsakts mit einer Befristung, einer

Bedingung, einem Vorbehalt des Widerrufs, einer Auflage

oder einem Vorbehalt der nachträglichen Aufnahme,

Änderung oder Ergänzung einer Auflage nach pflichtgemäßem

Ermessen erfolgen. Nach § 36 Abs. 3 VwVfG

darf eine Nebenbestimmung dem Zweck des Verwaltungsaktes

nicht zuwiderlaufen.

Bei der hier in Rede stehenden Deponierung freigemessener

Abfälle soll mit Hilfe von Nebenbestimmungen

– zu denken wird hier in erster Linie an eine (auflösende)

Bedingung oder einen Widerrufsvorbehalt (Unterfall der

auflösenden Bedingung) sein – sichergestellt werden,

dass, wenn die Deponierung nicht erfolgreich war, die

freigemessenen Abfälle wieder dem Atom- und Strahlenschutzrecht

unterfallen.

Diese neue Ausgestaltung der Freigabe erscheint jedoch

nicht frei von Bedenken.

Zwar sollen Nebenbestimmungen in der neuen

Strahlen schutzverordnung zugelassen sein. Insoweit

wären sie nach § 36 Abs. 1 VwVfG erlaubt. Es fragt sich

aber, ob eine auflösende Bedingung oder ein Widerrufsvorbehalt

dem Zweck des Verwaltungsaktes nicht

zuwiderlaufen. (§ 36 Abs. 3 VwVfG).

Zweck der Freigabe (siehe Definition der Freigabe in § 3

Abs. 2 Nr.15 StrlSchV) ist es, die Entlassung radioaktiver

Stoffe sowie beweglicher Gegenstände, von Gebäuden,

Bodenflächen ... aus dem Regelungsbereich des Atomgesetzes

und darauf beruhender Rechtsverordnungen

sowie verwaltungsbehördlicher Entscheidungen zur

Verwendung, Verwertung, Beseitigung, Innehabung oder

zu deren Weitergabe an Dritte als nicht-radioaktive Stoffe

zu bewirken. An der Definition der Freigabe in der

geltenden StrlSchV soll sich auch zukünftig nichts Wesentliches

ändern. Bei den zur Deponierung freigegebenen

Abfällen darf man annehmen, dass in der Regel die

Freimessung ordnungsgemäß erfolgt ist, die freige gebenen

Stoffe also die Dosis im Bereich von 10 Mikrosievert

nicht überschreiten. Die rechtlichen Voraussetzungen des

Inhalts des Verwaltungsaktes “Freigabe“ sind damit im

Regelfall erfüllt. Eine Bedingung dergestalt, dass die

Freigabe nur wirksam wird, wenn der Einbau in der

Deponie erfolgreich war, oder ein Widerrufsvorbehalt

dergestalt, dass bei nicht erfolgreicher Deponierung die

Freigabe widerrufen werden kann (z.B. weil wegen einer

Demonstration vor den Toren der Deponie die Abfälle

nicht auf das Deponiegelände verbracht werden können),

widerspricht dem Zweck der Freigabe, die Entlassung der

freigemessenen Stoffe aus dem Regime des Atom- und

Strahlenschutzrechts zu bewirken, und führt dazu, diese

Stoffe in diesem Regime zu belassen, obwohl sie die

Voraussetzungen der Freigabe erfüllen. Die Pflicht der

Behörde bzw. des Staates ist es sicherzustellen, dass

ordnungsgemäß freigegebene und im Rechtssinne

nicht-radioaktive Stoffe auch nach dem Kreislauf wirtschafts

recht in eine geeignete Deponie verbracht werden

können. Eine Neben bestimmung hier einzuführen, um

die im Rechtssinne nicht­ radioaktiven Stoffe dauerhaft zu

radioaktiven Stoffen zu machen, ist eine juristische

Krücke, die verdeckt, dass der Rechtsstaat sich offenbar

nicht in der Lage sieht, das von ihm gesetzte Recht (Recht

wie auch die Pflicht der Abfallablieferer, bestimmte freigemessene

Abfälle an eine Deponie abzuliefern) auch

durchzusetzen. Im Übrigen muss der Ablieferer auf den

der Behörde festgestellten Status seiner freigemessenen

Abfälle als im Rechtssinne nicht-radioaktive Stoffe

vertrauen dürfen.

Auch unter eher praktischen Gesichtspunkten bestehen

Bedenken gegen den Erlass von Nebenbestimmungen.

Wenn bei gescheiterter Deponierung die im Rechtssinne

nicht-radioaktiven Stoffe dauerhaft zu im Rechtssinne

radioaktiven Stoffen werden, wird dies zum einen in

der Öffentlichkeit nicht als eine vertrauensbildende

Maßnahme verstanden werden, sondern ohnehin besorgte

Bürger in ihrer Besorgnis bestätigen. An Deponiestandorten

dürfte dieses negative Signal zukünftig

als „Einladung“ verstanden werden, die Deponierung

scheitern zu lassen.

Zum anderen wird mit einer Bedingung oder einem

Widerrufsvorbehalt das Problem der Zwischen- und Endlagerung

der freigemessenen Abfälle nicht gelöst, sondern

nur auf die Ebene der im Rechtssinne radioaktiven Stoffe

verschoben (wohin diese Abfälle eigentlich gar nicht

gehören), wodurch – man darf annehmen – zulasten

der Abfall ablieferer die Menge der zwischen- bzw. endzulagernden

radioaktiven Abfälle erhöht wird.

Fazit

Das neue Strahlenschutzrecht wird weiterhin von den

Grundsätzen des Strahlenschutzes – Rechtfertigung,

Dosis begrenzung und Optimierung – geprägt sein und

auch im Übrigen viel Bewährtes aus dem bisherigen Recht

übernehmen. Zahlreiche Änderungen und Ergänzungen

sind gleichwohl im Strahlenschutzgesetz enthalten. Mit

den meisten wird man sich vermutlich anfreunden können.

Welche Neuerungen die künftige Strahlenschutzverordnung

neben der verwaltungsverfahrensrechtlichen

Änderung bei der Freigabe im Einzelnen bereithält, wird

sich vielleicht im Mai zeigen, sollte dann die Verbändeanhörung

eingeleitet werden.

Author

Ulrike Feldmann

Berlin, Germany

Spotlight on Nuclear Law

The New Radiation Protection Law and the Approval: May Makes Everything New? ı Ulrike Feldmann


atw Vol. 63 (2018) | Issue 5 ı May

Continuous Process of Safety Enhancement

in Operation of Czech VVER Units

J. Duspiva, E. Hofmann, J. Holy, P. Kral and M. Patrik

A continuous process of a safety enhancement of VVER units in the Czech Republic is briefly described

including a presentation of important milestones and examples of particular safety measures already implemented.

A special attention is given to the evaluation and implementation of safety measures following stress tests recommendations

and R&D activities supporting this process. As examples an implementation of the „design extension condition

without core melt“ concept and various activities related to severe accident mitigation strategies are presented in the

more detailed way.

1 Introduction

A safety enhancement process is a

continuous effort which has been

taking place in both VVER sites at the

Czech Republic since their commissioning.

Within this systematic and

focused process, checked regularly by

periodic safety reviews, international

missions and independent inspections,

a level of safety and safety culture has

been significantly increased within

the last decades. Despite the high

level of safety reached by the preventive

means mainly at both VVER sites

in the context of the Fukushima accident,

a new period of enhancement

process has been initiated following a

stress tests exercise. Use of advanced

deterministic analytical approaches

and implementation of new preventive

safety measures has then taken

place together with a special attention

to the mitigative part of potential accidents

and relevant strategies and

measures.

The important attributes for new

potential safety measures have been

diversity and use of alternate means.

As examples of such approach could

be mentioned the diverse power

sources, the diverse RCS/SFP/CNTM

(reactor cooling system/spent fuel

pool/containment) make up, the

alternate secondary heat sink, the

alternate key parameters monitoring,

or the new communication means,

etc.

Within the systematic approach to

enhance safety a PSA has been recognized

as a very useful tool too. The

PSA has been mainly used for identification

of weak points in design

and operation and for an evaluation

and prioritization of potential safety

measures. On the other hand also a

use of PSA monitoring system helps to

manage an operational risk level

particularly during outage periods.

The other PSA applications, which

took place during the last decade, can

be represented by various case studies

devoted to modifications of plant

tech- specs, modification of test and

maintenance intervals, event analysis

etc. Besides that, the methods of risk

analysis are individually and indirectly

applied for development of procedures

related to plant safety (symptom

based procedures for emergency

scenarios, procedures for operation

in plant abnormal status, severe

accident management guidelines,

plant Tech-Specs), for the support of

plant crew training, including training

of operators at full scope simulator,

for improvement of MMI in the main

control room as well as in local control

rooms, and in other areas of the

direct manipulation with the plant

equipment, including planning and

timing of actions carried out under

non-standard conditions.

As an example of using the PSA

for identification and evaluation of

safety measures, additional cooling

towers at the Dukovany site could be

mentioned. The ultimate heat sink

problem as a consequence of some

external events of high intensity was

identified in 2008, long time before

Fukushima event, where the extreme

wind dominated the external hazards

risk. In the first version of external

events PSA, very high risk contribution

of this external hazard was

related to unsatisfactory resistance

of “big” plant passive cooling towers

(the active cooling towers – so called

fume cooling towers – had been made

the elements of original plant design,

but were not built during plant construction).

The important general

conclusions from the PSA (missing

ultimate heat sink for external winds

with return time period of 10,000

years) were doubted in the discussions

before the Fukushima event, but

a significant change in the opinion

after the Fukushima event led to an

installation of active cooling towers

with significant positive impact on the

external events’ plant risk. A full effect

of this modification was reached as

soon as the modification in the design

was supported by changes in procedures.

In the following chapters more

attention is paid to advanced deterministic

analytical approaches and an

implementation of severe accident

management strategies.

2 Implementation of

design extension condition

without core melt

(DEC-A)

The work on DEC-A (previously

BDBA) safety analyses for Czech NPPs

was initiated in 2009 as a consequence

of the Periodical Safety Review (PSR)

after 20 year of the operation of the

Dukovany NPP. This effort has been

influenced also by initiatives and

suggestions from European Utility

Requirements (EUR), WENRA safety

reference levels and by the IAEA

introduction of the DEC term and

concept into the safety standards

series.

2.1 Methodology basis for

DEC-A assessment in the

Czech Republic

Aside from the IAEA recommendations

and the Czech Atomic Law,

the following regulations, directives

and reports constitute the legislative

and methodological basis for deterministic

analyses of DEC-A (BDBA) in

the Czech Republic:

• SUJB regulation 195/1999, Requirements

on Nuclear Facilities

for Ensuring of Nuclear Safety,

Radiation Protection and Emergency

Preparedness, 1995.

• SUJB directive BN-JB-1.6, Probabilistic

Assessment of Safety, 2010

(currently revised due to new

Atomic Law).

• SUJB directive BN-JB-1.7, Selection

and Assessment of Design and

Beyond Design Events and Risks

for Nuclear Power Plants, 2010

299

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atw Vol. 63 (2018) | Issue 5 ı May

OPERATION AND NEW BUILD 300

(currently revised due to new

Atomic Law).

• UJV, Proposal of Methodological

Procedure for Performing of Safety

Analysis of Beyond Design Basis

Accident, UJV Rez, 2010.

Analyses of DEC-A scenarios use the

best estimate computer codes with

combination of realistic initial and

conservative (or realistic) boundary

conditions. The robust design of VVER

reactors and their safety features

enable to fulfil DBA acceptance

criteria in most DEC-A cases including

radiological consequences. For the

most severe conditions comprising

multiple failures of safety systems or

safety groups providing protection in

the level 3a of Defense in Depth (like

SBO), the new measures imple mented

after post-Fukushima Stress tests in

the level 3b of the DiD provide an

additional robust protection against

the evolution of these scenarios into

the DEC-B category (severe accident).

The acceptance criteria applied to

DEC-A analyses are identical to those

applied to DBA analysis with exception

of criterion on primary and

secondary pressure and radiological

consequences.

The computer code used for NPP

safety analyses in the Czech Republic

must be approved by the regulatory

body according to the SUJB directive

VDS-030.

2.2 Selection of DEC-A events

to be analysed and

documented in SAR

The basic set of DEC-A (BDBA) events

to be analyzed is specified in

BN-JB-1.7 0. Supplemental events

and scenarios could be specified by

PSA outcomes and engineering

judgement.

It is important to mention that in

analyses of DEC (which are often

complex sequences or combinations

of events and failures) it is logical to

transfer from “frequency of initial

events” to “frequency of occurrence of

scenarios”.

The SUJB directive BN-JB-1.7 0

requires besides the standard set of

ATWS analyses, the following DEC-A

(BDBA) events to be analyzed:

• Total long-term loss of inner and

outer AC power sources;

• Total long-term loss of feed water

(„feed-and-bleed„ procedure);

• LOCA combined with the loss of

ECCS;

• Uncontrolled reactor level drop or

loss of circulation in regime with

open reactor or during refueling;

• Total loss of the component cooling

water system;

• Loss of residual heat removal

system;

• Loss of cooling of spent fuel pool;

• Loss of ultimate heat sink (from

secondary circuit);

• Uncontrolled boron dilution;

• Multiple steam generator tube

rupture;

• Steam generator tube ruptures

induced by main steam line break

(MSLB);

• Loss of required safety systems in

the long term after a design basis

accident.

The whole set of prescribed DEC-A

analyses was already performed both

for Dukovany NPP (VVER-440) and

for Temelín NPP (VVER-1000).

Analyses of DEC-A events for the

Czech NPP’s have been performed

with the RELAP5 computer code. It is

worth noting that the RELAP5 has

been in the UJV Rez validated against

experimental data from more than 20

tests carried out at various integral

test facilities (ITF) and that approximately

half of these tests were modelling

events of the DEC-A type.

2.3 Example of DEC-A analysis:

SBLOCA in VVER-1000 with

failure of ECCS and operator

start of HPSI at 30 min

The analysis of a small break loss

of coolant accident (SBLOCA) with

the break D50 mm in the cold leg

and with a failure of the start of

emergency core cooling systems

(ECCS) and operator manual start of

high pressure safety injection (HPSI)

at 30 min was performed for the

| | Fig. 1.

Nodalization scheme of VVER-1000 for RELAP5 (only primary circuit and 1 of 4 modeled loops depicted).

Operation and New Build

Continuous Process of Safety Enhancement in Operation of Czech VVER Units ı J. Duspiva, E. Hofmann, J. Holy, P. Kral and M. Patrik


atw Vol. 63 (2018) | Issue 5 ı May

VVER-1000. The nodalization scheme

of the VVER­ 1000 unit for the RELAP5

used and graphical courses of main

parameters are shown in Figure 1 and

Figure 2.

Loss of primary coolant through

the break D50mm in cold leg and

without an automatic actuation of

ECCS leads to a depletion of a primary

inventory and if not mitigated by

operator, the core uncovery and

overheating start. However with

respect to high “water volume to

power ratio” in the VVER-1000, there

is sufficient time for the operator

intervention. In the presented case,

the operator starts one HPSI pump at

30 min and soon after it the core is

quenched and its cooling restored and

stabilized.

2.4 Introduction of DEC-A

analyses into Safety

Analysis Report of

Dukovany and Temelín

NPP

The whole set of prescribed DEC-A

analyses was already performed both

for Dukovany NPP (VVER-440) and for

Temelín NPP (VVER-1000). It means

that from 15 to 20 DEC-A analyses for

each plant was elaborated.

As for the documentation of DEC-A

analyses in Safety Analysis Report, the

temporary solution was creation of a

new subchapter 15.9.1 which contains

basic results of all DEC-A (BDBA)

analyses required by BN-JB-1.7.

Beside that the ATWS analyses are

documented in subchapter 15.8 of the

SAR as usually.

The final foreseen solution is the

introduction of the new SAR chapter

19, that would contain both DEC-A

(BDBA without core melt) and DEC-B

(severe accident) analyses presented

in systematic and integrated way.

Then the Chapter 15 will be again

intended for analyses of events up to

DBA only.

3 Implementation of

severe accident

strategies and particular

measures

The solution of severe accident topics

for both Czech NPPs was initiated

at the end of 1980ies of last century

as the response to the Chernobyl

accident. The activities were supported

by IAEA via. regional projects

at the beginning and after the political

changes in the Czechoslovakia, the

main source of experience was shifted

to the CSARP program. All those

activities were performed by the

group on severe accident in the UJV

| | Fig. 2.

Reactor core temperatures (SBLOCA D50mm in VVER-1000 with failure of ECCS).

Rez. Later the group started to be

partner in international activities and

recently it plays an important role in

the safety enhancement of Czech

NPPs in the relation to the severe

accident management. Generally the

group is oriented on the analytical

activities, the experimental ones were

related only to the small scale test on

the corium properties using the cold

crucible facility. Recently the new

program on the reactor pressure coolability

during the in-vessel retention

strategy is under preparation and the

large scale facility under construction.

Historically the activities were at

the beginning focused on the plant

vulnerability studies, an identification

of typical timing of a severe accident

progression and the first basic source

term estimations. Later the activities

were extended to the identification of

potential plant modifications related

to mitigation of important severe

accident phenomena like hydrogen

issue and so on. The recent key activity

for the Czech NPPs is related to a

solution of the corium localization for

the Temelín NPP (VVER-1000/320),

but also other activities are on-going.

Any analytical activities in area of

severe accident must be done on

appropriate qualitative level. The

systematic approach in the UJV

consists of nine pre-conditions which

must be continuously filled up to be

granted that analytical results are

credible. The list of preconditions is

following

• Knowledge of SA phenomenology

• Knowledge of analyzed facility

(power plant unit or experimental

facility in a case of validation

analysis)

• Validated and state of the art

analytical code

• Knowledge of analytical tool by

analyst

• Close collaboration between code

users and developers

• Exchange of user experience,

recommendations and best

practices

• Code qualification for specific

design feature (or for specific

reactor design like in the

VVER-440/213 case)

• Plant specific input development

and testing

• Results evaluation and interpretation

taking into account the

objective and assumptions of the

analysis as well as the code limitations

– code must not be used as

black box

Extensive validation activities are very

important for the filling up of some of

above mentioned preconditions and

the UJV is very active. As mentioned

above the group is analytical and

the access to the experimental results

is possible only via. international

cooperation. The UJV participated in

several international projects during

last more than 15 years like Phebus FP,

projects of the 5th EC FWP (ARVI,

ICHEMM, LPP, OPTSAM), later in

SARNET and SARNET2, but also

within the OECD/NEA projects like

RASPLAV, MCCI, THAI, SFP, STEM

usually including their follow up.

Another important activity is a

participation in the international

benchmarks or exercises like the

OECD/NEA ISPs (ISP-45 on

Quench-06 test, ISP-46 on Phebus

FPT-1, ISP-41 on RTF, ISP-44 on

KAEVER or ISP-49 on OECD THAI HD

tests), SARNTET benchmarks on

Quench-11 test or Phebus FPT3 tests

or benchmarks within Analytical

Working Group of OECD THAI project

OPERATION AND NEW BUILD 301

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Continuous Process of Safety Enhancement in Operation of Czech VVER Units ı J. Duspiva, E. Hofmann, J. Holy, P. Kral and M. Patrik


atw Vol. 63 (2018) | Issue 5 ı May

OPERATION AND NEW BUILD 302

and its both follow up phases (latest

on THAI TH-27 test and THAI HR-49).

The past activities in the severe

accident for both Czech NPPs are were

mostly focused on the enhancement

of the safety using only existing

systems on site, but out of their operational

conditions. The Fukushima Daiichi

accident significantly changed the

approach to the solution of severe

accidents at the Czech NPPs and the

implementation of new, and only for

severe accident conditions dedicated,

systems was fully accepted as the

necessary step forward in the safety

enhancement. The proposals were not

started from the scratch, but both NPP

were evaluated within the Stress

Tests, which reports were issued to

the Czech Republic State Office for

Nuclear Safety (SONS), and based on

these reports the national Stress Test

report [A] was prepared and evaluated

within the evaluation process

under the ENSREG leadership resulting

in the forming of the National

Action Plan [B]. Concerning the

severe accident issues the eight main

areas were defined (as a selection of

those most important ones from much

higher number of issues)

• Increase of the capacity of the system

for liquidation of emergency

hydrogen (action 46 – Dukovany

NPP, action 47 – Temelín NPP) –

this title is undertaken from the

official National Action Plan and it

means increase of recombination

power of emergency hydrogen as

the term “liquidation” is unusual in

this meaning

• Cooling of the melt from the outside

of RPV (action 48 – Dukovany

NPP)

• Recriticality (action 61 – both

NPPs)

• Control room habitability (actions

58, 31, and 51 – both NPPs)

• The means for maintaining containment

integrity due to overpressure

(actions 46 – 50 – both

NPs)

• Corium in/ex vessel cooling

( actions 48, 49, 50 – Temelín NPP)

• Extension of SAMGs for shutdown/

severe accident in SFP ( action 56 –

for both NPPs)

• System setup of training (drills),

exercises and training for severe

accident management according to

SAMG, including possible solution

of multi-unit severe accident

The solutions of some topics listed

above are independently described

in following subchapters with the

pointing out the contribution of the

UJV to their solution.

3.1 Increase of capacity of

system for emergency

hydrogen removal

The UJV performed a full analytical

support for the increased design of

the hydrogen removal system. The

analytical program consisted from

several steps. The first step contained

the integral analysis of the severe

accident progression with the aim to

define mass and energy source into

the containment for selected severe

accident scenarios. The scenario

selection was based on several conditions

like – a location of the hydrogen

source in the containment, its

potential intensity, an operation of

the containment spray system, or conditions

of the molten corium concrete

interaction. Generally three initiating

events were selected and overall six

integral analyses performed with the

MELCOR 1.8.6 code. The sources to

the containment were later used in

the stand alone analyses of the containment

response using the very

detailed model of the containment

again for the MELCOR 1.8.6 code.

Two kinds of analyses were performed

– first the hydrogen risk evaluation

based on the Sigma and Lambda

criteria, which were used for the first

proposal of PAR design, second the

optimization analyses with the aim to

fullfil predefined succes criteria – an

elimination of Lambda criterion in all

parts of containment, an elimination

of Sigma in practically all parts of

containment (small individual spaces

allowed for temporary occurrence),

global and local concentration limits

after recalculation of hydrogen concentration

in dry air. The design of

the PARs of the NIS vendor was

succesfully implemented at both units

of the Temelín NPP with finalization

and starting of its full operation after

outages in 2015.

3.2 Recriticality of degraded

core due to reflooding

with demi water

The UJV is recently performing

analytical investigation of this topic

independently for each of Czech

NPPs, because of their principal

design differences. The methodological

approach is identical and consists

of the integral analyses of the severe

accident progression using the

MELCOR 2.2 code with an externaly

defined boric acid to analyze the

development of boric acid concentration.

In parallel the most important

configurations of the degraded core

are defined to be analysed using the

MCNP code to identify the minimum

concentration of boric acid leading to

the recriticality. The evolution of boric

acid concentrations vs. the minimum

value will allow to define conditions

for the applicability or the restriction

of application of demi water for

injecting into the degraded core under

various stages of severe accident

course.

3.3 Control room habitability

This study was performed again

independently for each of Czech NPPs

and the UJV performed these analyses.

The methodology consisted of two

main analytical steps – the first one

covered the integral analyses with the

MELCOR 1.8.6 code for the identification

of fission product distribution

and releases via different leak paths.

The second step included the analysis

of dose rates in the control room based

on the precalculated distribution

of fission products and shielding of

control room walls or window (if

presents).

3.4 Corium localization

for Temelín NPP

The issue of the corium localization

is very complex and determines

the solution of others activities, like

a solution of long-term containment

pressure control. Before the

Fukushima Dai-ichi event the analytical

activities were focused only on

the ex-vessel corium cooling (ExVC)

strategy, because of a restriction on an

implementation of any new equipment

for severe accidents. The request

from the NAcP opened a way for

the solution of the in-vessel corium

retention (IVR) strategy as an alternative

one, which applicability has to

be evaluated. The utility opened at the

beginning two preparatory projects,

which enabled to prepare the fisrt

analytical models for the corium

behaviour in the lower head and cooling

conditions of rector pressure

vessel from outside. As the outcomes

from the analytical work identified

some potentials, the complex project

on the IVR solution was initiated in

2015 with expected duration up to

5 years. The project is not focused only

on analytical work, but also on the

experimental confirmation of the

VVER-1000 RPV coolability under the

IVR strategy.

The project is focused on three

types of activities – analytical investigation

with proposing of additional

systems for strategy solutions, designing

of new systems required for the

implementation of strategies, and the

experimental program for the RPV

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coolability. As the activities have to

produce a reasonable contribution to

the safety enhancement, the following

conditions were defined

• Effectiveness, it means a quantified

benefit to safety (prevention of

early FP releases and minimize FP

releases) plus confirmed physical

fruitfulness with sufficient margin

• Reasonable technical feasibility

• No negative impact to reactor operation

(including all procedures/

activities during outages)

• Simplicity – applicability under

SA condition (limited personal

capacity, limited accessibility …)

• At least partial independence of

functionality assurance in comparison

with existing emergency

systems

• Consistency of approaches with

other utilities operating VVER-1000

(or reactors of similar power) and

VVER-1000 designers

Taking into account above defined

conditions the activities were defined

for following six main topics which

have very well defined structure and

relations each to other

• Primary circuit depressurization

under SA conditions

• Corium cooling with water injection

into RPV (new acronym IVR-IN

is used)

• In-vessel retention with external

reactor vessel cooling strategy

(originally used IVR, but newly

IVR-EX acronym is used)

• Ex-vessel corium cooling strategy

(ExVC)

• Containment response to severe

accident and long term issues

• Severe accident initiated in the

spent fuel pool

Short description of the effort done up

to know is included in following

sub-chapters excluding the last topic,

of which activities are not yet initiated,

but some analyses were already

performed in past.

3.4.1 Primary circuit depressurization

under SA conditions

The activities related to this topic were

mostly focused on the conditions for

the IVR-EX strategy as the primary

pressure is strogly determining

success of this strategy. The analytical

investigation using the MELCOR 1.8.6

code was performed to identify if the

depressurization initiated after the

entry to SAMGs is sufficiently fast to

meet predefined criterion – to reach

the pressure reduction below 2 MPa-g

till the time of starting relocation of

corium into the lower plenum of RPV.

The analytical studies, for variant

opening of one or two safety valves or

additionally the system for a removal

of gases from the pressurizer, confirmed

that the safety valves if can be

kept open are sufficient, but the contribution

of the gas removal system is

negligible. The important condition is

that the valves must be kept open, the

modification of a control of safety

valves is in the final phase of its

preparation with an expectation to

initiate an appropriate technical

negotiation with the SONS.

3.4.2 Corium cooling with

injection into RPV

(IVR-IN strategy)

The extensive analytical investigation

was performed with the MELCOR

1.8.6 code to analyze possibility of a

termination of the core degradation

due to water injection into the RPV.

The analyses varied with the assumed

mass rate of the water injected (to

alternate among the repaired LPI and

HPI, and the new alternative system

called TB50, which has the lowest

mass rate of water injection and also

relatively lower maximum pressure

head). The second variations were in

timing of the water injection activation,

it means different configurations

of core degradation at the time of

water injection initiation. The last

variation was in the initiating event as

the most of analyses were performed

for the large break LOCA initiated

envent (plus loss of all active systems)

and as an alternative the scenario

initiated with postulated SBO was

chosen with the early depressurization

after the entry to SAMGs and also

loss of all active systems.

The conclusion from the analytical

work is that the results are of course

uncertain, but they show that the

initiation of water supply before starting

of the corium relocation into the

lower plenum should prevent the loss

of RPV integrity. More detailed, the

analysis predicted un-failed RPV wall

if the lower head is not dried out, but

such answer seems to be too optimistic

due to some modeling assumptions in

the MELCOR code which favoured

cooldown with even very small

remaining mass of water in the lower

head.

3.4.3 Strategy IVR-EX

The greatest effort was done for this

topic during last two years and the

activity will continue with the experimental

program on the RPV coolability.

The first issue related to the IVR-EX

strategy is the flooding of reactor

cavity and long term water supply

to the cavity. The design of the

VVER-1000/320 containment is absolutely

un-favorable for this IVR-Ex

strategy, because the recirculation

sump is located one floor below the

cavity and the water from the containment

is drained to this sump. So it

is absolutely impossible to close the

water circulation inside of the containment,

like it is in case of the

AP1000 or VVER-440/213 reactors.

The water for flooding of the cavity

must be injected from sources outside

of the containment. The feasibility

study was performed to investigate all

necessary modifications to be done

for a possibility to cool the RPV

from outside. The preliminary project

solution was prepared for new systems

on fast cavity flooding (based on

pressurized tanks located on roof of

the auxiliary building) with the

second part for feeding water from the

second pair of pressurized tanks. This

solution enabled to cover the first

6 hours by this passive system and

thus to have a way sufficient time for

an activation of the new active system

which is capable to feed water from

own tanks till 72 hours of the accident.

The injection of the water is

not the only condition, but the second

one is the protection of water releases

from the cavity. The injection of

water is assumed via. the venting

system channels and those must be

equipped with new valves to prevent

water overflow to the containment.

Generally it means to install 9 new

valves which must all of them be

surely closed before starting of injecting

water to the cavity. This is the

most critical part of the solution,

because of possible failure, which

would cause the loss of the IVR-EX

strategy. The second critical point is

the control of water level for the

second set of pressurized tanks. The

control is proposed to prevent loss of

this water due to overflow, but this

control can be tested and tuned up

only at the experimental facility, so its

credibility is not too high.

The second issue is the release of

steam produced during the cooling

of RPV by boiling. The design drawings

of the Temelín NPP showed that the gap

between the RPV itself and its

supporting ring should be sufficient

(including thermal expansion of RPV),

but the data from measurement after

the vessel installation were not available.

So at least part of this gap was

measured during the outage at the unit

2 in 2016 and it confirmed sufficient

gap. But the gaps among the boxes of

thermal and biological shielding seem

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OPERATION AND NEW BUILD 304

to be insufficient as their position

varies after each operation with those

boxes due to replacement of probes.

The conclusion is that in case of the

IVR-Ex strategy implementation, the

replacement of some boxes is needed

with modified ones, which will include

the burst membrane and sufficient

flow channel.

The analytical studies on RPV

coolability showed very low margin to

critical heat flux (CHF) if the cooling

is in a water pool. So the idea of

the intensification of heat removal

appeared and feasibility study on

the application of the deflector was

performed. The conclusion is that the

deflector is technically feasible, but

too many new risks are related to its

implementation that the deflector

cannot be acceptable as the measure

for a safety increase. Just a few

examples of those risks. The deflector

would need to be removed during

every outage so the process of its

disassembling and later re-assembling

using the manipulators would be very

risk for the hiting RPV. Also storing of

the deflector components, which are

irradiated, would be high risk for

additional dose rates to personal. So

the conclusion is that the deflector is

not reasonably applicable.

Verification of IVR-EX efficiency

for VVER-1000. This topic covers

practically all plant analyses for the

IVR-EX and also the analytical support

for the designing of the experimental

facility (called THS-15) on the RPV

coolability. This topic also covers the

designing of the facility itself. As

examples of the activities, it can be

mentioned – a study on the heat flux

distribution from corium to RPV wall

using the FLUENT code, RELAP

analyses of RPV cooling with various

parameters of deflector, pre-test

analyses of the THS-15 facility,

supporting scaling analyses, but also

a preparation of the methodology for

an evaluation of experimental results,

a preparation of the experimental

matrix and so on.

Although it is not part of the project,

it is logical to put into this subchapter

the information on the building of the

THS-15 facility. The facility purpose is

to perform tests on the cooling of RPV

of the VVER-1000 reactor under IVR-EX

conditions. The tests will be focused on

the verification of a coolability of heat

fluxes, an identification of CHF, and to

produce data for the code validation.

The experiments will be performed as

the part of the WP4 of the EC H2020

Project IVMR (grant agreement

number 662157). The recent situation

in the facility building is that the

production of the last big component

is finished and the facility assembling

is ongoing. The scheduled com missioning

of the facility has to be at the

end of November 2017. Then the experimental

program will be launched.

3.4.4 Strategy ExVC

The alternative strategy to the IVR-EX

is the ex-vessel corium cooling one.

The idea of the strategy is to spread

the corium into the cavity neighboring

room (named GA302) and cover the

corium with the water to cool it down.

Several analytical studies as well as

the experimental data from the OECD

MCCI and MCCI2 project showed, that

the MCCI with the siliceous concrete

is not possible to terminate. Based

on this observation the idea of the

refractory material, which would

survive at least several hours till

significant corium temperature reduction

appeared. The project also

defined some topics of a solution for

this strategy, starting with additional

analysis, but also with the feasibility

study on modifications of doors

between cavity and room GA302 to

make spreading much faster and

easier. The second activity was a

feasibility study of an instalation of

the refractory material. It is very

complicated mainly in the cavity due

to need of the concrete cooling (it

must not be influenced) and also from

the perspective of those modification

performace as there is really high

radiation in the cavity, which practically

precludes any work in the cavity.

The updated measurement of the

radiation in the cavity (in 2017) not

only confirmed the level of radiation,

but it was more detailed and identified

sources, which is not only the

RPV itself, but also the thermal and

biological shielding near the cavity

walls. So it is practically impossible

to work there and this shielding was

expected to be replaced with new

one together with refractory layer

installation. So the situation in the

cavity is recently not yet solved and

will need an alternative solution,

which is not yet prepared.

Concerning the refractory material,

the preliminary study on potential

candidates was performed, but before

the final decision the experimental

testing has to be performed.

3.4.5 Stabilization of

containment conditions

The title of this chapter differs in the

wording with the bullet above, but its

meaning is same, because the main

task of the activities is to maintain the

containment pressure and temperature

on an appropriate level for long

time. During this year, the project

opened this task and the proposal

for technical solutions of the heat

removal from the containment

atmosphere are under development.

About nine technical proposals were

prepared as basic scheme and two of

them are be selected for the feasibility

study. Those nine solutions use

various approaches – spraying, steam

condensation and heat removal,

using sophisticated systems with

supercritical CO 2 and so on.

4 Conclusions

A continuous process of safety

enhancement of VVER units in the

Czech Republic has been briefly

described including a presentation of

important milestones and examples of

particular safety measures already

implemented. Despite the high level

of safety reached mainly by the

preventive means in the last decades

at both VVER sites in the context

of Fukushima accident, a new period

of enhancement process has been

initiated following the stress tests

exercise. Use of advanced deterministic

analytical approaches and

implementation of new preventive

safety measures has then taken place

together with special attention to

mitigative part of potential accidents

and relevant strategies and measures.

In the paper a special attention was

given to the evaluation and implementation

of safety measures following

stress tests conclusions and R&D

activities supporting this process. As

examples an implementation of the

„design extension condition without

core melt“ concept and various activities

related to severe accident mitigation

strategies are presented in more

detailed way.

Within the systematic approach to

enhance safety also the PSA has been

recognized as a very useful tool. The

PSA has been mainly used for identification

of weak points in design and

operation and for an evaluation

and prioritization of potential safety

measures. The example of new cooling

tower measure for the Dukovany

site has been used to demonstrate this

approach.

The work on DEC-A (previously

BDBA) safety analyses for Czech NPPs

was described as a consequence of the

Periodical Safety Review (PSR) after

20 year of the operation of the

Dukovany NPP. This effort has been

influenced also by initiatives and

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suggestions from European Utility

Requirements (EUR), WENRA safety

reference levels, and IAEA introduction

of the DEC term and concept into

the safety standards series. Examples

of a use of this approach for Dukovany

and Temelín safety analyses have been

presented.

The programme for severe accidents

for Czech NPPs is very extensive,

this contribution included mainly the

activities related to the Temelín NPP,

but also the Dukovany NPP has own

program and activities related to the

topic of severe accident. The common

for both NPPs and not yet mentioned

is the training of the staff. The UJV

supports both NPPs for longer time

with special lessons on the progress of

the severe accident and impact of an

application of measures to the SA

course. Recently the new tool for the

training is close to the completion, it is

named VINSAP (Visualization of NPP

Severe Accident Progress for Training

on SAM), the project is sponsored

by the Technology Agency of Czech

Republic (project no. TH01011086).

References

[A]

National Report on “Stress Tests” of NPP

Dukovany and NPP Temelín, Czech

Republic, December 2011.

[B]

[C]

Authors

National Action Plan (NAcP) on

Strenghtening Nuclear Safety of

Nuclear Facilities in the Czech Republic,

State Office for Nuclear Safety, rev2,

Jan 6, 2015.

SUJB directive BN-JB-1.7, Selection and

Assessment of Design and Beyond

Design Events and Risks for Nuclear

Power Plants, 2010.

J. Duspiva

J. Holy

P. Kral

M. Patrik

UJV Rez, a.s.

Hlavni 130

25068 Rez, Czech Republic

E. Hofmann

CEZ, a.s.

Duhova 2

14000 Prague, Czech Republic

Applications of Underwater-Robotics

in Nuclear Power Plants

OPERATION AND NEW BUILD 305

Gunnar Fenzel, Dr. Dietmar Nieder and Alexandra Sykora

1 Research project AZURo Cutting and packing of the reactor pressure vessel (RPV) is one important step

during decommissioning of nuclear power plants. The RPV and its internals are radiological activated caused by the

long standing neutron flux.

In particular the internals which –

amongst others – retained the fuel assemblies

have to be cut and packed

under water due to their high radiological

activity. In the past this was

largely done manually using remote

handled tools (such as rods, grippers

and cranes). The operation of the

remote handled tools is time consuming

and enables access to the respective

parts by one direction only. Thus

the accessibility is strongly restricted

slowing down the progress of the

work. Hence the costs of the decommissioning

are highly increased. In

addition, the risk of failures is

enlarged by the inflexibility of the

tools. Moreover and due to the

complex proceeding, the workers

are exposed to a certain radiation

level leading finally to an averaged

increased radiation exposure.

Therefore, it was the objective of

the research project Automated Cutting

of Reactor Pressure Vessels Internals

Using Underwater-Robotics (AZURo) to

(semi-) automate frequently repeated

activities by an underwater robot.

This joint research project was

sponsored by the German Federal

Ministry of Education and Research

(BMBF). It was executed together with

Fraunhofer-Einrichtung für Gießerei-,

Composite- und Verarbeitungstechnik

IGCV. The project AZURo started in

2012 and was finished in 2016.

The highest degree of innovation is

given in the research of the application

of industrial robot systems under

water and in radiation fields. Thereby

there is no direct contact possible,

neither with the robot itself nor with

the workstation. All works at the

system have to be performed remotely

monitored respectively remotecontrolled.

Key aspects of the development

were remote control of the

system, optical monitoring and the

development of an intuitively

designed simulation ambience with

automated path planning.

Arm of robot

Total mass

max. load

The system supports the operator in

planning and execution as well as in

handling steps but ensures the control

of a human being at all times. Just as

well the nuclear requirements such as

intervention capability, reproduc ibility

and health physics aspects during

developing have to be con sidered.

2 Advantages of AZURo

The (semi-)automation of cutting and

packing activities by means of robots

will lead to

• a reduction of the local radiation

exposure of the involved staff,

• a shortening of the performance

times of cutting and packing of

highly activated components,

• a lowering of costs for such projects,

and

Approx. 1,100 kg

150 kg

Protection class IP 68

Working space (spherically) without tools, max.

Working space (spherically) with tools, max.

| | Tab. 1.

Technical Data.

Ø 2,194 mm

Ø 2,764 mm

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OPERATION AND NEW BUILD 306

| | Fig. 3.

Mounting of AZURo in a nuclear power plant

(Sources: Orano GmbH and RWE Power AG, Biblis).

| | Fig. 1.

Under water test of the modified robot (source: Orano GmbH).

| | Fig. 2.

GUI with integrated simulation surroundings (Source: Orano GmbH).

| | Fig. 4.

GUI with integrated simulation surroundings (Source: Orano GmbH).

• an extensive prevention from handling

failures during decommissioning

of nuclear plants.

3 Development AZURo

3.1 Basics

Demands on the overall system and on

the different subsystems were derived

and defined. A detailed market investigation

based on those requirements

was carried out and a robot including

control system was selected (Table 1).

A detailed risk assessment of

the system was performed in the

framework of FMEA. For this, the

complete system was subdivided

into expedient subsidiary systems

and subsequently into structural components

which were assessed by

means of their risks. Based on this risk

estimation an intervention respectively

recovery concept was established.

In addition, a safety concept with

system-independent collision detection

has been elaborated complying

with the demands on applications in

nuclear power plants.

3.2 Upgrade of the robot

For preparation of its upgrade, the

robot has been investigated with

respect to its underwater capability.

The robot was then suitably adapted

and, subsequently, the water tightness

has been verified by respective underwater

tests (Figure 1).

Mock-ups as well as grippers and

tools were designed in order to test

the reference scenario in practice.

The intervention and retrieval

concepts of the robot were tested

successfully.

3.3 Software and control

ambit (development and

implementation)

The selected software enables the

modelling of the workspace in

simulation surroundings. The workspace

for the field test (c.f. chapter 4)

was modelled and tested successfully.

Superordinate control architecture

was developed and implemented.

The interface to the operator is

represented in the elaborated and

tailored Graphical User Interface

(GUI) (Figure 2).

Derived from the previously

prepared specification, solutions for

the position detection and for the

additionally necessary sensors have

been investigated and an appropriate

camera as well as a master arm have

been selected.

The master arm with developed

force feedback option was integrated

into the control system of the robot.

The communication of additional applications

(such as tool, camera, gripper,

further tools) have been established.

For safety reasons forerunning

collision detection was developed,

reviewed by expert and approved.

Investigations on calibration of the

workspace by means of a laser sensor

as well as mechanically by use of fixed

stops were carried out.

4 Field test in a nuclear

power plant

In cooperation with RWE Power AG

(today RWE Nuclear GmbH) a field

test of the system in the nuclear

plant Biblis was performed (Figure 3

and Figure 4). This field test was

carried out in the spent fuel pool

containing nuclear fuel at a water

depth of 14 meters.

For this the robot system was

qualified with expert’s supervision for

application in a nuclear power plant.

The following qualification topics had

to be demonstrated and verified:

• safekeeping of the fuel assemblies

integrity,

• safekeeping of the fuel pool liner

integrity,

• protection of the robot electronics

from high radiation exposure,

• electrical safe operation of the

device in the spent fuel pool,

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atw Vol. 63 (2018) | Issue 5 ı May

| | Fig. 5.

Sorting of temporarily stored operational waste using AZURo

(Sources: Orano GmbH and RWE Power AG, Biblis).

• ability for intervention in each

situation, and

• health physics aspects.

As a result from the expert’s supervision,

there were no safety-related

concerns against employment of the

remotely operated underwater robot

system AZURo.

The mission of the field test was

divided into 2 packages:

• Sorting of disorderly stored operational

waste (flow restrictors

and absorber element heads) into

appropriate disposal containers

• Segmenting and sorting of core

instrumentation detector fingers

into appropriate disposal con tainers

4.1 Sorting of disorderly

stored operational waste

Sorting of flow restrictors and absorber

element heads was performed

by means of the robot system

(­Figure 5). For this an appropriate

gripper was adapted to the robot

system. Using that tool, one piece

after another from the stock of flow

restrictors and absorber element

heads lying disorderly on the bottom

of the spent fuel pool was picked up.

Each item was then transferred into

the disposal container after successful

separation and dose rate determination.

For this application, the robot

was remotely controlled.

4.2 Segmenting and sorting

of core instrumentation

detectors fingers

The following work steps have been

applied to cut and pack the core

instrumentation detector fingers

(Figure 6):

• For determination of the cutting

length the detector finger was positioned

on the cutting tool installed

at the robot arm. The cutting tool

was then moved down vertically

until the cutting length for this

segment was reached. In doing so,

each finger was simultaneously

tracked according to the movement

of the cutting tool until the

length of the segment was reached.

• Subsequently, the crosscut segment

was cut and clamped by

means of the cutting tool.

• The cut-off segment was then

transferred to an appropriate

disposal container also using the

robot system.

4.3 Field Test in a nuclear

power plant – Summary

The testing in a nuclear power plant

proved the capability of the system

under realistic conditions.

The following significant advantages

of this system for underwater

application in comparison with manually

operated tools could be shown:

• High speeds.

• Exact compliance with the given

step sequences and trajectories;

high degree of safety (guarantee of

the integrity of the spent fuel pool

liner and of the fuel assembly

racks); malfunction caused by

breakdown of component(s) prevented

by the graduated safety concept,

that means by combination of

technical and administrative measures.

Moreover:

• High availability (tightness in a

depth of 14m), in the present case

during an application for a period

of ten weeks.

• The preparatory measures being

initially rather extensive (e.g. for

modelling, programming, arranging

and calibrating of the components

and for system checks) are

compensated during execution and

result in economic advantages (time

and cost savings) if there is a larger

extent of similar material quantities

to be dismantled and packed.

| | Fig. 6.

Segmenting and sorting of core instrumentation detectors fingers using

AZURo (Sources: Orano GmbH and RWE Power AG, Biblis).

5 Future prospects

The developed underwater robot

system AZURo clearly highlighted its

assets during the field test. The elaborated

fundamentals ease the adaption

of the robot system to other applications.

For instance, the system can be

used for handling operations (loading

of machines, charging of disposal

canisters etc.) during decommissioning

of nuclear power plants or service

projects as well. Cutting tools (saws,

cutting nozzles etc.) can be very well

adapted at reasonable effort.

The diversity of facilities is a major

asset of AZURo because of the robot

only being a “tool carrier” and being

able to handle the carried tools

arbitrarily in its workspace.

Acknowledgement

The research project AZURo has been

funded by German Federal Ministry

of Education and Research (BMBF)

(support code 02S9082A).

Post Research

After finalizing the research project

the underwater robot system AZURo

is qualified and ready for use. Orano

GmbH has been contracted for the

cutting and packing of the core waste

and the RPV internals of four NPPs.

First application of AZURo will be in

Brunsbüttel NPP.

AZURo is an innovative dose, time and

cost saving equipment; ready to use!

Authors

Gunnar Fenzel

Alexandra Sykora

Orano GmbH

Henri-Dunant-Str. 50

91058 Erlangen, Germany

Dr. Dietmar Nieder

RWE Power AG

Kraftwerk Biblis

Biblis, Germany

OPERATION AND NEW BUILD 307

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atw Vol. 63 (2018) | Issue 5 ı May

OPERATION AND NEW BUILD 308

360° Raumatlas im Kraftwerk Biblis

Jürgen Kircher

In Deutschland ist seit 2011 der Ausstieg aus der friedlichen Nutzung der Kernenergie politisch beschlossene

Sache. Er stellt Betreiber, Behörden, Gutachter und letztendlich die Ingenieure vor große Herausforderungen. Der

Restbetrieb und der Abbau der stillgelegten Kernkraftwerke muss lückenlos dokumentiert werden.

Ein großes hilfreiches Werkzeug

hierbei ist der Raumatlas. Er macht es

möglich, den Ingenieuren, Technikern

…, die außerhalb der Blöcke des

Kraftwerks arbeiten, die Räume in

Kernkraftwerken in hochauflösenden

360° HDR Bildern darzustellen und

technische Gegebenheiten daraus

abzuleiten.

Seit 2008 begleitet die Jürgen

Kircher Medienproduktion den Raumatlas

im Kernkraftwerk Biblis (Abbildung

1). Bis 2017 wurden hier circa

2.500 Panoramen (300.000 Einzelbilder)

erstellt.

Um aus dem Raumatlas aussagefähige

Daten zu entnehmen und

weiter zu verarbeiten, sind HDR­

Panoramen mit höchster Auflösung

und Präzision ein unabdingbares

Werkzeug.

Es wurde die Anforderung gestellt,

dass bei jeder Beleuchtung eine

Schrift von 1 cm Größe in 4 Meter

Entfernung noch lesbar ist. Die Standpunkte

und Höhen der Panoramen

sind auf 5 mm genau festgelegt, so

dass eine genaue Ableitung der Maße

von Armaturen, Rohren und sämtlichen

räumlichen Gegebenheiten

erfüllt werden konnte (Abbildung 2).

Nachdem die Panoramen überarbeitet

sind, werden sie in den

Raumatlas eingefügt. Je nach Größe

sind teilweise in den einzelnen

Räumen bis zu 15 Panoramaaufnahmen

abrufbar, so dass kein toter

Winkel existiert (Abbildung 3). In

| | Abb. 3.

Panoramen werden in den Raumatlas eingefügt, ohne einen toten Winkel.

| | Abb. 1.

Der Kernkraftwerksstandort Biblis in Hessen.

| | Abb. 2.

HDR-Panorama.

jedem Panorama des Raumatlas

sind zahlreiche weitere Parameter

auswählbar.

Zum Beispiel die Anzeige eines

Radars um den jeweiligen Blickpunkt

des Betrachters im Plan darzustellen,

Hotspots um an markanten Punkten

verschiedene Informationen abzurufen,

Links zu weiteren Bildern,

Zeichnungen, URLs und noch vieles

mehr.

Eine Herausforderung waren

die notwendigen Bedingungen des

Strahlenschutzes. So musste beispielsweise

die Vorgabe eingehalten

werden, alle Aufnahmen automatisch

zu erstellen. Dafür wurde ein automatischer

Panoramakopf der Firma Seitz

mit einer Canon EOS 5D Mark IV

angepasst und über die Jahre stetig

weiterentwickelt. Das System macht

es möglich auch unter extremen

Bedingungen (z.B. Temperaturen bis

65 °C) jederzeit schnell optimale

Ergebnisse zu erzielen.

Das System wurde 2011 vom TÜV

München zertifiziert.

2016 wurde der Beschluss gefasst,

mindestens alle Räume in den

Kontrollbereichen der Blöcke A und B

in Biblis lückenlos vor dem Abbau zu

dokumentieren. Vorgabe war, dass

das Projekt vor dem Beginn des

Abbaus abgeschlossen sein muss.

Operation and New Build

360 Degree Area Atlas in the Biblis Nuclear Power Plant ı Jürgen Kircher


atw Vol. 63 (2018) | Issue 5 ı May

Die Jürgen Kircher Medienproduktion

erhielt von RWE den Auftrag

zwischen Februar und Juni 2017

alle Räume zu dokumentieren. Dies

bedeutet, dass in Summe 2.600

HDR Panoramen (ca. 400.000

Einzelbilder) gemacht worden sind.

Nach Abschluss der Arbeiten belief

sich die Daten menge auf etwa 20 TB

RAW Daten.

Seit Mitte 2017 steht der Rückbau

an, der wieder lückenlos dokumentiert

wird. Eine Historie zu jedem

Raum wird im Raumatlas letztendlich

verfügbar sein. Man wird die Möglichkeit

haben, sich jederzeit hochauflösende

Panoramen vor und nach dem

Abbau mit den entsprechenden

Zwischenschritten anzuschauen und

sich ein realistisches Bild von der

Anlage zu machen.

Für die planenden Ingenieure,

Behörden und Gutachtern ist der

Raumatlas ein nicht wegzudenkendes

Werkzeug, welches bei der Planung

Zeit, Geld und Dosis einspart.

ELINDER – European Learning Initiatives

for Nuclear Decommissioning

and Environmental Remediation

Pierre Kockerols, Hans Günther Schneider and Daniela Santopolo

The decommissioning of nuclear facilities is an industrial activity that is expected to grow worldwide, creating

many attractive career opportunities (Figure 1). European industry has acquired know-how and today Europe can

position itself at the top level in the world decommissioning market. However, in view of the expected expansion of the

activities, efforts are necessary to share and enhance the underpinning knowledge, skills and competences and to

ensure the availability of the necessary workforce in the future.

Author

Jürgen Kircher

Medienproduktion

Carl-Zeiss-Str.41

63322 Rödermark Ober-Roden,

Germany

Learning Initiatives for Nuclear

Decom missioning and Environmental

Remediation) and is implemented

from 2018 onwards.

Revised version of

a paper presented

at the Eurosafe, Paris,

France, 6 and 7

November 2017.

309

DECOMMISSIONING AND WASTE MANAGEMENT

| | Fig. 1.

The ESSOR experimental reactor at the JRC-Ispra site.

In this perspective, the European

Commission’s Joint Research Centre

(JRC) investigated the opportunities

for stimulating the development,

coordination and promotion of

adequate education and training

programmes at EU level in nuclear

decommissioning.

Building on the existing experiences

available at European level, the

JRC and partners in the EU decommissioning

field have launched a project

to consolidate and improve existing

training programmes to facilitate

their promotion and enhance the

opportunities they can offer. The

overall aim is to raise the interest of

students and professionals and to

stimulate careers in this important

and emerging field, by offering a

modular, attractive set of theoretical

and practical learning opportunities.

Attention is paid to the international

validity of the qualification and

to a sustainable interaction with

interested industrial actors.

The joint training programme

project is called 'ELINDER' (European

1 Introduction

With the expectations of the decommissioning

market and related new

industrial activities over the coming

decades a clear global positioning of

the European Union will be an asset.

In preparation of future decommissioning

programmes, the availability

of qualified and experienced personnel

will be essential and will be

probably one of the most critical

issues to address. For many years, the

European nuclear sector has faced an

increasing difficulty in recruiting and

maintaining staff with the required

expertise. It can be expected that

the decommissioning industry will

face a similar or even more significant

shortage of competent personnel.

In a joint report with the University

of Birmingham [1], the JRC has

investigated the needs for competences

in nuclear decommissioning,

the existing education and training

opportunities and the ways on how to

attract new talent. It appeared that

some learning initiatives were taken

in European countries over the last

years to improve of knowledge, skills

ELINDER – European Learning Initiatives for Nuclear Decommissioning and Environmental Remediation

Decommissioning and Waste Management

ı Pierre Kockerols, Hans Günther Schneider and Daniela Santopolo


atw Vol. 63 (2018) | Issue 5 ı May

DECOMMISSIONING AND WASTE MANAGEMENT 310

1) A few such examples

are the JRC' Summer

School on Nuclear

Decommissioning

and Waste Management',

the ‘Technology

and Management

of the Decommissioning

of Nuclear

Facilities’ course at

the AREVA Nuclear

Professional School

(Karlsruhe Institute of

Technology (KIT),

Germany), the

Belgian Nuclear

Research Centre

SCK•CEN courses on

'Decommissioning of

Nuclear Installations',

the 'European

Decommissioning

Academy' organised

by the Slovak University

of Technology,

the CEA/INSTN

international course

on 'Dismantling

Experience of Nuclear

Facilities' and

through the IAEA ad

hoc training programmes

and possibilities

for e-learning.

2) ECVET or the

'European Credit

system for

Vocational Education

and Training':

http://ec.europa.eu/

education/policy/

vocational-policy/

ecvet_en.htm

and competences in nuclear decommissioning,

going from short professional

induction training programmes

to academic graduate and postgraduate

courses. Time has come to

support, coordinate, develop and

promote them in a joint project.

2 Competeneces in nuclear

decomissioning

Decommissioning programmes are in

general implemented over several

years via a sequence of projects and

activities of different nature (Figure

2). This explains, at least in part, the

variety of the skills required. Those

range from senior site managers,

programme managers and project

managers (including particularly the

speciality area of planning, scheduling

and cost estimating), to engineers

(electromechanical, chemistry, construction,

geology, ..), operational

managers, safety managers (safety

case and licensing), operational and

technical staff (decontamination,

dismantling, waste characterisation,

treatment and transport, maintenance)

and surveillance staff (radiological

protection, safety and security).

Complementary to operational

staff, the usual important number of

diverse contracts and their complexity

will also require relying on competent

contract management and legal

support, as well as the implementation

of an appropriate human

resources management.

In the transition phase from the

closure of an installation to its

de commissioning, part of the competences

can be acquired by professional

conversion of part of facility

operating staff. But experience shows

that companies embarking from operation

to decommissioning face an

| | Fig. 2.

Decommissioning – a sequence of projects

and activities of different nature.

important cultural change. Targets are

changing from operating and

maintaining a facility using known

technologies, to the dismantling and

final demolishment of the installations.

When moving to decommissioning,

the nature of the work

changes significantly and will require

more flexibility. Activities become

mainly project, cross discipline based,

necessitating a broader knowledge,

especially of new technologies. The

available competences do not neatly

map from operations to decommissioning

and new recruitment or

outsourcing is needed.

The usual significant time scales of

decommissioning processes require

that specific attention goes into the

long-term strategic planning of

recruitment and training needs with

an appropriate profile in terms of

both time and scale. Obviously this

approach is essential for the key

disciplines for which currently a

shortage is already experienced or is

forecasted. A clear vision should exist

on possibilities for personal career

development, which can be facilitated

by the various job opportunities that

decommissioning activities can offer.

Mobility of decommissioning

experts is also an important factor to

consider. Decommissioning activities

are implemented sequentially. For

some steps specific competences are

required for a well-defined period,

limited in time (which can extend up

to a few years). There opportunities

exist for teams of decommissioning

experts to move from project to project

thus maximising the cost benefit

return to companies that develop the

skills-base.

3 What are the

education and training

opportunities?

The evolution of nuclear decommissioning

activities over the last decades

has triggered the development of

several programmes, particularly in

the three main 'nuclear' EU countries:

France, Germany and the UK.

Higher education in decommissioning

and waste management is currently

provided through PhD programmes

and dedicated Professorships, through

two-to-three year or postgraduate

taught Masters courses focussed on

decommissioning knowledge, dedicated

modules in decommissioning

integrated in a more general Master

course in nuclear science or nuclear

engineering or Bachelor degrees

with specialisation (about one year)

in decommissioning.

Some programmes allow students

the flexibility to develop also

managerial skills aimed at running

decommissioning projects. It is indeed

essential to develop also non­ technical

skills such as com mercial awareness,

project organisation, communication,

team leadership.

Complementary to those education

programmes which are addressed

to students, several shorter vocational

training programmes exist focussing

on professionals having already a

working experience in the nuclear

field but whose job evolution requires

new competences linked to decommissioning

activities 1 .

Decommissioning activities also

require specific technical skills related

to decontamination, dismantling,

waste treatment, waste measurement,

radiological checks and surveillance,

transport, accountancy, etc... All these

more operational hands-on skills

require ad hoc training but they are,

in general, organised by the industrial

organisations or nuclear research and

training laboratories.

The education and training programmes

are expected to grow to

meet a future demand. This evolution

will require more harmonisation of

the outcomes and further enhancing

the collaboration with all participants

involved in decommissioning (industry,

safety authorities and associated

technical support organisations,

waste management and decommissioning

agencies, research centres).

Academic education programmes

with modules are weighted according

to the ‘ECTS’ credits which importantly

creates a high degree of transparency

across the EU [2]. For vocational

training programmes (aiming

e.g. at specific job qualifications), the

system ECVET 2

is being introduced

for the identification and mutual

recognition of the requested learning

outcomes [3]. Where applicable,

ECVET points are awarded to

learning packages; some of them are

developed for the nuclear domain.

A harmonised definition of the

necessary profiles needed in decommissioning

combined with mutual

recognition schemes across the EU

could support the development of the

adequate training modules and clarify

the learning outcomes.

4 How to attract new

talent?

Trends in the evolution of the nuclear

workforce in Europe have been

analysed by the ‘European Human

Resources Observatory in Nuclear’

Decommissioning and Waste Management

ELINDER – European Learning Initiatives for Nuclear Decommissioning and Environmental Remediation

ı Pierre Kockerols, Hans Günther Schneider and Daniela Santopolo


atw Vol. 63 (2018) | Issue 5 ı May

| | Fig. 3.

Decommissioning – a challenge requiring

knowledge and skills.

(EHRO-N), whose reports are published

periodically [4]. EHRO-N

statistics show that the number of

students graduating in nuclear­ related

disciplines has slightly increased over

the last five years. The workforce of

nuclear educated staff involved in

decommissioning and waste management

activities represents today only a

fraction (< 20 %) of the total nuclear

employment. Most of the human

resources are dedicated to the operation

of nuclear facilities, to R&D and

to design purposes; decommissioning

is still a ‘niche’ activity in the entire

nuclear business.

At a first glance, undertaking a

career in decommissioning could be

perceived as not particularly exciting;

at face value it involves mainly

clearing, cleaning and demolishing

of reactors and facilities. This is often

seen as less attractive than constructing

something new. However,

the finality of decommissioning is

material recycling and environmental

plus economic valuation, once a site is

cleaned and can be released from

regulatory control and reused for

other purposes. Decommissioning can

be challenge/problem led, due to

the variety of issues to be resolved,

requiring the mastery of a diverse set

of knowledge and skills, with the

development of a bespoke set of

solutions (Figure 3).

However, the many possibilities

offered to study and to start a career

in nuclear decommissioning are

presently not visible enough. In addition,

the on-going decommissioning

programmes and the difficulties they

face are in general presented too

negatively, instead of highlighting

the achievements made so far. Promotion

could start by clarifying the

existing education, training and

career opportunities in Europe.

Advertising the challenge and excitement

linked to decommissioning

could be stimulated and integrated

within existing campaigns for the

promotion of education and training.

And more generally, promotion of

decommissioning could be helped by

improving the public understanding

on its finality and as such presenting

the activity in a more objective way.

5 Elinder

The overall aim of the present

ELINDER 3 project to start as of 2018 is

to raise the interest of students and

professionals and to stimulate careers

in this important and emerging

field, by offering a set of attractive

theoretical and practical learning

opportunities. In this sense ELINDER

envisages to elaborate and promote a

modular, coherent and commonly

qualified training programme in

nuclear decommissioning and pave

the way for an ECVET application to

well-defined decommissioning job

profiles.

The target audience for ELINDER

are students at the end of their

education cycle, young professionals

at the start of their career and experienced

professionals and managers

who change their career orientation

towards nuclear decommissioning.

The programme will be achieved

by pooling and enhancing already

existing learning initiatives of different

European partners active in the

field of decommissioning.

5.1 ELINDER partners

and actors

The ELINDER concept has been

created by the European Commission's

Joint Research Centre in collaboration

with European universities

and institutes (Table 1).

The IAEA has been participating

from the beginning to the elaboration

of the concept and will practically

support the ELINDER project through

its networks of experts, training tools

and its technical cooperation programme.

The development of the courses

and the coordination of training will

be led by the partners, but interested

industrial actors and organisations

will be invited for providing ad hoc

lectures or practical case studies and

exercises. Periodically, a round table

will be organised to assess the lessons

learnt from the programme in view of

its continuous improvement.

5.2 ELINDER approach

The project is conceived as an

integrated set of learning opportunities.

The training programme consists

of a series of courses including

lectures, practical hands-on exercises

at the premises of the organising

partners and visits to relevant facilities

in the vicinity (Table 2). Five

' generic training modules' serve as a

general introduction with a synopsis

of the main decommissioning aspects.

They are addressed to different audiences

of starting professionals and

one of them to students. Additionally

'specific, topical training modules'

address more in detail specialised

topics which have been identified as

pinch-point areas, i.e. areas in which

knowledge, skills and competences

need to be improved.

A complementary e-learning programme

is in elaboration and will be

based on existing courses. It will allow

an introduction of students and

professionals with a view to participate

to future ELINDER trainings.

The most relevant opportunity

will depend on the professional

experience of each participant as

3) ELINDER or

European Learning

Initiatives for

Nuclear Decommissioning

and

Environmental

Remediation, see

webpage: https://

ec.europa.eu/

jrc/en/trainingprogramme/elinder

SCK•CEN Studiecentrum voor Kernenergie – Centre d'Etude de l'Energie Nucléaire Belgium

KIT Karlsruher Institut für Technologie Germany

CEA Commissariat à l’énergie atomique et aux énergies alternatives France

UoB University of Birmingham UK

STUBA Slovenská Technická Univerzita v Bratislave Slovakia

UT University of Tartu Estonia

NUVIA NUVIA France

SOGIN Società Gestione Impianti Nucleari Italy

ENEN European Nuclear Education Network association Europe

ENSTTI European Nuclear Safety Training & Tutoring Institute Europe

ENS European Nuclear Society Europe

FORATOM European Atomic Forum Europe

JRC European Commission Joint Research Centre EU

| | Tab. 1.

ELINDER Partners.

DECOMMISSIONING AND WASTE MANAGEMENT 311

ELINDER – European Learning Initiatives for Nuclear Decommissioning and Environmental Remediation

Decommissioning and Waste Management

ı Pierre Kockerols, Hans Günther Schneider and Daniela Santopolo


atw Vol. 63 (2018) | Issue 5 ı May

DECOMMISSIONING AND WASTE MANAGEMENT 312

e-learning, induction

'generic' training course

as introduction

specific, topical courses

for specialisation

| | Tab. 2.

ELINDER Decommissioning Training Modules.

Basics on nuclear industrial applications and radiation safety

Overview of: regulation and standards, status of the play, experience

feedback, waste management, technical and organisational issues, radiation

safety, stakeholder involvement

Decommissioning planning and cost assessment

Licensing and environmental impact assessment

Programme and project management

Decommissioning Safety

Waste and material management

Decontamination and Dismantling techniques

Metrology for Waste Characterisation and Clearance

Environmental remediation and site release

well as on her/his actual and targeted

level of knowledge, skills and competences.

5.3 ELINDER qualification

To ensure a coherent and harmonised

approach, shared minimum quality

criteria including learning outcomes

will be defined for acceptance of the

course modules within the ELINDER

programme and receiving the

“ELINDER stamp” (Figure 4).

In a next step, the programme will

be aligned to enable the certification

of specific job profiles in nuclear

decommissioning, following the

ECVET credit system.

6 Conclusions

The nuclear decommissioning business

is expected to grow in the coming

decades. The delivery of related

education and training programmes is

still at an early stage of development.

The opportunity could be taken to

harmonise quality criteria and learning

outcomes, which allows more

transparency for the industrial actors

but also facilitate the promotion of

competences in this field.

The ELINDER project aims to

provide a European answer to these

prospects. The programme has been

prepared with a variety of experienced

partners and with the IAEA,

and is actually implemented from

2018 on.

References

[1] Education and Training in Decommissioning

– Needs, Opportunities and

Challenges for Europe,

ISBN 978-92-79-51836-2 (2015).

[2] ECTS Users' Guide, European Commission,

ISBN 978-92-79-09728-7 (2009).

[3] Recommendation of the European

Parliament and of the Council on the

establishment of a European Credit

System for Vocational Education and

Training, 2009/C 155/02, O.J. of the EU

dd. 8.7.2009.

[4] Top-Down Workforce Demand from

Energy Scenarios: Sensitivity Analysis,

European Commission (2016).

Authors

Pierre Kockerols

Hans Günther Schneider

Daniela Santopolo

EUROPEAN COMMISSION

Joint Research Centre

21, Rue Champ de Mars

1049 Bruxelles, Belgium

The New CASTOR® geo – A Comprehensive

Solution For Transport and

Storage of Spent Nuclear Fuel, MOX

and Damaged Fuel

Linus Bettermann and Roland Hüggenberg

Dry interim storage has become a common solution for the disposal of spent fuel in recent years worldwide.

However, in particular the complete defueling of NPP prior to decommissioning and dismantling will dramatically

increase the demand especially for non-standard fuel. Here we present the new dry storage system by GNS for

international markets with its capability to also store MOX and damaged spent fuel.

Introduction

Dry interim storage systems for spent

fuel assemblies have been in use

worldwide for more than three

decades by now. Starting with the first

CASTOR® dry storage systems by GNS

in the early 80s, this proven and

reliable technology has enhanced the

safe storage of spent fuel in countless

NPP worldwide. More than 1300

CASTOR® casks have been loaded and

safely stored over the past decades all

over the world, including Germany,

the US, South Africa and several

eastern European countries. This

made the CASTOR® cask system a

well known and internationally established

trademark for the safe transport

and storage of spent nuclear fuel

and high-level waste.

The sound operational record is

backed by a common design philosophy

that remained unchanged for

various CASTOR® cask types. The

casks feature a monolithic cask

body made of ductile cast iron with

machined cooling fins to improve the

heat dissipation. Neutron shielding is

provided by means of polyethylene

neutron moderators, filled in drilled

bore holes in the casks wall. This is a

major benefit in safety compared to

neutron moderators that are attached

to the outside of the cask wall, when

it comes to thermal accidents. The

CASTOR® casks are closed by a bolted

double lid system. Both independent

lids are sealed with metal gaskets that

are suitable for longterm interim

storage. During storage both lids are

permanently monitored to observe

leak tightness. All cask components

Decommissioning and Waste Management

The New CASTOR® geo – A Compre hensive Solution For Transport and Storage of Spent Nuclear Fuel, MOX and Damaged Fuel ı Linus Bettermann and Roland Hüggenberg


atw Vol. 63 (2018) | Issue 5 ı May

DECOMMISSIONING AND WASTE MANAGEMENT 314

| | Fig. 1.

Different types of CASTOR® geo casks.

a) CASTOR® geo21B for 21 PWR FA b) CASTOR® geo24B for 24 PWR FA including up to 8 MOX FA

c) CASTOR® geo32CH for 32 PWR FA including up to 8 MOX FA d) CASTOR® geo69 for 69 BWR FA including up to 16 MOX FA.

score with the complete absence of

any welding seams, a potential weak

link in many other dry storage systems

on the market.

Another striking advantage of the

CASTOR® design philosophy is its true

dual-purpose design. The casks do not

require any additional overpack for

storage or transportation but rather

fulfill the demands on shielding and

safe enclosure set for a dual purpose

system by itself. The only difference

between storage and transport is

the surveillance of leak tightness

(in case of storage) and the shock

absorbers (in case of transportation),

respectively.

However, the existing designs of

the established CASTOR® types and

especially its fuel baskets for the

spent-fuel assemblies (FA) were

mainly driven by extreme boundaries

and requirements in GNS' German

home market. The larger geometric

dimensions of the German PWR-FA

and higher burnups in typical German

NPPs in combination with requested

short cooling times, limited the

number of FA per cask. This limitation

is very much in contrast to the internationally

increasing demand for

storage systems with larger capacities

in terms of accommodated FA per

cask.

Facing the challenge of the demand

of higher FA capacities per cask, GNS

introduces a new transport and

storage system, the CASTOR® geo.

The CASTOR® geo includes all the

well known and established safety

features of the existing CASTOR®

systems, while it is able to accommodate

a significantly higher number of

FA per cask.

The new CASTOR® geo

The new CASTOR® geo cask system is

a product line based on standardized

modules and components featuring

different cask dimensions and basket

designs. The cask system is designed

to meet the individual requirements

of customers worldwide rather than

focusing on the German market.

CASTOR® geo casks are designed for

storage and transport of both PWR

and BWR FA.

A high degree of standardization

between the different cask types of

the CASTOR® geo system allows for

savings in terms of time and funds

especially for licensing. Even though

different regulators will still review

the respective documents independently

and separately, major parts

of the safety cases remain unchanged.

The approach of standardization also

yields to savings for the equipment

needed for handling and dispatch of

the casks and for training of the

personnel. Finally the weights of the

individual cask types are optimized according

to internationally established

crane capacities and can be further

customized to individual needs.

In late 2016 GNS and Kernkraftwerk

Gösgen-Däniken AG have signed

a contract for the development and

manufacturing of CASTOR® geo32CH

casks and associated equipment. After

the approval process in Switzerland,

the up to 51 casks will be manufactured

at the GNS facility in Mülheim/

Ruhr.

Also in late 2016 GNS and Synatom,

a subsidiary of the Belgian ENGIE

Electrabel, have signed a contract for

the development, licensing, and

manufacturing of 30 transport and

storage casks of the type CASTOR®

geo24B and CASTOR® geo21B. From

2021, the casks will be delivered to

the Belgian nuclear power stations of

Doel and Tihange. The contract also

includes the option for further casks

to serve the future demand for storage

casks until 2030.

Actual examples of different

cask types

CASTOR® geo casks are able to

accommodate up to 37 PWR-FA or 69

BWR-FA respectively with a maximum

initial enrichment of approx. 5 wt-%

235 U, and more than 40 kW heat load.

At the moment there are four

different types of the CASTOR® geo

system in the development or licensing

process, respectively. Three different

cask types for PWR reactors and

one cask type for BWR reactors, customized

for the specific needs, while

still based on standardized modules

and components.

The three different types for PWR

CASTOR® geo casks will be utilized

in six reactors of two European

countries. The first of the new PWR

CASTOR® geo casks is the CASTOR®

geo24B (Figure 1) for Tihange NPP

1&2 and Doel NPP 3. The cask is able

to take up 24 FA of which a maximum

of 8 FA might be MOX-fuel. The initial

enrichment of the fuel is 4.5 wt-%

235 U and 7.7 wt-% Pu fiss (Pu+U)

respectively. The cask features a

maximum average burn-up of

55 GWd/MTU and a maximum heat

load of 33 kW. The maximum mass

of the cask during handling inside

the reactor filled with water is

117 Mg. The mass in transport configuration

is somewhat higher due to

the attached shock absorbers. It is

134 Mg.

The second cask in the series of

new CASTOR® geo casks is the

CASTOR® geo21B (Figure 1) for

Tihange NPP 3 and Doel NPP 4. This

cask is somewhat longer and slimmer

compared to the CASTOR® geo24B to

be able to accommodate longer FA. To

remain at the same handling and

transport masses the cask takes up 21

FA, thus slightly less FA per cask. The

nuclear parameters are also adjusted

to the customer specific needs. The

initial enrichment of the fuel is

4.4 wt-% 235 U with a maximum

average burn-up of 55 GWd/MTU and

a maximum heat load of 29 kW.

The third of these new PWR casks

is the CASTOR® geo32CH (Figure 1)

for Gösgen NPP in Switzerland. The

cask accommodates 32 PWR FA

with a maximum of 8 MOX-FA. The

Decommissioning and Waste Management

The New CASTOR® geo – A Compre hensive Solution For Transport and Storage of Spent Nuclear Fuel, MOX and Damaged Fuel ı Linus Bettermann and Roland Hüggenberg


atw Vol. 63 (2018) | Issue 5 ı May

| | Fig. 2.

A CASTOR® geo24B cask with shock absorbers. The shock absorbers for the CASTOR® geo casks are

generally comparable and suitable for the different categories of transportation.

initial enrichments are 5 wt-% 235 U

and 4.8 wt-% Pufiss (Pu+U) respectively

with a maximum heat load of

35 KW. The maximum mass in transport

configuration is 150 Mg, while it

is 135 Mg in storage configuration.

Besides the new casks for PWR fuel

assemblies GNS currently designs a

new cask for BWR fuel particularly for

Asian customers. The CASTOR® geo69

will be able to accommodate 69 BWR

fuel assemblies including up to 16

MOX FA.

Transport

The transport of the casks of the

CASTOR® geo cask system is similar to

the established transportation of the

previous CASTOR® casks. Again the

casks are generally able to be transported

by road, rail, inland waters and

sea. Two shock absorbers attached to

the top and bottom respectively form

the transport package in combination

with the cask itself (Figure 2). There

are no other auxiliaries necessary

for the formation of the transport

package. Related equipment such as

transport frames etc. are designed in

accordance to with the respective

transport vehicle.

same burn-up. Decay heats are

roughly two times higher and neutron

source strengths even up to seven

times higher. The maximum number

of eight MOX-FA in the CASTOR®

geo24B cask type is therefore on the

edge of the physically possible configurations.

However, the exact take up of MOX

fuel is dependent on the customerspecific

needs and might be slightly

adjusted.

Damaged spent fuel disposal

Complete defueling is a prerequisite

for decommissioning and dismantling

of NPPs. In particular the defueling

of damaged spent fuel and the

following dry storage remains a

challenge, since most of the damaged

spent fuel rods have been collected

during the NPPs lifetime in the spent

fuel pools. It was the aim to complement

the existing dry storage

technology for intact fuel assemblies

with a comprehensive solution for

damaged fuel rods comprising

concepts for transport as well as

for storage

| | Fig. 3.

GNS IQ. A Quiver for dry storage and transport

damaged spent fuel of PWR and BWR reactors.

Therefore GNS developed the

Integrated Quiver System “GNS IQ”

(Figure 3) for damaged spent fuel

which complies with the requirements

of the transport and storage cask

CASTOR® V (for the German market)

and CASTOR® geo (for the international

market). The dimensions of

the Quivers allows it to load them

directly into the slots of the fuel

baskets of the casks. The Quivers are

designed like a “second cladding” and

it offers especially leakers a way to be

re-dried inside the Quiver. However, it

was the aim of the Quiver project to

develop a one-fits-it-all solution for all

different kinds of damaged spent

fuel, therefore other defects like

deformations of rods, buckling of the

DECOMMISSIONING AND WASTE MANAGEMENT 315

Mox fuel disposal

Since many dry storage systems on the

market are not licensed for MOX fuel,

the shortage of systems capable to

take up MOX-FA remains a challenge

for many utilities. Therefore the

existing CASTOR® design has two

major advantages over many other

dry storage systems. In addition to

standard UOX spent fuel it is also able

to take up MOX fuel from reprocessing

plants. This achievement can also

be found in the new CASTOR® geo

product line.

The challenges in the storage and

transport of MOX-fuel are higher heat

loads and higher neutron source

strengths compared to UOX-fuel of the

| | Fig. 3.

Schematic dispatch of the Quiver on the reactor floor level.

a): The damaged rods are loaded under water into the body of the Quiver, which is subsequently

transferred into a so-called primary shielding that allows the Quiver to be taken out of the pool.

After that the Quiver inside the primary shielding is lifted out of the pool into the handling station

on the reactor floor. b) Dewatering of the Quiver inside the handling station. c) Subsequently a

simplified hot cell is attached to the handling station. In there the drying and the welding of the Quiver

take place. All work is carried out fully remote controlled. The picture provides a view into the simplified

hot cell on top of the handling station.

Decommissioning and Waste Management

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DECOMMISSIONING AND WASTE MANAGEMENT 316

FA skeleton or damages to the foot and

head pieces or even certain kind of

debris can be accommodated by the

Quiver.

In contrast to the regular dispatch

and loading of spent fuel assemblies

under water in the spent fuel pool, the

dispatch of the Quiver is performed

outside the spent fuel pool on

the reactor floor. This approach is

motivated by the use of a much

simpler technology and increase in

process stability, than it would be

required if processing and especially

drying and welding is done under

water in the spent fuel pool. This also

yields an increase in process stability.

However, this approach requires some

addi tional equipment especially in

terms of shielding. Figure 4 describes

the dispatch i.e. the drying of the fuel

and the closure of the Quiver by

means of welding in general. The

dispatch of a Quiver is assumed to

last not longer than one week and

results in a collective dose of less

than 3,7 mSV including independent

inspectors.

The Quiver for the PWR cask

CASTOR V/19 has received its Type B

transport license in spring 2017 and

its first storage license at an interim

storage facility of a German NPP is

expected in spring 2018. However, the

first hot loadings and internal transport

between different blocks of a

multi-block NPP already took place in

summer and fall 2016, respectively.

Conclusion

With the development of the new

CASTOR® geo cask system GNS provides

state-of-the-art high capacity

dry storage cask system for costumers

worldwide. In Combination with

the newly developed GNS IQ Quiver

system, it provides a comprehensive

solution for the dry storage of

spent UOX fuel, MOX fuel and even

damaged fuel.

Authors

Linus Bettermann

Roland Hüggenberg

GNS Gesellschaft für Nuklear-

Service mbH

Frohnhauser Straße 67

45127 Essen, Germany

Optimal Holistic Disposal Planning

– Development of a Calculation Tool –

Johannes Schubert, Anton Philipp Anthofer and Max Schreier

1 Optimisation potential of disposal planning The expected volume of radioactive waste from

dismantling of nuclear facilities in the forthcoming scope and the opening of the Konrad disposal requires an optimised

planning of the removal of radioactive waste. For the treatment of radioactive raw waste, with negligible heat

generation, different conditioning processes are available. Thereby different waste volumes and masses with different

properties can result even from the same raw waste. For final storage, each container has to be filled completely

according to the repository conditions of the approved repository site Konrad. There are different variants available to

combine the number of barrels and type of container. For example if 100 barrels should be packaged into ten containers

there are 1.7x10 13 possibilities of combination if the type of container is chosen before. In addition there are variants of

container load with bulk material and huge possibilities how components can be fixed in different types of container.

These packaging variants can also be combined.

For each variant, compliance to the

final storage conditions regarding to

the criteria radiology, material,

volume and mass must be checked.

Furthermore there are boundary

conditions like missing places for

| | Fig. 1.

Workflow of calculation for optimisation of repository container packaging.

handling, missing conditioning facilities

or design of the site for using only

specific containers.

An optimisation can be realised

according to the parameters repository

volume, radiological utilization of

the containers, exposure time and container

costs. Moreover an optimisation

of container loading requires a comparison

of the loading variants regarding

to the optimisation parameters.

An optimisation of disposal and

packaging planning saves repository

volume, time, costs and exposure time

for the staff. Therefore it is indispensable

for an integrated disposal planning

in the forthcoming volume [1].

Accordingly, for existing waste, all

conditioning and loading variants

must be calculated with all criteria

like radiology, mass, volume, exposure

time and costs. A combination

of different loading variants results

in iteration loops to find an optimal

solution. This workflow is shown in

Figure 1.

For the planning of logistics, handling

and shielding for determining

the dose rate for the executive staff in

individual handling steps, established

simulation programs are available.

The consideration of the change of

Decommissioning and Waste Management

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| | Fig. 2.

Parts of disposal planning, which can be replaced by a calculation tool.

waste properties by conditioning

processes and the possible combinations

of packaging variants resulting

therefrom are currently calculated

manually. This complex process can

be carried out by a calculation tool.

With the data obtained at the removal

planning, the calculation tool can

carry out the planning and optimisation

of conditioning and packaging

and supports a repository documentation.

Therefore the calculation tool

supports the planning, optimisation

and calculation of packaging according

to the final storage conditions and

prepares and simplify the repository

documentation. This workflow is

shown in Figure 2.

2 Characteristics of

packaging planning

During the post-operational phase of

a nuclear installation, dismantling can

be planned. This implies the planning

of the dismantling as well as the planning

of conditioning and packaging of

the radioactive waste with a final

disposal documentation. When disassembling,

statements about the

properties of the waste can be made.

The required conditioning processes

are dependent of the material properties.

These properties such as volume,

mass, state of matter and flammability

will be changed by conditioning processes.

For example, by high-pressure

compression, the volume of raw waste

can be reduced by up to 80 %, using

incineration a reduction by 98 %

can be achieved and by a combination

of high-pressure compression and

incineration, the waste can be reduced

by up to 99 % [2]. By reducing the

volume, the radioactivity is concentrated.

Depending on raw waste and

conditioning process, different volumes

of radioactive waste with

different properties result. This is

crucial for packaging planning. The

needed parameters for final storage of

the waste results of material analyses

and calculations. For all conditioning

processes a qualification is necessary.

Therefore evidences for the realization

of the conditioning according to

the given restrictions and corresponding

conditioning systems at the site

are needed.

Moreover, there are various types

of containers available in various

categories for the final disposal packaging

of radioactive waste. Furthermore,

restrictions in terms of mass,

volume, radiology and other waste

properties are given in the final

disposal conditions [3]. These restrictions

must be checked for each

container. The evidence for the permissibility

of the used containers is

also required. This could be implemented

by manufacture certificates

and handling instructions.

Further influencing factors for

choosing the type of container can be

given by local boundary conditions of

the site. Equipment for handling of only

a specific type of container without the

possibility to adapt the transport system

to another type of container can be

such an example. The available storage

area inside of a site can be a logistic

challenge which has to be accounted.

Compliance to the transport regulations

has to be given at every time

inside the site and during transport.

These restriction parameters for

packaging which are necessary to be

taken into account by a calculation

tool for holistic waste management

planning are shown in Figure 3.

3 Development of the

calculation tool

The calculation tool has a modular

structure. In individual modules, the

conditioning methods are determined,

the change in waste properties

such as volume and mass is calculated,

the locally available conditioning

procedures are determined and

compared with the required procedures,

loading time and equivalent

dose are estimated and the compliance

of the disposal conditions

regarding volume, mass and radiology

is checked. The modules are illustrated

in Figure 4.

From the calculation results of

the individual modules, the optimal

loading variant is determined and

entered into the waste data sheet in

accordance with the selected optimisation

parameter like repository

volume, loading time, container costs,

volume utilization of the last container

or radiological utilization. This

is realised by the main module, where

all information from the separate

modules are evaluated.

The user interface of the developed

calculation tool consists of an input

mask for waste-, conditioning- and

container data and an output,

where the optimal loading variant is

described.

In addition, a waste data sheet

is created, where all information determined

by the calculation tool are

inserted automatically. The individual

details, e.g. for description of the

included material or dose rate are

| | Fig. 3.

Characteristics of packaging planning which must be accounted by a calculation tool for holistic disposal planning.

DECOMMISSIONING AND WASTE MANAGEMENT 317

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DECOMMISSIONING AND WASTE MANAGEMENT 318

| | Fig. 4.

Modules of the Bethoven calculation tool.

| | Fig. 5.

Logo of the calculation tool “Bethoven”.

provided with instructions, in which

the required evidence for the repository

documentation are described.

The acronym for the calculation

tool for holistic waste management is

Bethoven according to Figure 5.

4 Validation and discussion

Extrapolations have been realized

according to the optimisation parameters

repository volume, container

costs, loading time and volume utilization

of the last container.

The validation of the calculation

tool was based on existing final

storage containers with complete and

certified final disposal documentation

and certified suitable for disposal, e. g.

[4] and [5] as well as on raw waste

data from national disposal program

(NaPro) [6].

The validation of the calculation

tool has been successful and the

extrapolations give meaningful results

that illustrate the potential of optimising

packaging planning.

The input data can be taken from

the planning of dismantling and

material analysis as single data

or from software like waste flow

tracking and product control system

(AVK) or the residue-tracking control

program (ReVK). The data of a loaded

final storage container can also be

the data input for Bethoven to create

the waste data sheet. In the mentioned

order, the number of data is

decreasing.

Bethoven also can be used to estimate

number and type of necessary

containers to pack a certain amount of

waste according to an optimisation

parameter. Another function of the

calculation tool is to give a statement

about the general packability of waste

according to the Konrad disposal

conditions and handling possibilities

at the site. For detail planning and

optimisation the Bethoven calculation

tool can give specifications like costs

and exposure time for the executive

staff. As highest level of detail, the tool

is able to check the packability of

waste with respect to a particular

container. For that the calculation tool

gives a list including the parameters

for the package variant according

to the Konrad disposal conditions

and a waste data sheet. An overview

about these classifications is given in

Figure 6.

The validation of Bethoven demonstrates

the range of possibilities

for optimisation of packaging by using

a calculation tool for packaging

planning. For individual waste packages,

the calculation tool is able to

generate a detailed waste data sheet.

For locations with small amounts of

radioactive waste, optimising the

volume utilization of the last container

is expedient and for large

amounts of waste. Bethoven can be

used to optimise the packaging

according to disposal volume, loading

time or container costs.

5 Summary and outlook

Optimisation of packing radioactive

waste is necessary with regard to the

volume limit of the repository and the

exposure times for the executive staff.

For the development of the calculation

tool, the planning of waste

management for radioactive waste

| | Fig. 6.

Input and output for raw, product and container data.

Decommissioning and Waste Management

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atw Vol. 63 (2018) | Issue 5 ı May

with negligible heat generation has

been considered from the generation

of waste through the conditioning and

packaging of the various waste

streams according to the Konrad

disposal conditions to the final

disposal documentation.

The developed calculation tool

considers the change of the waste

properties by the respective conditioning

procedures and determines

the optimal variant for a container

loading according to the selected

parameter. The restrictions regarding

radiology, volume, mass, conditioning

processes and available technical

facilities at the site are taken into

account and inadmissible loading

variants are excluded by the calculation

tool.

The calculation tool is based on a

modular structure. By using the

different input data, Bethoven creates

an overview about the needed information

to the determined optimal

packaging variant and a waste data

sheet according to the Konrad disposal

conditions.

The calculation tool has been

validated on the basis of data from

already conditioned repository containers.

Furthermore extrapolations

were made for raw waste on the basis

of data from NaPro. The validation

demonstrates the range of possibilities

for optimisation by using a calculation

tool for packaging planning.

With different level of detail, the calculation

tool can be used to optimise

the packaging process.

Perspective, the calculation tool is

to be supplemented with established

planning and calculation tools for

dismantling, handling, logistics and

material flows. This coupling establishes

the calculation tool for using at

holistic disposal planning for the

decommissioning activities of nuclear

facilities and guarantees the optimal

utilization of the resources like time,

exposure time for the executive staff,

costs and available volume of the

Konrad repository.

References

[1] Anthofer, A.; Schubert, J.: Repository

Documentation Rethought. ATW,

Vol. 62 (2017), Issue 11, November.

Page 649-653.

[2] Arbeitskreis Abfallmanagement des

VGB PowerTech e.V.: Entsorgung von

Kernkraftwerken: Eine technisch gelöste

Aufgabe. Essen, 2011.

[3] Bundesamt für Strahlenschutz:

Anforderungen an endzulagernde

radioaktive Abfälle (Endlagerungsbedingungen).

Endlager Konrad,

Fachbereich Sicherheit nuklearer

Entsorgung. Stand: Dezember 2014.

Salzgitter, 2014.

[4] Baumann, R.: Nachweis der Endlagerfähigkeit

von radioaktiven Abfallgebinden,

die nach den (vorläufigen)

Konrad-Endlagerungsbedingungen

(Dez. 1995) hergestellt wurden.

Siemens AG, Symposium Endlagerung

radioaktiver Abfälle. Oktober 2014.

[5] Karbstein, L.; Anthofer, A.; Borchardt,

R.; Reithmeier, H.: Konradgerechte

Konditionierung der ANTARES-Shutterbaugruppe

aus dem FRM II; Kontec 17;

Dresden, 2017.

[6] Bundesministerium für Umwelt,

Naturschutz, Bau und Reaktorsicherheit:

Verzeichnis radioaktiver Abfälle,

Bestand zum 31. Dezember 2013 und

Prognose.

Authors

Dipl.-Ing. Johannes Schubert

Dr.-Ing. Anton Philipp Anthofer

Dipl.-Ing. Max Schreier

VPC GmbH

Fritz-Reuter-Straße 32c

01097 Dresden, Germany

DECOMMISSIONING AND WASTE MANAGEMENT 319

Scope for Thermal Dimensioning

of Disposal Facilities for High-level

Radioactive Waste and Spent Fuel

Joachim Heierli, Helmut Hirsch, Bruno Baltes

Introduction The objective of final disposal of high-level radioactive waste in deep geological formations is to

isolate the radionuclides from the accessible biosphere for a sufficient period of time [IAEA 2011; IAEA 2012]. To achieve

this, both the functionality and the integrity of the disposal system must be assured under ambient conditions that

depend both on the geological environment and on engineering choices taken in the planning of the facility.

In particular, the amplitude of the transient temperature increase caused by the release of nuclear decay heat in the

disposal area is scalable through design strategies and thermal dimensioning.

The trade-offs between hotter and

cooler repository designs are multiple

and complex [Whipple et al. 1999]. In

many ways, the ambient temperature

influences the physical processes

taking place in and around the repository.

For example, hotter designs

delay liquid water contact with

corrodible barrier materials and, with

it, liquid-phase oxidation-reduction

reactions. In some environments, they

cause faster convergence of circumambient

rock [Mönig et al. 2013] or

tend to eliminate the excavation

damage zone as a possible by-pass

for radionuclides to escape [Beswick et

al. 2014]. On the other hand, cooler

designs decrease the rates of thermally

activated diffusion-reaction processes,

induce smaller changes to the natural

system, alleviate pore or crack water

pressures in the circumambient rock

[Gens et al., 2017; Wieczorek 2017]

and are likely easier to analyse [Whipple

et al. 1999]. In geologic repositories

relying on the barrier properties

of smectite-rich materials such as

argillaceous rock or bentonite,

increased peak temperatures in the

hydro-chemical environment raise the

difficulties of substantiating the safety

case with a satisfactory level confidence

[Heierli 2016; Huang et al.

1993]. Pertaining to operational

aspects, high temperatures produce

adverse working conditions for human

or machine operated underground

activities. For example, the retrievability

of waste, either limited or

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DECOMMISSIONING AND WASTE MANAGEMENT 320

Access gallery

Access gallery

disposal area (1 cluster)

waste batches

disposal area (1 cluster)

waste batches

| | Fig. 1.

Spatial layout for (a) configuration C a and (b) configuration C b . The disposal area can be divided in more

than one cluster. Red: center batch of cluster.

unlimited in time, has recently been

considered by the Nuclear Energy

Agency [OECD-NEA 2012] and is

required by national law in an increasing

number of countries (e.g. Finland,

Germany, Switzerland). As temperatures

increase fast after waste

emplacement, the ambient conditions

for waste retrieval give rise to operational

uncertainties, which should be

taken into account in an early stage of

planning [Heierli and Genoni 2017].

Last but not least, it has been pointed

out that hotter repository designs are

intrinsically more complicated and

that their uncertainties in behaviour

are too large to accept [Long and

Ewing 2004; Whipple et al. 1999].

The trade-offs and uncertainties

are to be analysed in safety cases. The

purpose of the safety cases is to

determine whether an adequate level

of confidence in safety can be achieved

and whether the safety criteria can be

fulfilled [IAEA 2012]. An important

aspect hereby is that the temperature

of components remains within boundaries

determined in safety analyses.

Formally, this step is handled by using

criteria of admissibility in the form of

inequalities [e.g. Hökmark et al. 2009;

Eikemeier et al. 2013; Ikonen and Raiko

2012; Jobmann et al. 2016; Kommission

Lagerung hoch radioaktiver Abfallstoffe

2016]. Unilateral criteria do no

bring about a unique solution, however,

but a scope of admissible choices.

To select amongst those, the prevailing

procedure is to take into account

the use of spatial resources underground,

resulting in setting the space

requirements as high as judged necessary

and as low as judged possible.

drifts (n)

(a)

(b)

drifts (n)

D1

D2

D1

D2

Currently, many national disposal

programmes are in the stage of siteselection.

In this stage, site boundaries

are determined under the leadership

of national governments. In order

clarify the challenges of the corresponding

decision-making process,

this contribution explores the interdependence

between spatial and

thermal dimensioning. Temperatureaffecting

design options are parameterised

to evaluate their benefit on

the one side and the engineering

effort to realise that benefit on the

other. It is emphasised that neither the

optimisation of peak temperatures nor

the optimisation of spatial resources

are the primary objectives of nuclear

waste disposal. It is understood

Project leadership

| | Tab. 1.

Baseline configuration of study cases (10,15). SF = “spent fuel.”

P2

20 m

0.88 m

P3

P1

1.5 m

throughout this work that the main

objective is to enhance both the safe

confinement of waste and the confidence

in the functionality of its

elements.

Materials and method

The configuration of a repository for

high-level waste and spent fuel can be

represented by a configuration vector

C = (x 1 , x 2 , …) of engineering parameters

x i (Figure 1). In the present

context, of interest are those that

affect temperatures most. These are:

the cooling time from reactor retrieval

to disposal of the waste (t cool ), the

number of fuel elements per waste

batch (k), the spacing of disposal

drifts (D 1 ), the spacing of batches

within a drift (D 2 ), the number of

disposal drifts (n) and the size of

sub-clusters of batches for disposal

(s). Parameters that are not freely

adjustable by the engineers, such as

the total amount of waste to be

disposed or the depth of the repository,

are not varied in this study.

Let P i be a decision point for the

dimensioning of temperature in a

component indexed i, e.g. the canister

core, the canister surface, the backfill

material, the ambient rock, the

nearest significant aquifer, etc. For

each P i , there exists a decision

criterion T 0 (P i ) + u(t, P i ) < T i , where

T 0 (P i ) is the undisturbed temperature

at P i , u(t, P i ) is the temperature

increase at P i at time t. The right-hand

side term T i is the admissible boundary

temperature for the component

at P i . A configuration C = (n, t cool , k,

D 1 , D 2 , s, …) is considered admissible

if the criterion is satisfied for all P i .

symbol

C a

Nagra

C b

Posiva

Type of host rock (fixed) clay crystalline

Depth of the repository (fixed) 650 m 420 m

SF type (fixed) UO 2 +MOX UO 2

burnup (fixed) 48 MWd/kg 40 MWd/kg

Cooling time of SF t cool 55 y 33 y

Initial average decay power per SF element π 0 (t cool ) 337.5 W 189 W

Number of SF elements for disposal (fixed) F 8748 8100

Number of SF elements per batch k 4 9

Initial average decay power per waste batch p 0 = kπ 0 1350 W 1700 W

Number of disposal drifts n 27 30

Number of batches per disposal drift m 81 30

Number of batches total N= F/k = nm 2187 900

Spacing of disposal drifts D 1 40 m 25 m

Spacing of batches within a drift D 2 7.6 m 8.92 m

Cluster size s 27×81 30×30

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The fulfilment of the admissibility

conditions, however, is not the

endpoint of engineering endeavor.

Supplementary benefits may be

achiev able within reasonable efforts.

It is therefore important to explore

repo sitory configurations beyond

their strict admissibility.

Since every disposal project needs

to fulfil particular national regulations,

it is not possible to address all

situations at once. In the following,

two situations shall be considered as

examples to illustrate the interdependence

between spatial and thermal

dimensioning: (1) A facility

planned for argillaceous rock in

northern Switzerland by the Swiss

National Cooperative for the Disposal

of Radioactive Waste (Nagra), with

baseline configuration C a (Table 1).

This project is currently in the stage

of site-selection under the supervision

of the Swiss Federal Office of Energy.

(2) A facility targeted for crystalline

rock in southern Finland by the

Finnish expert organisation for

nuclear waste management (Posiva

Oy), with baseline configuration C b

(Table 1). This project has been approved

by the Finnish Radiation and

Nuclear Safety Authority.

The thermal dimensioning of a

repository for high-level waste and

spent fuel requires the mutual

comparison of a large number of

engineering configurations comprising

large amounts of individual

heat sources. Therefore, the evolution

of temperature at decision-critical

points P i must be evaluated in reasonable

CPU time. For the purpose of this

study, a computational model based

on the numerical integration of

analytically calculated Green’s functions

for all relevant heat sources has

been implemented [Carslaw and

Jaeger 2011; Myers et al. 2015; Heierli

2016]. The method remains insensitive

to length scales, which greatly

facilitates parameter studies.

The purpose of the Green’s function

model is to predict the temperature

increase u(t,P) within a few

degrees Kelvin at location P and time t

in a homogeneous (but not necessarily

isotropic) volume of the circumambient

rock to a repository of

arbitrary configuration. The model is

not designed to compute the temperature

in engineered components such

as tunnel lining, inside the backfill or

waste containers, nor in remote rock

formations.

The positions of heat sources in

the repository are modelled in 3

dimensions. The (real) heat sources

are supplemented by (imaginary)

images sources to account for isotherm

conditions on a given boundary

surface, e.g. the earth surface in the

present work. The image sources

are sited on the mirror image of

the repository with respect to the

boundary surface. The instant power

release of each image source is equal

and opposite in sign to the power

release of the corresponding heat

source. The heat sources and the

image sources are modelled in

dependence to their distance to P.

Sources sited remotely from the point

of observation are modelled as single

point sources, whereas nearby sources

are modelled as a mesh of points sited

on the surface of the waste canister,

to account for container geometry.

By virtue of the superposition principle,

contributions from the repository

are treated separately from

those originating from geothermal

heat. Aspects of secondary importance

are omitted in the model: It is

assumed that the radioactive decay of

the activated waste represents the

dominating source of heat in the

repository. Other sources or sinks,

such as enthalpy of reactions taking

place underground, initial heat content

of waste canisters, ventilation

and construction of caverns are

neglected. Heat is conservatively assumed

to be transported by conduction.

Thermal properties are assumed

invariant in time. Waste disposal is

assumed to take place instantly, rather

than in a period of one or two decades.

The geometry is assumed to remain

stable over the time range considered.

Further implementation details are

given in Heierli [2016].

The main advantage of the Green’s

function method is that the temperature

at an arbitrary point in

the underground can be computed independently

of the temperature of any

other points at previous times,

resulting in a fast algorithm allowing

for the calculation of a large variety of

configurations. The CPU time for the

simulation of the configurations considered

in this study, calculated over

100’000 y in full 3D for P, range

between 30 s and 1 minute per

Point P Location Significance of P

con figuration. In previous studies, the

Green’s function method has been

compared with results from sophisticated

numerical models and found to

produce unbiased results within 2 °C

standard deviance or 2 % of maximum

value, provided that P is chosen in the

appropriate domain of application

(see above) and that the thermal

properties of the host rock formation

are homogeneous in space [Myers et

al. 2015; Heierli 2016].

Results

The evolution of temperature at

decision points P 1 and P 2 (Figure 1a,

Table 2) has been calculated for the

baseline configuration C a in argillaceous

rock, as well as for a number

of parameter variations thereof. The

results for u(t,P 1 ; C a ) and for u(t,P 2 ;

C a ) are presented in Figures 2 and 3

respectively. The same procedure

was applied for decision point P 3 in

con figuration C b for crystalline rock

(Figure 1b, Table 2). Corresponding

results for u(t,P 3 ; C b ) are presented

in Figure 4. The y-axis in the graphs

designates the change in temperature

at P i . In order to obtain the resulting

temperature, the local undisturbed

temperature has to be added (this is

approximately 38 °C in 650 m depth

in Figure 2 and 37 °C in 630 m depth

in Figure 3). In panels a to f, one and

only one engineering parameter has

been changed with respect to the

baseline configuration. Panel g shows

the temperature for both the baseline

configuration and alternative configurations

with three parameters

changed simultaneously, leading to a

lower and shorter temperature peak.

The uppermost curves in every panel

are identical and represent the temperature

evolution in the baseline

configuration. The dashed curves

represent asymptotic limits. The

arrows indicate the path of the temperature

peak under variation of one

parameter, e.g. cooling time in panel

a. The arrows points towards increasing

technical effort.

Discussion

In the baseline configuration C a for

argillaceous rock, the temperature

P 1 P 0 + (x = 0, y = 1.50, z = 0 ) Retrievabilty, degradation of backfill material.

P 2 P 0 + (0, 0. 20.0) Pore water pressure

P 3 P 0 + (0, 0.88, 0) Retrievability, degradation of bentonite.

| | Tab. 1.

Location of decision points P i relative to the midpoint P 0 of a central batch. The x-axis is taken parallel

to the drift axis, the y-axis perpendicular to x in the plane of the repository and the z-axis is taken

perpendicular to the repository plane.

DECOMMISSIONING AND WASTE MANAGEMENT 321

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atw Vol. 63 (2018) | Issue 5 ı May

(a) Parameter variation: t cool

(b) Parameter variation: k

(c) Parameter variation: D 1

DECOMMISSIONING AND WASTE MANAGEMENT 322

60

50

40

30

20

10

0

60

50

40

30

20

10

0

60

50

40

30

20

10

0

u(t,P) [°C]

55 y

100 y

200 y

300 y

450 y

0 1 10 10 2 10 3 10 4 10 5

u(t,P) [°C]

(d) Parameter variation: D 2

7.6 m

9 m

11 m

15 m

asymp.

0 1 10 10 2 10 3 10 4 10 5

u(t,P) [°C]

(d) Parameter variation: D 2

7.6 m

9 m

11 m

15 m

asymp.

0 1 10 10 2 10 3 10 4 10 5

| | Fig. 2.

Temperature evolution at P = P 1 for the

baseline configuration C a (black) and

alternative configurations (blue).

A-F: One parameter changed (parameter

in title); G: Three parameters changed.

Host rock composition: opalinus clay (argile)

with anisotropic heat conductance 1.2 W/mK

(vertical), 2.15 W/mK (horizontal), heat

capacity 2.3 MJ/m 3 K.

60

50

40

30

20

10

0

60

50

40

30

20

10

0

u(t,P) [°C]

k = 4

k = 3

k = 2

k = 1

0 1 10 10 2 10 3 10 4 10 5

u(t,P) [°C]

(e) Parameter variation: n

0 1 10 10 2 10 3 10 4 10 5

Time [y]

n = 27

n = 7

n = 3

n = 1

Param. Value Benefit / unit

t cool

55 y

100 y

200 y

k 4

3

2

D 1

D 2

40 m

50 m

60 m

7.6 m

9 m

11 m

n 27

7

3

(a)

0.25 °C/y

0.14 °C/y

0.07 °C/y

16 °C/FE

16 °C/FE

16 °C/FE

1.3 °C/m

0.5 °C/m

0.2 °C/m

6.8 °C/m

5.1 °C/m

3.4 °C/m

-0.1 °C/row

0.7 °C/row

5.5 °C/row

0 1 10 10 2 10 3 10 4 10 5

| | Tab. 2.

Expected benefit on peak temperature per unit change of a parameter, starting with the baseline

configuration. (a) Configuration C a ; (b) Configuration C b .

Example for case (a): Increasing t cool = 55 y by one year yields a benefit of 0.25°C. Extending D 2 by 1 m

results in a benefit of 6.8 °C (27 times more in comparison). FE = “fuel element.”

60

50

40

30

20

10

0

60

50

40

30

20

10

0

u(t,P) [°C]

u(t,P) [°C]

(f) Parameter variation: s

Time [y]

40 m

50 m

60 m

80 m

asymp.

27 x 81

15 x 45

9 x 27

3 x 9

1 x 1

0 1 10 10 2 10 3 10 4 10 5

Param. Value Benefit / unit

t cool

55 y

100 y

200 y

k 4

3

2

D 1

D 2

40 m

50 m

60 m

7.6 m

9 m

11 m

n 27

7

3

(b)

0.25 °C/y

0.14 °C/y

0.07 °C/y

16 °C/FE

16 °C/FE

16 °C/FE

1.3 °C/m

0.5 °C/m

0.2 °C/m

6.8 °C/m

5.1 °C/m

3.4 °C/m

-0.1 °C/row

0.7 °C/row

5.5 °C/row

at P 1 increases to 100 °C in a few decades,

peaks at 103 °C after 316 y and

remains in this range for several hundred

years (Figure 2, black curves).

Examination of Figure 2 leads to the

following observations.

The increase of cooling times is an

efficient option for reducing peak

temperature, but beyond a few

decades the benefits rapidly decrease

(Figure 2a, Table 3a). As t cool is raised

from 55 y to 100 y for example, peak

temperature decreases by 8 °C, representing

a benefit of 1.8 °C per decade.

A further rise to 300 y decreases peak

temperature by another 16 °C, representing

a benefit of 0.8 °C per decade.

For comparison, increasing t cool from

30 y to 55 y represents a benefit of

5.5 °C per decade. The rate of heating

decreases with increasing cooling

times, which is beneficial for retrieval

in the first few hundred years, but

peak time increases (the time for

which peak temperature is attained).

On the detriment side, intermediate

storage of large inventories on the

land surface represents a considerable

effort and is vulnerable to natural

or human-made hazard, including

malicious acts. Its deployment requires

monitoring, inspection, maintenance,

repair and, not least, a stable

society.

The reduction of batch charge k

leads to a reduction of peak temperature

by about 16° per fuel

element, while peak times roughly

remain constant (Figure 2b). The

considerable thermal benefits are

associated with high efforts as a

greater number of batches have to

be manufactured and handled over

farther distances, with possible

adverse effects regarding occupational

safety. Accordingly, the disposal

area and total drift length increases.

For example, the effort for decreasing

batch charge from 4 fuel elements to 1

implies quadrupling the number of

waste containers and quadrupling the

total length of the disposal drifts

altogether. For these reasons, the

benefit-to-effort ratio decreases as

k is reduced.

Increasing the spacing of disposal

drifts leads to a substantial reduction

of peak temperature as D 1 is increased

from 40 m to 50 m, but the positive

trend markedly decreases thereafter

(Figure 2c). Further reduction practically

stops at u = +50 °C. Accordingly,

the reduction of peak temperature

per meter drift spacing is important

for small D 1 and almost inexistent for

large D 1 (Table 3a). On the positive

side, the temperature peak becomes

narrower for high D 1 , with peak time

occurring earlier. Counting 316 y to

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atw Vol. 63 (2018) | Issue 5 ı May

(a) Parameter variation: t cool

(b) Parameter variation: k

(c) Parameter variation: D 1

50 u(t,P) [°C]

40

30

20

10

0

0 1 10 10 2 10 3 10 4 10 5

50 u(t,P) [°C]

40

30

20

10

0

55 y

100 y

200 y

300 y

450 y

(d) Parameter variation: D 2

0 1 10 10 2 10 3 10 4 10 5

50 u(t,P) [°C]

40

30

20

10

0

7.6 m

9 m

11 m

15 m

asymp.

(g) Parameter variation: n, D 1 , D 2

27, 40 m, 7.6 m

9, 65 m, 12 m

9, 65 m, 15 m

1, −, 15 m

asymp.

0 1 10 10 2 10 3 10 4 10 5

| | Fig. 3.

Time [y]

Temperature evolution at P = P 2 for the

baseline configuration C a (black) and

alternative configurations (blue).

A-F: One parameter changed (parameter in

title); G: Three parameters changed.

Host rock composition as in Figure 2.

peak for D 1 = 40 m, this is reduced to

just 32 y to peak for D 1 = 60 m. The

envelope area required for siting the

repository increases proportionally

to drift spacing. This option requires

to increase the length of the access

tunnel to the drifts, while the total

length of the disposal drifts remains

unchanged. An increase from 40 m to

60 m increases space requirements by

50 % in envelope area. The cost and

side effects have to be weighed against

the benefit of a significantly lower and

earlier temperature peak. Also, with

more space available between drifts,

retrieval could be facilitated.

Increasing batch spacing D 2 from

7.6 m to 15 m leads to a remarkable

reduction of peak temperature at P 1 ,

cutting the amplitude by over 40 %

(Figure 2d, Table 3a). Peak time

remains roughly constant for small

D 2 and decreases abruptly between

11 m and 15 m. This option requires

doubling the total length of the disposal

drifts, while the number of

batches remains unchanged. The area

for disposal increases proportionally

50 u(t,P) [°C]

40

30

20

10

0

0 1 10 10 2 10 3 10 4 10 5

50 u(t,P) [°C]

40

30

20

10

0

k = 4

k = 3

k = 2

k = 1

(e) Parameter variation: n

0 1 10 10 2 10 3 10 4 10 5

Time [y]

n = 27

n = 7

n = 3

n = 1

to batch spacing. This cost has to be

weighed against a significant reduction

of peak temperature. Retrieval is

handicapped on the one side by longer

drifts and favoured on the other by

substantially decreased temperatures.

Passing from 27 to 3 disposal drifts

leads to a limited reduction of peak

temperature, while passing from 3

to 1 drift leads to a more significant

drop (Figure 2e). Accordingly, the

reduction in peak temperature per

row is slight for large n and important

for small n (Table 3a). Similar to the

effect of drift spacing, the temperature

peak shortens significantly. For

n = 3, the temperature peaks 40 y

after waste disposal (compared with

316 y in base line configuration).

Decreasing the numbers of drifts

significantly influences the geometry

of the repository; the area required

remains constant, but its envelope

becomes long and narrow. For low

values of n, the difficulty of “fitting”

the repository into a host formation

may increase. In panel e, the number

of drifts is reduced in steps from 27 to

1. Accordingly, their length increases

from 616 m (27 drifts) to 2.4 km

(7 drifts) to 5.6 km (3 drifts) and to

16.7 km for a single drift. The technical

effort consists in ensuring longer

operating times for the individual

drifts.

Finally, the repository can be

divided into clusters (in some cases, it

might have to, due to the presence of

determining geological features). The

clusters are assumed to be sited in

sufficient distance to remain thermally

unaffected from each other. The

effect is shown in Figure 2f. Thermal

benefits on peak temperature are

limited unless the clusters are made

50 u(t,P) [°C]

40

30

20

10

0

0 1 10 10 2 10 3 10 4 10 5

50 u(t,P) [°C]

40

30

20

10

0

(f) Parameter variation: s

Time [y]

40 m

50 m

60 m

80 m

asymp.

27 x 81

15 x 45

9 x 27

3 x 9

1 x 1

0 1 10 10 2 10 3 10 4 10 5

very small (i.e. holding well below

10 % of the total number of batches).

The required area per cluster decreases

proportionally to the number

of batches it holds, which can be an

advantage for “fitting” the repository

in perturbed sites. In panel f, the

disposal area is fractioned into clusters

of various size (indicated by n×m,

with s = nm| option / nm| baseline ). As

for panel c, this option requires to

increase the length of the access

tunnels.

Panel g presents results for

alter native configurations with n

decreased and D 1 , D 2 increased, other

parameters unchanged. The asymptotic

case is represented by the configuration

D 1 , D 2 → ∞. The hatched area

represents the scope for thermal

dimensioning below the temperature

curve belonging to the baseline configuration.

A large fraction of the

scope for thermal dimensioning is

available if triple the amount of space

with respect to the baseline configuration

is reserved, allowing for efficient

action on D 1 , D 2 and n.

Another point of interest is the

thermal benefit at the time horizon

10 4 y, when steel waste canisters are

expected to fail. At this point, the

ambient conditions in the canister core

change from dry to wet, thus enabling

liquid-phase reactions that were previously

excluded. For thermally activated

diffusion-reaction processes, the

rates depend on the ambient temperature.

While all engineering parameters

can be used to substantially reduce

the temperatures in first 10 3 y, this is

no longer the case in the following 10 4

y. Except for the asymptotic cases,

there is little response in temperatures

at the horizon 10 4 y for changes in t cool ,

DECOMMISSIONING AND WASTE MANAGEMENT 323

Decommissioning and Waste Management

Scope for Thermal Dimensioning of Disposal Facilities for High-level Radioactive Waste and Spent Fuel ı Joachim Heierli, Helmut Hirsch, Bruno Baltes


atw Vol. 63 (2018) | Issue 5 ı May

(a) Parameter variation: t cool

(b) Parameter variation: k

(c) Parameter variation: D 1

DECOMMISSIONING AND WASTE MANAGEMENT 324

60 u(t,P) [°C]

50

40

30

20

10

0

0 1 10 10 2 10 3 10 4 10 5

60 u(t,P) [°C]

50

40

30

20

10

0

33 y

66 y

100 y

200 y

300 y

(d) Parameter variation: D 2

0 1 10 10 2 10 3 10 4 10 5

60 u(t,P) [°C]

50

40

30

20

10

0

8.9 m

10 m

12 m

16 m

asymp.

(g) Parameter variation: n, D 1 , D 2

30, 25 m, 8.9 m

10, 40 m, 12 m

10, 40 m, 15 m

1, −, 15 m

asymp.

0 1 10 10 2 10 3 10 4 10 5

| | Fig. 4.

Time [y]

Temperature evolution at P = P 3 for the baseline

configuration C b (black) and alter native

configurations (blue). Panels a-f: One parameter

changed (parameter in title); g: Three

parameters changed. Host rock composition:

crystalline rock with isotropic heat conductance

2.82 W/mK, heat capacity 2.09 MJ/m 3 K.

D 1 and D 2 (Figure 2). Instead, changes

in n and s lead to more substantial

thermal benefits. Changes in k result

in an intermediate situation. In this

sense, it can be said that action on t cool ,

D 1 and D 2 mainly lead to short-term

thermal benefits, while action on n

and s lead to long-term benefits. A

comprehensive thermal dimensioning

should therefore include action on

parameters from both groups.

Figure 3 shows the temperature increase

at a point P 2 sited 20 m above

the repository centre in the baseline

configuration C a for argillaceous rock,

as well as for parameter variations

thereof. P 2 is of relevance for the

admissibility of pore water pressure

resulting from the increase in temperature.

At this point, the temperature

raises by 49 °C within 500 y and

decreases thereafter. Regarding efforts

and benefits, the discussion is qualitatively

identical to the dis cussion of

Figure 2. Differences are mainly

quantitative in nature. For example,

the asymptotic limits (dashed, blue)

60 u(t,P) [°C]

50

40

30

20

10

0

0 1 10 10 2 10 3 10 4 10 5

60 u(t,P) [°C]

50

40

30

20

10

0

k = 9

k = 6

k = 3

k = 1

(e) Parameter variation: n

0 1 10 10 2 10 3 10 4 10 5

Time [y]

n = 30

n = 6

n = 3

n = 1

for D 1 → ∞, D 2 → ∞,

n = 1 and s → 0 are considerably

smaller at P 2 than at P 1 . For D 2 and s

they almost vanish. One important

aspect for the present discussion is the

observation that action on D 1 and n

leads to comparatively greater benefits

for the same effort at P 2 than at P 1 . For

example, at P 2 , peak temperature

decreases by a factor of 1.8 (from 48 °C

down to 26 °C) when passing from

n = 27 to n = 3. The same action at P 1

results in a benefit of a factor of 1.1

only (from 65 °C down to 58 °C). In

Figure 3a-f, only a single engineering

parameter has been changed at a time.

Results for alternative configurations

with n decreased and D 1 , D 2 increased,

other parameters unchanged, are

shown in Panel g. The graphs show

that considerably lower temperatures,

even approaching the asymptotic case,

can be envisaged if sufficient space is

reserved to allow for reasonable action

on D 1 , D 2 and n.

Qualitatively similar findings apply

to point P 3 and to the case of a repository

sited in crystalline rock with

configuration C b (Table 3b, Figure 4).

Altogether, the results indicate that

the patterns of interdependence

between spatial and thermal dimensioning

are non-trivial and cannot be

parameterised by the space-averaged

heat load alone. For example the

average initial heat loads for D 1 =

80 m in Figure 2c and for D 2 = 15 m in

Figure 2d are both close to 2.2 W/m 2 ,

yet their peak temperatures differ by

more than 12 °C.

Conclusions

There is a close interdependence

between spatial and thermal dimensioning

of disposal facilities for

60 u(t,P) [°C]

50

40

30

20

10

0

0 1 10 10 2 10 3 10 4 10 5

60 u(t,P) [°C]

50

40

30

20

10

0

25 m

30 m

50 m

60 m

asymp.

(f) Parameter variation: s

0 1 10 10 2 10 3 10 4 10 5

Time [y]

30 x 30

21 x 21

12 x 12

3 x 3

1 x 1

high­ level radioactive waste and spent

fuel and an important potential for

further reduction of the transient

temperature peak, resulting in supplementary

thermal benefits. After a few

decades of cooling storage, the most

efficient adjusting screws for thermal

dimensioning beyond strict admissibility

are drift spacing D 1 , batch spacing

D 2 and number of drifts n. Action

on batch charge k results in large

benefits but also large efforts to obtain

them. Cooling storage becomes

rapidly less efficient as cooling time

t cool increases. Dividing the repository

into clusters is inefficient with regard

to the reduction of peak temperatures

unless clusters are made very small.

Action on t cool , D 1 and D 2 leads to

short-term thermal benefits, while

action on n and s to long-term thermal

benefits. Except for t cool , all options

are related to geometrical side-effects

that increase space requirement. If, in

the current process of site selection,

spatial reserves are limited tightly in

proportion to space-saving configurations,

the options for final thermal

dimensioning by action on D 1 , D 2 , k, n,

and s in later project phases are

severely reduced. If on the other hand,

the scope for action on these parameters

is preserved, considerably cooler

repository designs remain possible in

later project phases. Therefore, the

preservation of technical options for

final thermal dimensioning should

always be part of early decisionmaking

in the site selection process.

This particularly applies for temperature-sensitive

configurations in argillaceous

or crystalline rock. It should

equally apply to rock salt when

retrievability of waste is legally

required.

Decommissioning and Waste Management

Scope for Thermal Dimensioning of Disposal Facilities for High-level Radioactive Waste and Spent Fuel ı Joachim Heierli, Helmut Hirsch, Bruno Baltes


atw Vol. 63 (2018) | Issue 5 ı May

Acknowledgements

This work was funded by the State

Council of the Canton of Schaffhausen,

Switzerland. The authors declare no

conflict of interest. The source code of

the computational tool used for

thermal calculations in this study has

been made available by J.H. to the

German nuclear science web-platform

nucleonica, who provide a considerably

speeded-up web-application

thereof.

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(27 pp.). Paris: OECD/NEA Publishing.

18. Wieczorek, K., Gauss, I., Mayor, J.C.,

Schuster, K., García-Siñeriz, J.L., Sakaki,

T. (2017). In-situ experiments on

bentonite-based buffer and sealing

materials at the Mont Terri rock

laboratory (Switzerland). Swiss Journal

of Geosciences, 110, 253-268.

19. Whipple, C.G., Budnitz, R.J., Ewing, R.C.,

Moeller, D.W., Payer, J.H., Witherspoon,

P.A. (1999). Final report total system

performance assessment peer review

panel (144 pp). Washington:

Department of Energy.

Authors

Dr.-Ing. Joachim Heierli

Interkantonales Labor

Mühlentalstrasse 188

8200 Schaffhausen, Switzerland

Dr. Helmut Hirsch

Consultant to the Environment

Agency Austria

Spittelauer Lände 5

1090 Vienna, Austria

Dr. Bruno Baltes

Senior adviser on long-term safety

of radioactive waste disposal

Germany

325

RESEARCH AND INNOVATION

Heavy Ions Irradiation as a Tool to

Minimize the Number of In-Pile Tests

in UMo Fuel Development

H. Breitkreutz, J. Shi, R. Jungwirth, T. Zweifel, H.-Y. Chiang and W. Petry

Introduction Irradiation with heavy ions from an accelerator source is an increasingly often used tool to quickly

reproduce and simulate certain effects of in-pile irradiation tests, avoiding the complexity and high costs of handling

highly radioactive samples. At the Maier-Leibnitz Laboratorium (MLL) of the Technische Universität München (TUM),

swift heavy ions have been applied in the development of Uranium-Molybdenum (UMo) based research reactor fuels

for more than 10 years. Since then, the technique has been advanced from feasibility over qualitative analysis to

quantitative prediction, including fission gas implantation.

Accelerator based

out-of-pile irradiation

In accelerator based out-of-pile

experiments, Iodine-127 with energy

of 80 MeV serves as representative

fission product. The usage of Iodine,

which has only one stable isotope,

permits for efficient extraction, acceleration

and – if necessary – detection

in the irradiated sample.

Samples are either prepared from

as-manufactured fuel plates or produced

as tailored model systems, depending

on the application. Cut-outs

of fuel plates usually are employed for

Research and Innovation

Heavy Ions Irradiation as a Tool to Minimize the Number of In-Pile Tests in UMo Fuel Development ı H. Breitkreutz, J. Shi, R. Jungwirth, T. Zweifel, H.-Y. Chiang and W. Petry


atw Vol. 63 (2018) | Issue 5 ı May

RESEARCH AND INNOVATION 326

| | Fig. 1.

Irradiation set-up for heavy ions irradiation of small fuel samples with I-127

at the MLL accelerator facility. The sample holder offers 3 irradiation positions

with temperature sensors, electrical heating, air cooling and a faraday blend.

An IR camera is used for monitoring the samples and for adjusting the beam.

studies of UMo dispersion fuel, where

small fuel particles of a few dozen

micrometre diameters are dispersed

in an aluminium matrix. Model

systems normally consist of an Al

substrate, an optional coating layer of

a few 100nm and a ≤5 µm thin UMo

layer on top. In certain cases, the

order is reversed and UMo serves as

substrate with a ≤13 µm thick Al layer

on top. The model systems have

been produced by Physical Vapor

Deposition (PVD) in the Uranium lab

of FRM II, dispersion plates were

manufactured by AREVA-CERCA.

The irradiations have been performed

at the Maier-Leibnitz Laboratorium

in a dedicated irradiation setup

(Figure 1). The MLL houses a

tandem accelerator that provides

I-127 ions at the desired 80 MeV with

high intensity. Ions are targeted perpendicular

on the sample surface.

Penetration depth ranges between

5 µm and 13 µm, depending on the

material of the top layer. The sample

can be cooled and heated in a temperature

range between 30 °C and

~200 °C. Optionally, a wobbler can be

hooked up to the beam to irradiate a

larger area with a beam that is still

focused. In that case, a quantitative

evaluation is only possible for specific

applications. Otherwise, quantitative

analysis is possible whenever the

geometric beam intensity profile can

be determined.

Uranium-Molybdenum and

the interdiffusion layer

To minimize the usage of highly

enriched Uranium (HEU) in the civil

nuclear cycle, the High Performance

Research Reactors (HPRR) that are

currently using HEU are seeking to

convert their cores to low enriched

Uranium (LEU). As the core geometry

in these reactors usually can only be

varied to a small extent, the current

fuels need to be replaced by an alloy

with considerably higher chemical

Uranium density to account for the

lower fraction of fissionable isotopes

in LEU. In this context, the most

promising candidates for such a

conversion are fuels based on the

Uranium-Molybdenum alloy with

7-10wt % Mo, either as dispersion or

as monolithic fuel. These two variants

commensurate to the two sample

variants discussed beforehand.

One particularly important application

for the ions in this context has

been the study of the growth of the

interdiffusion layer (IDL) that forms

between UMo and the aluminium

matrix and/or the Al cladding during

irradiation [1]. Due to its inferior

irradiation properties, the IDL can

lead to exponential fuel swelling and

therefore needs to be avoided [2, 3,

6]. Therefore, numerous experiments

have been carried out to understand

the development of this IDL and to

test countermeasures like the addition

of silicon to the Al matrix, the application

of diffusion barriers between

UMo and Al and even the substitution

of the Al matrix by Magnesium. Ionbased

experiments have con siderably

accelerated this develop ment; some

exemplary results will be discussed in

the following.

Understanding the IDL

Growth dynamics

Kim and Hofman [11] have developed

an Arrhenius-like formula to predict

the thickness d IDL of interaction layer

formation between UMo and Al

during irradiation, based on the data

of several in-pile experiments:

(1)


Here, A = 2.6 ∙ 10 -8 µm 2 cm 3p s p-1 is the

proportionality factor, p = 0.5 the

power of the fission rate f that has

been averaged over the irradiation

time t and q = 3,850 K the fit parameter

for the average irradiation temperature

T. f Mo and f Si are correction

factors for the molybdenum content

of the fuel and the silicon content the

matrix.

Based on a SRIM/TRIM [11]

dataset for the deposition of energy

and the creation of vacancies by

in-pile fission products as well as the

I-127 ions, a conversion of ion flux

and fluency to the corresponding

fission rate and -density equivalents is

possible. Jungwirth [8], has discussed

that the irradiation enhanced diffusion

which leads to the creation of this

layer is mainly driven by thermal spiking,

from ionization as well as recoils.

Therefore, in this case, φξ, where

ξ is the relative total energy deposition

that was calculated using SRIM/

TRIM. The factor 0.5 originates from

a real fission producing 2 fission

products. The calculation of ξ is laid

out in detail in [17].

An inverse Al/UMo system with

the Bragg peak near the interface was

irradiated [18, 22, 35] to study the

growth dynamics of the IDL and to

verify the flux-fission rate conversion

| | Fig. 2.

Deviation of the measured IDL thickness compared to the value expected from in-pile irradiations,

i.e. y-axis is difference between expected and measured value, divided by measured value.

The use of a cubic fit is arbitrary to visualize a trend.

Research and Innovation

Heavy Ions Irradiation as a Tool to Minimize the Number of In-Pile Tests in UMo Fuel Development ı H. Breitkreutz, J. Shi, R. Jungwirth, T. Zweifel, H.-Y. Chiang and W. Petry


atw Vol. 63 (2018) | Issue 5 ı May

mechanism in eq. 1. The beam had a

2D-Gaussian profile and was not

wobbled, the six samples therefore

cover a large variety of fission rate and

burn-up equivalents. The samples

were irradiated at six different

temperatures between 383 K and

548 K and for different time spans

between 5 h and 27.85 h. For this

irradiation geometry, ξ = 0.21 µm -1

was calculated.

Figure 2 shows the deviation

between expected and measured IDL

thickness. Even though significant

scattering is present, which is attributed

to the general fluctuation of IDL

thicknesses that is also observed

in-pile and to suboptimal sample

quality, prediction and measurement

match well within the uncertainties

of the experiment and the assumed

uncertainties of the equation of Kim

and Hofman (~15 %). The proportionality

constant A was reproduced

with an accuracy of 10 %. It was

furthermore shown that p is constant

over several orders of fission rates and

the dependency d IDL ∝√t was established

also for the early growth of the

IDL [18]. The temperature normalization

q, which includes the activation

energy, was verified by comparing the

value obtained from the fit results of

ion irradiation data (3,906K ±30 K)

and the one from in-pile data

(3,850K).

In summary, the matching thicknesses

and growth dynamics of outof-pile

produced IDLs with predictions

based on in-pile data support

the current understanding of the

Arrhenius­ like in-pile IDL growth as

well as the conversion between ion

flux and fission rate.

fission density, therefore exact

comparisons are difficult. The same

observations were made for ground as

well as atomized UMo powder [6, 8].

Ion and in-pile experiments did

not show ternary U x Mo y Al z phases

like purely thermal-driven diffusion

couple experiments [24, 25], which

underlines the necessity of irradiation

to correctly reproduce IDL growth.

Nearest neighbour distances in

neutron irradiations were found

between 0.239 nm and 0.251 nm [2,

23], which is in agreement with

0.248 nm measured after Iodine

irradiation [20].

Fission gas behaviour

The experiments presented above

cover the early phase of the fuel in the

reactor, up to a fission density of

n = 5.8 ∙ 10 20 , about 15 % of the

target burn-up for that fuel. Even

though fission gases play a less pronounced

role at this point, their

behaviour already can provide insight

into the mechanism that leads to

the break-away swelling that was

observed around n ≈ 2.5 ∙ 10 21

in early in-pile UMo irradiation

experiments.

To extend the ion irradiation technique

to this field, Krypton-82 ions

with 45 MeV have been implanted

on a UMo/Al sample at the GANIL

facility. The sample was irradiated

with I-127 beforehand to a burn-up

equivalent of n = 1.2 ∙ 10 20 [9, 10]

at the MLL. Kr was preferred over

the more common fission gas Xenon

as it is easier to accelerate. At

c Kr = 2.68 ∙ 10 19 , the concentration

of Kr ions reached the desired

equivalent of 25 % of “virtual” fission

products. Even though the two irradiations

have been carried out consecutively

and the irradiation environment

in the second did not reach the fission

rate equivalent of an in-pile irradiation

due to limited beam intensity,

a similar gas bubble growth was

observed as in-pile (Figure 3): Fission

gases have accumulated at the IDL

interfaces in macroscopic bubbles.

Secondary Ion Mass Spectrometry

(SIMS) revealed that the bubbles are

indeed filled with Kr and that the Kr

has accumulated in the IDL [9].

Avoiding the IDL

As outlined above, several approaches

are available to limit or prevent IDL

growth in UMo/Al fuel. Silicone

added to the matrix can form a protective

Si layer around the particles,

since U x Si y preferably forms over UAl x .

From a manufacturing perspective,

such an addition would be superior

over the coating of particles with a

diffusion barrier material. A more

rigorous approach would be to replace

the Al matrix of dispersion fuel by

Magnesium, which does neither interact

with Uranium nor Molybdenum.

Additions to the matrix

If Si is added to the Al matrix, a Si rich

diffusion layer (SiRDL) will form

around the UMo matrix during production

of the plate if the Si fraction

exceeds 2-5 %. This layer consists of a

mixture of U x Si y -phases. During irradiation,

the SiRDL is consumed and

finally a conventional IDL grows, i.e.

its presence does not prevent the

formation of an IDL (Figure 4). The

RESEARCH AND INNOVATION 327

Composition

IDLs generated in-pile usually consist

of UAl 3 with far smaller amounts of

UAl 2 and UAl 4 . Early ion experiments

lead to a partially crystalline IDL.

UAl 3 was identified using XRD [8]

but SEM/EDX showed a higher Al

fraction (3.5 – 5), which was attributed

to partial amorphization.

Improved sample cooling and the

usage of monolithic samples allowed

reproducing a fully amorphous IDL

[19]. RBS measurements showed a

U/Al ratio of ½, while EDX yielded ²∕ 3

[20]. The discrepancy is attributed to

the measurement techniques. Latest

experiments without beam wobbling

led to IDLs with a U/Al ratio between

¹∕ 3 and ¼ as identified by EDX [22].

This fairly agrees with in-pile results;

IDL composition depends on irradiation

temperature, fission rate and

| | Fig. 3.

Kr implantation in an Iodine pre-irradiated UMo/Al layer system: The Kr beam targets pre-irradiated as

well as as-fabricated parts of the sample (A). Fission gas bubbles can only be found inside the IDL (B),

not in the as-fabricated part (C). No bubbles have been found in locations that were not previously

irradiated with Iodine but only exposed to Kr (D). The white areas in (A) were analyzed using SIMS to

confirm the presence and depth profile of Kr.

Research and Innovation

Heavy Ions Irradiation as a Tool to Minimize the Number of In-Pile Tests in UMo Fuel Development ı H. Breitkreutz, J. Shi, R. Jungwirth, T. Zweifel, H.-Y. Chiang and W. Petry


atw Vol. 63 (2018) | Issue 5 ı May

RESEARCH AND INNOVATION 328

| | Fig. 4.

IDL growth around UMo particles in a UMo/AlSi7 miniplate, protected by a SiRDL, in three steps, from low to high ion dose.

Left: The UMo particle located at the sample surfaces is surrounded by first a SiRDL and then an IDL. Center: In the upper part of UMo

particle embedded in the Al matrix, the SiRDL has been destroyed by the heavy ion irradiation and the IDL is directly in contact with

the UMo particle core. Right: A standard UMo/Al interaction occurs and the shape of the UMo particle indicates clearly that UMo has

been consumed in the interaction.

same effect was observed in the IRIS-3

and RERTR-7 test [28, 29]. It was also

found that the IDL is free of Si except

when precipitates were present from

manufacturing using blended instead

of alloyed AlSi-powder.

Ion irradiations were carried out

on samples from real test plates of the

E-FUTURE II (EF2) irradiation experiment

with 7 and 12 wt% Si addition to

the Al matrix. E-FUTURE [7] samples

were not irradiated due to their

likeness to other samples irradiated

before. For the EF2 samples, it was

found that the SiRDL similarly gets

consumed during irradiation [31].

Yet, extrapolation predicted sufficient

Si around the particles and in the

matrix until very high burn-up. In the

end, the EF2 test failed unexpectedly

due to macroscopic deformation of

the plates at comparably low burn-up

[4]. One of the possible reasons is the

very low creep of the hard AlSi12

matrix. Thereby, EF2 is an experiment

demonstrating the limits of out-of-pile

ion irradiation.

Other additions to the matrix like

2-5 wt % Bi and 2 wt % Ti have been

tested using ions but not in-pile as no

beneficial effect was found [8].

Several other coating materials

would be available from a reactor

physics and manufacturing point of

view, e.g. Mo, Nb and Ti. None of

these has been tested in-pile so far,

but in a series of ion experiments

[20] and analysed using Rutherford

Backscattering and µ-XRD. The

experiments have correctly predicted

the efficacy of several coating

materials as well as their interaction

with surrounding matrix and

fuel materials. Several interactions

between the coating, UMo and Al

were found: For Ti, the presence of

Ti 0.04 U 0.96 indicates a transformation

of γ-UMo to α-UMo. Nb forms

compounds on both interfaces that

have poor crystallinity, Nb 3 Al and

Mo 0.1 Nb 0.45 U 0.45 .

For the current standard coating of

monolithic fuel, Zr, it was found that it

forms Mo 2 Zr with the Mo from the

UMo fuel, leaving behind a small area

that is poor in Mo in the fuel zone. In

this area, the UMo would no longer be

sufficiently stabilized, i.e. α-U would

form and macroscopic bubbles would

appear – a precursor of plate failure.

Such an effect has indeed turned up

in the RERTR-12 in-pile irradiation

at very high fission densities

(n > ∼ 6 ∙ 10

21

[33].

No detrimental effects could

be identified for a Molybdenum

coating, which is now tested in-pile

in the SEMPER FIDELIS experiment

since September 2017. Most important,

it could be shown that the

driver for the interaction between

the barrier material, the UMo fuel

and the Aluminium matrix is

indeed the chemical potential of

the respective constituents. This

allows for a prediction of the protective

properties of coating material

candidates [20].

Magnesium as matrix material

If the Aluminium matrix was replaced

by Magnesium, no chemical reaction

between UMo and the matrix would

take place. Instead, TEM analysis

showed that spinodal decomposition

would occur at elevated temperatures

(~200 °C, Figure 5), which favours

embrittlement. However, this temperature

is at the upper limit of what is

usually found in research reactor fuel

under nominal conditions (~140 °C),

where no spinodal decomposition

could be detected. At the lower

temperature, only a ~50 nm thin

interaction layer at the interface with

decreased crystallinity was found. Mg

would therefore be a viable alternative

for the matrix material once all

manufacturing-specific difficulties are

solved.

Besides its use in the 50’s and 60’s 1 ,

an UMo/Mg fuel has already been

tested in the RERTR-3 and -8 experiments

under increasingly aggressive

irradiation conditions up to temperatures

between 145 °C and 171 °C

[27]. In agreement with the ion

1) See [21] for a

comprehensive

overview

Coating of UMo particles

A coating aims on physically – and

thereby chemically – separating UMo

and Aluminium. For monolithic UMo,

a layer of Zr usually is applied between

foil and cladding. UMo particles in

dispersion fuel are coated with ZrN

and Si. In-pile, a Si coating has shown

little to no benefit over the Al-Simatrix

[5, 31]. The protective properties

of ZrN were confirmed on actual

dispersion samples from SELENIUM.

Cracks in the coating led to formation

of a conventional IDL in ion and

neutron experiments [4, 30].

| | Fig. 4.

Spinodal decomposition in the UMo/Mg system at 200°C. No decomposition and only very little

amorphous interaction was found at 140°C, the typical operating temperature for HPRR cores.

Research and Innovation

Heavy Ions Irradiation as a Tool to Minimize the Number of In-Pile Tests in UMo Fuel Development ı H. Breitkreutz, J. Shi, R. Jungwirth, T. Zweifel, H.-Y. Chiang and W. Petry


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atw Vol. 63 (2018) | Issue 5 ı May

RESEARCH AND INNOVATION 330

experiments, only negligible interaction

occurred near the interface

[21] and no indications for spinodal

decomposition were found [26].

UMo phase behaviour

Few experiments have been carried

out so far using heavy ions that

directly target effects in the UMo

kernels. One major finding was the

reduction of the orthorhombic

α-phase of Uranium during irradiation,

i.e. the stabilization of the bcc

γ-phase [8]. This is an imperative

requirement for the fuel. The same

effect was observed in-pile [34],

though quantitative comparisons are

still outstanding.

Ternary alloys like U 8 wt % Mo

1 wt % Pt which originally were

considered to improve irradiation

behaviour showed no improvement

and partially even a destabilisation

of the γ-phase of the fuel and were

therefore discarded, too [32].

Conclusion

Altogether, the experiments very well

demonstrate the applicability of the

approach to use Iodine-127 irradiation

for qualitative and even quantitative

experiments to reliably simulate

numerous in-pile irradiation effects.

Even though the burn-up equivalent

that has been reached up to now

with ions was comparably low

(n < ∼ 6 ∙ 10

20

) for HPRR conditions,

important contributions were made to

the development of UMo fuels.

In the future, improved ion beam

diagnostics will open the door to the

analysis of high-burn-up effects like

the recrystallization of UMo kernels.

The necessary fission density equivalents

can be reached with only a few

days of beam time, depending on the

influence of fission rate/ion flux.

Subsequent fission gas implantation

additionally allows studying gasdriven

effects – first demonstration

experiments were successfully carried

out with Kr ions.

The main advantage of this

technique is its effectiveness: No

additional radioactivity is involved,

the complete cycle from experiment

design over irradiation to post irradiation

examinations can be carried

out in a few weeks. Even though the

method has obvious limitations, it is

well able to minimize the number of

costly and time consuming in-pile

irradiation experiments.

Acknowledgements

The authors would like to thank the

MLL staff for their great support

during our beam-times and also in

the times in-between as well as all

members of the working group

“Hochdichte Kernbrennstoffe”.

The work was supported by a

combined grant (FRM1318) from the

Bundesministerium für Bildung und

Forschung (BMBF) and the Bayerisches

Staatsministerium für Bildung und

Kultus, Wissenschaft und Kunst

(StMBW).

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[2] A. Leenaers et. Al.: Post-irradiation

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Research and Test Reactors (RERTR),

2016.

[19] H.-Y. Chiang et. Al.: Evidence of

amorphous interdiffusion layer in heavy

ion irradiated U-8 wt %Mo/Al interfaces,

Journal of Nuclear Materials 440

(2013) 117–123.

[20] H.-Y. Chiang, Material selection of UMo

fuel for research reactors: Swift heavy

ion irradiation studies, Ph.D. thesis,

Technische Universität München

(2014).

[21] D.D. Keiser et. Al.: Microstructural

Characterization of a Mg Matrix U-Mo

Dispersion Fuel Plate Irradiated in the

Advanced Test Reactor to High Fission

Density: SEM Results; Metallurgical and

Materials Transactions 3E (2016) 71-89

[22] J. Hingerl, Evolution of the Inter

Diffusion Layer of UMo/Al Nuclear Fuel

during Heavy Ion Irradiation, Bachelor

thesis, Technische Universität

München, 2016

[23] S. van den Berghe et. Al., Transmission

electron microscopy investigations of

irradiated U-7 wt % Mo dispersion fuel,

Journal of Nuclear Materials 375

(2008) p. 340

[24] M.I. Mirandou et. Al.: Characterization

of the reaction layer in U-7 wt %Mo/Al

diffusion couples, Journal of Nuclear

Materials 323 (2003) 29-35

[25] H. Palancher et. Al.: Evidence for the

presence of UMoAl ternary compounds

in the UMo/Al interaction layer grown

by thermal annealing: A couple micro

X-ray diffraction and micro X-ray

absorption spectroscopy study, Journal

of Applied Crystallography, 40 (2007)

1064-1075

[26] D.D. Keiser, Private communication.

Publication in preparation.

Research and Innovation

Heavy Ions Irradiation as a Tool to Minimize the Number of In-Pile Tests in UMo Fuel Development ı H. Breitkreutz, J. Shi, R. Jungwirth, T. Zweifel, H.-Y. Chiang and W. Petry


atw Vol. 63 (2018) | Issue 5 ı May

[27] G.L. Hofman et. Al.: Proc. of the

RRFM-2008 International Meeting,

Hamburg, Germany, March 2–5, 2008.

[28] A. Leenaers et al. Microstructural

analysis of irradiated atomized UMo

dispersion fuel in an Al matrix with Si

addition. Proc. International Meeting

on Research Reactor Fuel Management

(RRFM), Hamburg, 2008

[29] D.D. Keiser et al. Microstructural

development of irradiated U-7Mo/6061

Al alloy matrix dispersion fuel, Journal

of Nuclear Materials, 393 (2009)

311-320

[30] R. Jungwirth et. Al.: Heavy ion

irradiation of UMo/Al samples with

protective Si and ZrN layers (SELENIUM),

Proc. International Meeting on Reduced

Enrichment for Research and Test

Reactors (RERTR), Santiago de Chile,

Chile, 2011

[31] R. Jungwirth et. Al.: Ion irradiation of

UMo/Al fuel samples with 7wt% and

12wt% Si inside the matrix (E-FUTURE

II), Proc. International Meeting on

Research Reactor Fuel Management

(RRFM), Prague, Czech Republic, 2012

[32] R. Jungwirth et. Al.: Study of heavy ion

irradiated UMo/Al Miniplates: Si and Bi

additions into Al and UMo ground

powders, Proc. International Meeting

on Research Reactor Fuel Management

(RRFM), Marrakech, Morocco, 2010

[33] M.K. Meyer et. Al.: Irradiation

performance of U-Mo monolithic fuel,

Nuclear Engineering and Technology,

46 (2014) p. 169-182

[34] M.L. Bleiberg et. Al.: Phase changes in

pile-irradiated uranium-base alloys,

Journal of Applied Physics, 27 (1956),

p. 1270-1283

[35] J. Shi et. Al.: Quantitatively simulation

fission-enhanced diffusion in U-Mo/Al

bilayer systems by swift heavy ion

irradiation, Proceedings of European

Research Reactor Conference RRFM

2018, Munich, Germany, March 11-15,

2018

Authors

H. Breitkreutz

J. Shi

R. Jungwirth 1)

T. Zweifel

H.-Y. Chiang

W. Petry

Forschungs-Neutronenquelle Heinz

Maier-Leibnitz (FRM II)

Technische Universität München

Lichtenbergstr. 1

85747 Garching, Germany

1) Author ist now with

European Commission.

The views and

opinions expressed

in this article do

not necessarily

reflect those of the

European Commission.

331

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KTG Inside


atw Vol. 63 (2018) | Issue 5 ı May

Top

Presidents of Russia and

Turkey Vladimir Putin and

Recep Tayyip Erdogan kicked

off large-scale construction of

Akkuyu NPP

(rosatom) On the 3rd of April 2018 in

Gülnar Municipality (Mersin Province,

Turkey) the function of first concrete

was held at the construction site of

Akkuyu NPP. It marked the start of

large-scale work to build the first

Turkish nuclear power plant being

constructed by ROSATOM.

President of the Russian Federation

Vladimir Putin and President of

the Republic of Turkey Recep Tayyip

Erdoğan took part in the function by

video conference.

Addressing the gathering, Putin

said: “The first unit of Akkuyu NPP

must be put on line in 2023. Owing to

Akkuyu project, new up-to-date wellpaid

jobs will be created and new

cutting­ edge productions and technologies

will develop in Russia and Turkey.

Frontline engineering solutions and

economically effective and sound technologies

will be used in construction.

Highest safety standards and most

stringent environmental requirement

will be met. I am sure that in 2023 entire

Turkey will feel the feedback of the

energy to be generated by this plant,

this high-technology facility.”

On his part, President of Turkey

Recep Tayyip Erdoğan said that in

2023 by the start of NPP Turkey celebrates

the declaration of the republic.

“Commissioning of Akkuyu NPP in

2023 will make Turkey a member of

the family of nuclear power countries.

We will meet 10% of our demand in

electricity owing to Akkuyu,” he said.

After welcome addresses, leaders

of the two countries gave the permit to

start construction and this was the

start of the first concrete pouring to

the basemat of Unit 1 reactor building.

The function was witnessed by

500 people who included local folks,

workers, schoolchildren, local authorities,

representatives of related

ministries and agencies, journalists,

business partners, and representatives

of ROSATOM’s companies participating

in the projects. Turkish young

graduates of related Russian universities

who had received high education

certificates less than a month ago and

would work soon in the project company

also took part in the function.

Commenting on the event, Director

General of ROSATOM Alexei Likhachev

stressed: “ROSATOM will build in

Turkey an up-to-date and reliable

nuclear power plant of new Generation

III+ with four powerful units

VVER-1200 which meets all international

safety requirements. In Russia

we are building these new generation

units in series: Novovoronezh Unit 6

was commissioned in the commercial

operation and in February we put on

line Leningrad Phase II’s Unit 1. Successful

operation of these units confirms

reliability of our technologies.”

Likhachev noted that Akkuyu NPP

was a large investment project of the

Russian-Turkish cooperation. “And it

is important to underline that the project

is implemented according to the

plan and in effective cooperation with

the Turkish partners,” he added.

The day before the project company

JSC Akkuyu Nuclear had obtained an

Akkuyu construction license from the

Turkish Atomic Energy Authority

(TAEK). Late March, the permit for

construction of Unit 1’s reactor building

was obtained from Gülnar local

administration.

Earlier, JSC Akkuyu Nuclear

received the full-fledged status of a

strategic investor. The renewed certificate

of a strategic investment was

issued to the Project Company by a

request filed to the Ministry of Economy

of Turkey by the company. The certificate,

in particular, provides for reduction

or relief from taxes (including the

profit tax and VAT) and relief from

paying customs charges and duties.

For information:

The Akkuyu NPP construction project is

the largest project in the history of the

Russian-Turkish relations. The plant

will consist of four VVER-1200 units

(Generation III+ reactors) with 4.8 GW

total capacity. This is a modernized

series­ made design of a nuclear power

plant based on Novovoronezh Phase II

(Russia, Voronezh Region). The project

model is BOO (build-own-operate).

The General Customer and Investor

of the Project is JSC Akkuyu

Nuclear ( AKKUYU NÜKLEER ANONİM

ŞİRKETİ, the company specially

founded to manage the project; the

Architect General is JSC Atomenergoproekt;

the General Construction

Constructor is JSC Atomstroyexport

(both companies are part of

ROSATOM’s Engineering Division); the

Project Developer is JSC Rusatom

Energy International. The Technical

Customer is Rosenergoatom Concern

JSC; the Scientific Supervisor is FSA

NRC Kurchatov Institute; the licensing

consultant is InterRAO-WorleyParsons.

The designed period of Akkuyu NPP is

60 years with a possibility to prolong it

20 years more.

| | First concrete pouring ceremony at the Akkuyu NPP (03.04.2018, Turkey)

Akkuyu NPP meets all up-to-date

requirements of the world nuclear

community written down in the safety

standards of the International Atomic

Energy Agency (IAEA) and International

Nuclear Safety Advisory Group

and EUR Requirements. The power

unit includes a reactor plant and

turbine; it will be fitted with the

cutting-edge active and passive safety

systems to meet all potential scenarios

and their combinations. All works to

implement Akkuyu NPP project are

carried out in close cooperation with

the Turkish Ministry of Energy and

Natural Resources and the independent

regulator TAEK.

| | (181061159), www.rosatom.ru

Company News

Framatome upgraded

Borssele nuclear power plant’s

digital instrumentation &

control system

(framatome) Framatome has performed

a comprehensive modernization

of the instrumentation & control

(I&C) technology of the Borssele Dutch

nuclear power plant, operated by EPZ

(Elektriciteits-Productiemaatschappij

Zuid-Nederland). The project started

in 2014 and included the installation of

a new reactor control and limitation

system to monitor the operation of

the plant and interfere in case of any

deviations to shut down the reactor

safely.

Framatome’s I&C teams designed

and engineered the new systems,

manufactured the cabinets, performed

a six months test period and finally

installed and commissioned the

systems at the plant during the 2017

outage.

The project was completed

according to budget and schedule.

“The successful implementation

of the digital systems based on the

TELEPERM XS platform marks

another milestone in the long-lasting

cooperation between EPZ and

Framatome, already starting with the

333

NEWS

News


atw Vol. 63 (2018) | Issue 5 ı May

Operating Results December 2017

334

NEWS

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated. gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto BWR FI 910 880 744 677 427 7 422 331 254 654 186 100.00 94.62 99.13 93.24 100.06 93.11

OL2 Olkiluoto BWR FI 910 880 744 687 897 6 482 042 244 299 181 100.00 81.41 99.91 80.52 101.60 81.31

KCB Borssele PWR NL 512 484 744 381 468 3 402 478 158 206 919 99.89 76.34 99.89 76.67 100.45 74.52

KKB 1 Beznau 1,2,7) PWR CH 380 365 0 0 0 124 746 087 0 0 0 0 0 0

KKB 2 Beznau 7) PWR CH 380 365 744 285 817 2 932 717 131 164 873 100.00 88.29 100.00 87.83 101.09 87.42

KKG Gösgen 7) PWR CH 1060 1010 744 795 652 8 583 952 305 194 587 100.00 93.02 99.99 92.67 100.89 92.44

KKM Mühleberg BWR CH 390 373 744 287 210 3 125 900 124 338 145 100.00 92.90 99.74 92.30 98.98 91.50

CNT-I Trillo PWR ES 1066 1003 744 789 963 8 530 707 239 024 424 100.00 92.10 100.00 91.83 99.16 90.86

Dukovany B1 PWR CZ 500 473 744 363 432 2 820 108 108 630 483 100.00 65.98 97.72 65.46 97.70 64.39

Dukovany B2 PWR CZ 500 473 700 349 496 3 299 909 104 622 538 94.09 76.84 93.68 76.31 93.95 75.34

Dukovany B3 PWR CZ 500 473 744 373 964 2 997 571 102 622 427 100.00 77.81 100.00 68.84 100.53 68.44

Dukovany B4 PWR CZ 500 473 744 371 658 2 743 590 103 271 741 100.00 71.76 100.00 62.75 99.91 62.64

Temelin B1 PWR CZ 1080 1030 189 188 794 8 853 135 106 481 294 25.40 93.66 25.24 93.61 23.50 93.58

Temelin B2 PWR CZ 1080 1030 744 806 383 7 625 624 101 489 946 100.00 80.18 99.98 79.88 100.36 80.60

Doel 1 PWR BE 454 433 744 324 212 3 601 775 134 214 747 100.00 91.55 95.55 90.68 95.68 90.33

Doel 2 PWR BE 454 433 744 330 500 3 598 619 132 252 268 100.00 92.13 97.01 91.54 97.48 89.99

Doel 3 PWR BE 1056 1006 0 0 6 732 621 251 169 221 0 72.26 0 72.10 0 72.40

Doel 4 PWR BE 1084 1033 744 818 714 7 873 391 254 545 842 100.00 84.51 100.00 83.82 100.49 82.19

Tihange 1 PWR BE 1009 962 744 760 691 3 575 802 290 838 876 100.00 41.96 100.00 41.23 101.59 40.46

Tihange 2 PWR BE 1055 1008 744 792 848 7 430 470 248 949 538 100.00 83.93 100.00 80.45 101.66 80.76

Tihange 3 PWR BE 1089 1038 740 799 873 9 414 133 268 894 830 99.43 99.74 98.50 99.59 98.65 98.58

Operating Results February 2018

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 672 896 786 1 854 307 342 046 366 100.00 100.00 93.97 93.97 89.88 88.15

KKE Emsland DWR 1406 1335 672 952 068 1 962 705 337 285 988 100.00 100.00 100.00 100.00 100.88 98.60

KWG Grohnde 1,2) DWR 1430 1360 554 750 245 1 727 583 368 355 162 82.44 91.67 77.64 86.68 77.60 84.79

KRB C Gundremmingen SWR 1344 1288 672 908 675 1 890 834 322 470 727 100.00 100.00 99.41 99.72 100.04 98.75

KKI-2 Isar DWR 1485 1410 672 1 000 793 2 083 701 343 682 024 100.00 100.00 100.00 99.99 100.09 98.84

KKP-2 Philippsburg DWR 1468 1402 672 981 334 2 043 937 357 211 453 100.00 100.00 100.00 99.96 98.08 97.02

GKN-II Neckarwestheim DWR 1400 1310 672 946 400 1 952 600 322 075 734 100.00 100.00 100.00 99.68 101.03 98.81

*)

Net-based values

(Czech and Swiss

nuclear power

plants gross-based)

1)

Refueling

2)

Inspection

3)

Repair

4)

Stretch-out-operation

5)

Stretch-in-operation

6)

Hereof traction supply

7)

Incl. steam supply

8)

New nominal

capacity since

January 2016

9)

Data for the Leibstadt

(CH) NPP will

be published in a

further issue of atw

BWR: Boiling

Water Reactor

PWR: Pressurised

Water Reactor

Source: VGB

construction of the plant. It is yet

another proof of the quality of our I&C

solutions”, said Frédéric Lelièvre,

Framatome’s Senior Executive Vice

President in charge of Sales, Regional

Platforms and the Instrumentation

and Control Business Unit.

“We see the implementation of

TELEPERM XS and different functional

improvements as a big step

forward to operate safely until the end

of lifetime of our power plant”, said

Jack de Waal, Project Manager at EPZ.

The Borssele pressurized water

reactor is the only operating nuclear

power plant in the Netherlands. The

plant was connected to the grid in

1973 and has a net electric output of

482 megawatt. In 2006, the Dutch

authorities approved a prolonged

operation until 2034.

| | (181061143), www.framatome.com

Market data

(All information is supplied without

guarantee.)

Nuclear Fuel Supply

Market Data

Information in current (nominal)

U.S.-$. No inflation adjustment of

prices on a base year. Separative work

data for the formerly “secondary

market”. Uranium prices [US-$/lb

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

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

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

(Separative work unit)].

January to December 2014

• Uranium: 28.10–42.00

• Conversion: 7.25–11.00

• Separative work: 86.00–98.00

January to December 2015

• Uranium: 35.00–39.75

• Conversion: 6.25–9.50

• Separative work: 58.00–92.00

2016

January to June 2016

• Uranium: 26.50–35.25

• Conversion: 6.25–6.75

• Separative work: 58.00–62.00

July to December 2016

• Uranium: 18.75–27.80

• Conversion: 5.50–6.50

• Separative work: 47.00–62.00

2017

January to June 2017

• Uranium: 19.25–26.50

• Conversion: 5.00–6.75

• Separative work: 42.00–50.00

July to December 2017

• Uranium: 19.50–26.00

• Conversion: 4.50–6.00

• Separative work: 39.00–43.00

News


atw Vol. 63 (2018) | Issue 5 ı May

2018

January 2018

• Uranium: 21.75–24.00

• Conversion: 6.00–7.00

• Separative work: 38.00–42.00

February 2018

• Uranium: 21.25–22.50

• Conversion: 6.25–7.25

• Separative work: 37.00–40.00

March 2018

• Uranium: 20.50–22.25

• Conversion: 6.50–7.50

• Separative work: 36.00–39.00

| | Source: Energy Intelligence

www.energyintel.com

| | Uranium spot market prices from 1980 to 2018 and from 2008 to 2018. The price range is shown.

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

335

NEWS

Cross-border Price

for Hard Coal

Cross-border price for hard coal in

[€/t TCE] and orders in [t TCE] for

use in power plants (TCE: tonnes of

coal equivalent, German border):

2012: 93.02; 27,453,635

2013: 79.12, 31,637,166

2014: 72.94, 30,591,663

2015: 67.90; 28,919,230

2016: 67.07; 29,787,178

| | Separative work and conversion market price ranges from 2008 to 2018. 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.

2017

I. quarter: 95.75; 8,385,071

II. quarter: 86.40; 5,094,233

III. quarter: 88.07; 5,504,908

IV. quarter: 94.07; 6,754,798

2017, year: 91.28, 25,739,010

| | Source: BAFA, some data provisional

www.bafa.de

EEX Trading Results

March 2018

(eex) In March 2018, the European

Energy Exchange (EEX) achieved a

total volume of 222.4 TWh on its

power derivatives markets (March

2017: 311.2 TWh). In Dutch Futures,

EEX achieved its highest volume so far

at 4.1 TWh. On the French market,

volumes increased by 19% to

24.4 TWh (March 2017: 20.4 TWh)

while volumes on the Italian market

increased by 12% to 36.0 TWh (March

2017: 32.1 TWh). In total, the German

and Austrian markets (Phelix-DE,

Phelix-AT and Phelix-DE/AT) recorded

a volume of 141.6 TWh (March

2017: 225.9 TWh).

The March volume comprised

119.2 TWh traded at EEX via Trade

Registration with subsequent clearing.

Clearing and settlement of all

exchange transactions was executed

by European Commodity Clearing

(ECC).

The Settlement Price for base

load contract (Phelix Futures) with

delivery in 2019 amounted to

36.08 €/MWh. The Settlement Price

for peak load contract (Phelix Futures)

with delivery in 2019 amounted to

44.73 €/MWh.

On the EEX markets for emission

allowances, the total trading volume

more than doubled to 246.6 million

tonnes of CO 2 in March (March 2017:

117.5 million tonnes of CO 2 ). In

particular, the development was

driven by a significant increase of

volumes on the secondary market

where EEX achieved a new monthly

record of 99.1 million tonnes of CO 2 .

Primary market auctions contributed

75.1 million tonnes of CO 2 to the total

volume. Furthermore, 72.5 million

tonnes of CO 2 were traded in EUA

Options which is the highest monthly

volume so far in this product and

almost 3-fold the previous record

volume which was traded in February

2018.

The EUA price with delivery in

December 2018 amounted to

9.99/13.68 €/ EUA (min./max.).

| | www.eex.com

MWV Crude Oil/Product Prices

February 2017

(mwv) According to information and

calculations by the Association of the

German Petroleum Industry MWV e.V.

in February 2018 the prices for

super fuel, fuel oil and heating oil

noted inconsistent compared with

the pre vious month January 2018.

The average gas station prices for Euro

super consisted of 137.27 €Cent

( January 2018: 136.84 €Cent, approx.

+0.31 % in brackets: each

information for pre vious month or

rather previous month comparison),

for diesel fuel of 119.50 €Cent

(120.48; -0.81 %) and for heating oil

(HEL) of 59.15 €Cent (62.27 €Cent,

-5.01 %).

The tax share for super with

a consumer price of 137.27 €Cent

(136.84 €Cent) consisted of

65.45 €Cent (47.68 %, 65.45 €Cent)

for the current constant mineral oil

tax share and 21.92 €Cent (current

rate: 19.0 % = const., 21.85 €Cent)

for the value added tax. The product

price (notation Rotterdam) consisted

of 38.08 €Cent (27.74 %, 40.17 €Cent)

and the gross margin consisted of

11.82 €Cent (8.61 %; 9.37 €Cent).

Thus the overall tax share for super

results of 66.68 % (66.83 %).

Worldwide crude oil prices

(monthly average price OPEC/Brent/

WTI, Source: U.S. EIA) were slightly

lower, approx. -4.26 % (+8.36 %) in

February 2018 compared to January

2018.

The market showed a stable

development with higher prices; each

in US-$/bbl: OPEC basket: 63.48

(66.85); UK-Brent: 65.32 (69.08);

West Texas Inter mediate (WTI):

62.23 (63.70).

| | www.mwv.de

News


atw Vol. 63 (2018) | Issue 5 ı May

336

Operating results 2017 – Part I*

REPORT

In 2017 the German nuclear power plants generated

84.63 billion kilowatt hours (kWh) of electricity gross.

The Gundremmingen B nuclear power plant ceased operation

on 31 December 2017 due to the revision of the

German Atomic Energy Act in the political aftermath of

the accidents in Fukushima, Japan, in 2011. Eight nuclear

power plants with an electric gross output of

11,357 MWe were in operation on 31 December 2017.

Three power plants in operation until 31 December

2017 achieved operating results with a gross production

greater than 10 billion kilowatt hours and two power

plants even produced more than 11 billion kilowatt hours.

German nuclear power plants achieved two of the

world’s ten best production results in 2017 (“Top Ten”). At

the end of 2017, 449 reactor units were in operation in 31

countries worldwide and 56 were under construction in 16

countries. The share of nuclear power in world electricity

production was around 11 %. German nuclear power

plants have been occupying top spots in electricity production

for decades thus providing an impressive demonstration

of their efficiency, availability and reliability.

The Shin Kori 3 nuclear power plant in the Republic Korea

(capacity: 1,475 MWe gross) achieved the world record

in electricity production in 2017 in its first year of commercial

operation with 12,921,000 MWh. The German nuclear

power plants Isar 2 (KKI 2, 11,523,513 MWh) and Emsland

(KKE, 11,323,704 MWh) took the ninth and tenth place.

* The reports with additional operating results of European

nuclear power plants will be published in a further issue of atw.

D

German nuclear power plant

Top Ten: Electricity production 1981 to 2017

Year

1981

1982

1983

1984

1985

1986

1987

1988

1989

1990

1991

1992

1993

1994

1995

1996

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

2007

2008

2009

2010

2011

2012

2013

2014

Top Ten: Nuclear Power Plants

World's

best

2 3 4 5 6 7 8 9 10

D D D

D D D D

D D D D

D D D D

D D D D D D D

D D D D D D

D D D D D D

D D D D D

D D D D D D D

D D D D D D

D D D D D D D

D D D D D D D

D D D D D D D

D D D D D D D

D D D D D D D

D D D D D D D

D D D D D D D

D D D D D D

D D D D D D D

D D D D D D

D D D D D D D D

D D D D D

D D D D

D D D D D

D D D D D D

D

D

D D D D

D D D

D D D

D

D D D

D

D D D D D

D D D D D D

D

D

D

D

D

D

D D D

D

D D D D

D D D D

2015 D D D D

2016 D D D

2017 D D

D

Report

Operating results 2017 – Part I*


atw Vol. 63 (2018) | Issue 5 ı May

337

Operating results of nuclear power plants in Germany 2016 and 2017

Nuclear power plant Rated power Gross electricity

generation

in MWh

Availability

factor*

in %

Energy availability

factor**

in %

REPORT

gross

in MWe

net

in MWe

2016 2017 2016 2017 2016 2017

Brokdorf KBR 1,480 1,410 11,503,003 5,778,146 93.28 51.68 93.08 48.23

Emsland KKE 1,406 1,335 11,113,993 11,323,704 94.25 93.28 94.13 93.13

Grohnde KWG 1,430 1,360 8,903,639 9,684,880 75.11 86.10 73.08 82.40

Gundremmingen KRB B**** 1,344 1,284 10,015,303 9,689,710 89.81 93.10 89.30 92.20

Gundremmingen KRB C 1,344 1,288 9,396,741 9,929,820 85.98 87.85 85.46 85.93

Isar KKI 2 1,485 1,410 11,990,925 11,523,513 95.86 91.53 95.68 91.15

Neckarwestheim GKN II 1,400 1,310 11,391,770 10,540,800 94.69 88.93 94.26 88.60

Philippsburg KKP 2 1,468 1,402 10,318,992 7,853,827 82.32 63.18 82.19 63.12

Total*** 11,357 10,799 84,634,367 76,324,400 88.91 81.95 88.42 80.47

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

of available time of a plant (tV) and the reference period (tN). The time availability factor is a degree

for the deployability of a power plant.

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

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

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

considerations. The available energy WV is the energy which can be generated in the reference period

due to the technical and operational condition of the plant. Energy availability factors in excess of 100 %

are thus impossible, as opposed to energy utilisation.

*** Inclusive of round up/down, rated power in 2017.

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

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

All data in this report as of 31 March 2018.

Report

Operating results 2017 – Part I*


atw Vol. 63 (2018) | Issue 5 ı May

338

Brokdorf

REPORT

Operating sequence in 2017

100

90

80

70

60

50

40

30

20

10

Electrical output in %

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

In the year 2017 nuclear power plant Brokdorf (KBR) was with an

availability factor of 48.23 % in total 4,527.21 operating hours on the

grid.

Due to an increased oxide layer thickness on the fuel element

cladding tubes (ME 2017/02) occurred an unplanned revision

extension of 156 days. The revision was completed on 30 July 2017

after 177 revision-days. Due to the determination of ME 02/2017

“ increased oxide layer thickness on the cladding tubes of fuel

elements” the thermal reactor power was first limited to a maximum

88 % until 14 September 2017. Afterwards the thermal reactor power

was limited in the framework of the modified operating conditions

for the 30 th operation cycle, with a coolant temperature lowered

100 by 3 K, to maximum 95 %.

On 9 December 2017 a failure occurred on one of the primary

coolant 80 pumps. The plant stabilised itself with a 3-pump operation at

a reactor capacity of 40 % and a generator power of approx. 570 MW.

The 60 cause of the failure was a malfunction of the under-voltage

release inside the circuit breaker. The under-voltage release was

exchanged.

40

Unplanned shutdowns and reactor/turbine trip

None.

Power reductions above 10 % and longer than for 24 h

See operating diagram:

• Operation with lowered reactor power of max.. 88 % due to an

oxid layer problem at the fuel element cladding tubes

• Runback for a grid-supporting power control

Delivery of fuel elements

During the reporting year no fuel elements transportations were

carried out.

Waste management status

By the end of the year 2017 30 loaded CASTOR © cask were located at

the on-site intermediate storage Brokdorf.

20

Planned shutdowns

After reaching the natural cycle end on 31 December 2016 a 35-day

0

stretch-out operation directly adjoined. On 4 February 2017 the

plant was shutdown for the 29 th refuelling and plant revision:

The revision included the following priorities:

• Reactor Offload of the RPV.

Positionierung: Inspection of FE, SE, SC.

Clarification of cause ME2017/02.

Bezug, links, unten

Exchange of the BAT-cable.

Load of 72 fresh fuel elements.

• Main VGB: coolant HKS6K pump YD20 30 Ring % exchange of e-motor.

atw: 100 60

Inspection

0 0

of the axial bearing.

• Recuperation heat exchangers Pressure testing.

• HP-Cooler TA11/12 B001 Pressure testing.

• Coolant Works on pump stationary head VE10/20.

• Turbine Standard service, Bearing inspection SB14.

• Transformers Exchange CS22/CT41.

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

Report

Operating results 2017 – Part I*


atw Vol. 63 (2018) | Issue 5 ı May

339

Operating data

Review period 2017

REPORT

Plant operator: PreussenElektra GmbH

Shareholder/Owner: PreussenElektra GmbH (80 %),

Vattenfall Europe Nuclear Energy GmbH (20 %)

Plant name: Kernkraftwerk Brokdorf (KBR)

Address: PreussenElektra GmbH, Kernkraftwerk Brokdorf,

25576 Brokdorf, Germany

Phone: 04829 752560, Telefax: 04829 511

Web: www.preussenelektra.de

100

90

80

93

Availability factor in %

Capacity factor in %

79

84

92

93

93

93

First synchronisation: 10-14-1986

Date of commercial operation: 12-22-1986

Design electrical rating (gross):

1,480 MW

Design electrical rating (net):

1,410 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

70

60

50

40

44

The following operating results were achieved:

Operating period, reactor:

4,527 h

Gross electrical energy generated in 2017: 5,778,146 MWh

Net electrical energy generated in 2017: 5,480,413 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2017:

340,192,058 MWh

Net electrical energy generated since

first synchronisation until 12-31-2017:

323,410,879 MWh

Availability factor in 2017: 51.68 %

Availability factor since

date of commercial operation: 89.82 %

Capacity factor 2017: 48.23 %

Capacity factor since

date of commercial operation: 86.41 %

Downtime

(schedule and forced) in 2017: 48.32 %

Number of reactor scrams 2017: 0

Licensed annual emission limits in 2017:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.0 · 10 15 Bq

6.0 · 10 9 Bq

5.55 · 10 10 Bq

30

20

10

0

10

9

8

7

6

94 84

2010 2011

84

2012

93

2013

93

2014

93

2015

Collective radiation dose of own

and outside personnel in Sv

93

2016

52

2017

Proportion of licensed annual emission limits

for radioactive materials in 2017 for:

Emission of noble gases with plant exhaust air: 0.017 %

Emission of iodine-131 with plant exhaust air: 0.001 %

Emission of nuclear fission and activation products

with plant waste water (excluding tritium): 0.000 %

Collective dose:

0.132 Sv

5

4

3

2

1

0

0.18 0.22

2010 2011

0.13

2012

0.22

2013

0.17

2014

0.14

2015

0.14

2016

0.13

2017

Report

Operating results 2017 – Part I*


atw Vol. 63 (2018) | Issue 5 ı May

340

Emsland

REPORT

Operating sequence in 2017

Electrical output in %

100

90

80

70

60

50

40

30

20

10

0

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

Apart from the overall maintenance outage in May and June and the

refuelling outage by the end of the year 2016 the Emsland nuclear

power plant had been operating uninterrupted and mainly at full

load during the review period 2017. Producing a gross power

generation of 11,323,704 MWh with a capacity factor of 93.13 %

the power plant achieved again a very good operating result.

Planned shutdowns

29 th refuelling and overall maintenance inspection:

Bye the end of 2016 the KKE did a second refuelling outage ending at

7 January 2017. Main task was the replacement of 16 fuel elements.

30 rd Refuelling and overall maintenance outage:

100 The annual outage was scheduled for the period 13 May 1 June. The

outage took 18.5 days from breaker to breaker. In addition to the

replacement 80

of 24 fuel elements the following major maintenance

and inspection activities were carried out:

• 60 Inspection of core and reactor pressure vessel internals.

• Inspection of a reactor coolant pump.


40

Inspection of pressurizer valves.

• Eddy current testing of steam-generator tubes.

20

• Pressure test on different coolers and tanks.

• Inspection of a main condensate pump.

0

• Maintenance works on different transformers.

• Different automatic non-destructive examinations.

Unplanned shutdowns and reactor/turbine trip

None.

Power reductions above 10 % and longer than for 24 h

Stretch-out operation from 16 April to 13 May.

National Peer Review

Risk assessment.

WANO Review/Technical Support Mission

WANO Follow-up.

Delivery of fuel elements

• 24 Uranium-fuel elements.

Waste management status

5 CASTOR © cask loading were carried out during the review period

2017. At the end of the year 43 loaded casks were stored in the local

interim storage facility.

Positionierung:

Bezug, links, unten

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

VGB: HKS6K 30 %

atw: 100 60 0 0

Report

Operating results 2017 – Part I*


atw Vol. 63 (2018) | Issue 5 ı May

341

Operating data

Review period 2017

REPORT

Plant operator: Kernkraftwerke Lippe-Ems GmbH

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

PreussenElektra GmbH (12,5 %)

Plant name: Kernkraftwerk Emsland (KKE)

Address: Kernkraftwerk Emsland,

Am Hilgenberg , 49811 Lingen, Germany

Phone: 0591 806-1612

Web: www.rwe.com

100

90

80

Availability factor in %

Capacity factor in %

94 95 95

95

95

91

94

93

First synchronisation: 04-19-1988

Date of commercial operation: 06-20-1988

Design electrical rating (gross):

1,406 MW

Design electrical rating (net):

1,335 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

70

60

50

40

The following operating results were achieved:

Operating period, reactor:

8,183 h

Gross electrical energy generated in 2017:

11,323,704 MWh

Net electrical energy generated in 2017:

10,751,526 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2017:

335,323,283 MWh

Net electrical energy generated since

first synchronisation until 12-31-2017:

317,914,871 MWh

Availability factor in 2017: 93.28 %

Availability factor since

date of commercial operation: 94.04 %

Capacity factor 2017: 93.13 %

Capacity factor since

date of commercial operation: 93.90 %

Downtime

(schedule and forced) in 2017: 6.72 %

Number of reactor scrams 2017: 0

Licensed annual emission limits in 2017:

Emission of noble gases with plant exhaust air:

Emission of iodine-131 with plant exhaust air:

(incl. H-3 and C-14)

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

1.0 · 10 15 Bq

5.0 · 10 9 Bq

3.7 · 10 10 Bq

Proportion of licensed annual emission limits

for radioactive materials in 2017 for:

Emission of noble gases with plant exhaust air: 0.09 %

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

(incl. H-3 and C-14)

Emission of nuclear fission and activation products

with plant waste water (excluding tritium): 0.00 %

Collective dose:

0.093 Sv

30

20

10

0

10

9

8

7

6

5

4

3

2

1

95 95 95

2010 2011 2012

95

2013

95

2014

91

2015

Collective radiation dose of own

and outside personnel in Sv

94

2016

93

2017

0

0.15 0.07 0.09

2010 2011 2012

0.08

2013

0.06

2014

0.10

2015

0.05 0.09

2016 2017

Report

Operating results 2017 – Part I*


atw Vol. 63 (2018) | Issue 5 ı May

342

Grohnde

REPORT

Operating sequence in 2017

100

90

80

70

60

50

40

30

20

10

Electrical output in %

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

During the reporting year 2017 the nuclear power plant Grohnde

was put off the grid for a 39,5-day revision with refuelling and

achieved an availability factor of 86.1 % The gross production

amounted to 9,684,880 MWh.

Opposite to the scheduled 33-day downtime the revision extended

by 159 hours due to the assembly of a pipe blend inside the secondary

cooling water system for secured intermediate cooling system, the

exchange of plate springs at the upper central framework, the

recovery of a missing fuel element centring pin on the placement

console as well as the repair on a input valve at the main steam valve

station.

The plant was additionally taken off the grid for eleven days in

100 November due to the repair of a small leakage at the Δp- measurement

on one of the primary coolant pumps.

80

Planned shutdowns

460March to 12 April: 34 th Refuelling and revision:

Nuclear power plant Grohnde was shut down as scheduled after a

48-day

40

stretch-out operation on 4 March for the 34 th revision and

refuelling.

20

The main planned works during this year’s revision were:

• Refuelling and implementation of 60 fresh fuel elements.

0

• Complete inspection of 23 fuel elements.

• Eddy current testing of 37 control elements.

• Visual examination of 16 throttle bodies.

• Oxide layer thickness measuring on fuel elements.

• Pressure Positionierung:

testing on primary circuit.

• Secondary pressure testing on the steam generator.

Bezug, links, unten

• Ultrasonic testing of the reactor pressure vessel (RPV) and RPV

nozzles. X = 20,475 Y = 95,25 B = 173,5 H = 38,2 power plant Krsko (Slovenia).

• Change VGB: of HKS6K 28 hold down 30 spring % assemblies of the upper central

framework.

atw: 100 60 0 0

• Examination due to breakage of a fuel element-centring pin in the

lower central framework.

• Exchange of the motor on one of the primary coolant pumps.

• Exchange of the mechanical seal on two main coolant pumps.

• Revision of the axial bearings on a main coolant pump.

• Leakage testing of the containment vessel.

• Works and examination in the redundancies with the focus on the

tasks in the main redundancy 4/8 maintenance works on valves

and actuators as well as examination on vessels, batteries and

electronic junctions.

• Cleaning of nuclear intercooler.

• Revision of the turbine rotor low pressure-part 1 + 2 including

ultrasonic testing of the wheel discs.

• Reconstruction of a pump bow of the secondary cooling water

system for secure interim-coolant systems.

Due to a recognized and non-lockable leakage at the flanges of one of

the pre-control valve inside the valve station, during the short

starting process on 9 April a SCRAM was released at 18:05 at 8 % of

reactor capacity in order to shutdown the plant in subcritical cold

state for the repair.

Unplanned shutdowns and reactor/turbine trip

5 to 17 November: Downtime for repair of a small leakage.

Due to a small leakage at the Δp-measurement of main coolant pump

YD40 the plant was shut down on 5 November 2017. After the repair

of the defective pipe and non-destructive examinations as well as

on-site inspections of comparable small-bore pipes the plant was put

to operation on 17 November.

Power reductions above 10 % and longer than for 24 h

Load sequence operation was carried out in January due to the

requirements of the load distributor.

National Peer Review

In 2017 a National Peer Review on the issue “Risk Assessments” took

place at KWG.

WANO Review/Technical Support Mission

In 2017 KWG attended a technical support mission on the issue of

“Prevention of Contamination by Foreign Bodies“ at the nuclear

Delivery of fuel elements

In 2017 two fuel elements of the ANF GmbH Lingen and

28 Westinghouse fuel elements were delivered.

Waste management status

In 2017 no spent fuel casks were loaded.

Report

Operating results 2017 – Part I*


atw Vol. 63 (2018) | Issue 5 ı May

343

Operating data

Review period 2017

REPORT

Plant operator: Gemeinschaftskernkraftwerk Grohnde GmbH & Co. OHG

Shareholder/Owner: PreussenElektra GmbH (83,3 %),

Stadtwerke Bielefeld (16,7 %)

Plant name: Kernkraftwerk Grohnde (KWG)

Address: Gemeinschaftskernkraftwerk Grohnde GmbH & Co. OHG,

P.O. bx 12 30, 31857 Emmerthal, Germany

Phone: 05155 67-1

Web: www.preussenelektra.de

100

90

80

94

Availability factor in %

Capacity factor in %

84

95

89

84

89

82

First synchronisation: 09-05-1984

Date of commercial operation: 02-01-1985

Design electrical rating (gross):

1,430 MW

Design electrical rating (net):

1,360 MW

Reactor type:

PWR

Supplier:

Siemens/KWU

70

60

50

40

73

The following operating results were achieved:

Operating period, reactor:

7,552 h

Gross electrical energy generated in 2017: 9,684,880 MWh

Net electrical energy generated in 2017: 9,113,021 MWh

Gross electrical energy generated since

first synchronisation until 12-31-2017:

366,627,569 MWh

Net electrical energy generated since

first synchronisation until 12-31-2017:

346,630,034 MWh

Availability factor in 2017: 86.10 %

Availability factor since

date of commercial operation: 91.70 %

Capacity factor 2017: 82.40 %

Capacity factor since

date of commercial operation: 91.30 %

Downtime

(schedule and forced) in 2017: 13.90 %

Number of reactor scrams 2017: 0

Licensed annual emission limits in 2017:

Emission of noble gases with plant exhaust air: 9.0 · 10 14 Bq

Emission of iodine-131 with plant exhaust air: 7.5 · 10 9 Bq

Emission of nuclear fission and activation products

with plant waste water (excluding tritium):

5.55 · 10 10 Bq

30

20

10

0

10

9

8

7

6

94 84 95

2010 2011 2012

90