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AMNT 2018:

Opening Address

374 ı AMNT 2018: Best Paper

TESPA-ROD Code Prediction of the Fuel

Rod Behaviour During Long-term Storage

379 ı Research and Innovation

Safety Assessment of the Research Reactors FRM II and FR MZ

ISSN · 1431-5254

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389 ı Energy Policy, Economy and Law

The Unnoticed Loss of Carbon-free Generation in the United States

404 ı Decommissioning and Waste Management

Radiological Characterization of High-level Radioactive Waste


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atw Vol. 63 (2018) | Issue 6/7 ı June/July

Summary: Nuclear Power 2017 | 2018:

Figures, Facts, Assessments and Two Remarkable Conclusions

With the available data, the middle of the year is certainly a suitable time for a summary and an assessment of the

development of nuclear energy in the course of the previous year. But first, for a given reason, two current and certainly

noteworthy findings.

• With the commissioning of the 1,086 MWe (gross)

Yangjiang-5 nuclear power plant in China, not only has

the country's 40 th nuclear power plant gone into operation,

the total number of nuclear power plants in operation

worldwide has also reached another historic high.

After the Leningrad 2-1 (1,199 MWe) and Rostov-4

(1,070 MWe) units went into operation in Russia in

2018, 451 nuclear power plants are currently in operation

worldwide. With 421,553 MWe gross output and

398,094 MWe net output, there is also more nuclear

energy capacity available worldwide than ever before.

• Additionally, the contribution of nuclear power plants

increases in terms of security of supply and system

stability, depending on the structure of production in

the countries. Nuclear power plants have outstanding

characteristics in terms of load following capability.

With high possible load change rates and large power

strokes, i.e. the difference between minimum and

maximum power, nuclear power plants together with

other conventional power plants are not stopgap or

fillers in power supply systems, but the basis for grid

stability and integration of the volatile feed from

renewables. Nuclear power plants and conventional

power plants supply electricity when the renewables

are not available and their controllability opens the way

for renewable electricity when it is available at high

output due to the whims of the nature.

Now, back to the summary for 2017: With 449 nuclear

power plants one unit less were in operation than at the end

of the previous year. In particular, four units became critical

and/or were synchronised for the first time with the grid:

Fuqing-4 and Tianwan-3 in China, Chasnupp-4 in Pakistan,

and Rostov-4 in Russia. Five units stopped operation in

2017 and were finally shutdown: in Germany, after 33 years

of successful operation, the Gundremmingen-B NPP was

shutdown due to the political decision in 2011; in Japan,

the prototype fast breeder reactor Monju ceased operation;

in Sweden the Oskarshamn-1 unit and in Spain, after five

years of unsuccessful waiting for an application for further

operation, the Santa Maria de Garona plant stopped

electricity production.

For electricity generation capacities, the global nuclear

energy gross capacity of 420,383 MWe exceeded the

400,000-MW-mark again and the capacity was quasi on

the same level as in 2016 with 420,534 MWe due to high

capacities of the new reactors. Even, the net capacity

increased from 397,003 MWe in 2016 up to 397,193 MWe

with a plus of 190 MWe.

A further increase can also be registered for electricity

generation. The net production of 2,490 TWh is about 1 %

higher compared to last year with 2,477 TWh. Due to the

35 non-operating nuclear power plants in Japan in 2017

this is considerably lower than before the Fukushima

incident in 2011. However, two nuclear power units,

Takahama-3 and Takahama-4, were reconnected to the

grid in Japan in 2017. This means that a total of seven plants

have been put back into operation in Japan since 2011.

Thus, the nuclear power share to the overall energy

production remains at 11 %, the share of nuclear energy in

the entire global energy supply at about 4.5 % – these are

certainly two remarkable figures: The currently 414 active

nuclear power plants are able to provide electricity to

every tenth person worldwide or every twentieth person

worldwide covers its energy needs with nuclear power ...

as mentioned: Regionally and in each single nuclear

energy using countries the share of nuclear power in the

electricity production differs with a range of 4 % in China

– which means a doubling in the past five years – up to

almost 28 % in France. 13 states cover more than 30 % of

its electricity generation with nuclear. With 182 reactors

Europe remains the most important region using nuclear

energy. With a share of about 27 % almost every fourth

kilowatt-hour of electricity spent in European is generated

in nuclear power plants.

Regarding newly started projects in 2017, three projects

were implemented: In Bangladesh the Rooppur project

started; India started its third project at the Kudankulam

site and in the Republic of Korea the project Shin-Kori-5

was initiated. Thus globally 55 nuclear power plant units

with a 59,872 MWe gross – and 56,642 MWe net capacity

were under construction; due to commissioning two less

than in the previous year. Furthermore around 125 new

build projects were registered, which are currently in a

specific planning stage. Besides many projects are planned

in countries, which plan to enter the nuclear sector. For

another 100 nuclear power plant units exist already

preliminary plans.

These nuclear figures reflect a rather unspectacular

development of global nuclear energy with a constant or

rather slight decreasing share. A glance at the details, both

globally and from each single country show, that nuclear

energy can definitely expand its important role in the global

energy supply. On the one hand, this can be explained by

the advised operations times of existing reactors. Today,

60 years of operating time are technically and economically

reality and 80 years are under preparation, safety-related

feasible without any compromises and thus for many

today’s older plants already in preparation or realisation

phase. Thus, the nuclear “age pyramid” with many plants in

the range of 25 to 40 operating years, will have a rather

small influence during the next decades. The regulatory

and political environment shows, that these strategies are

accepted or even receive support through arguments such

as conserving resources, climate protection, favourable

and stable costs as well as supply reliability. The contribution

of innovative technical developments such as Small

Modular Reactors, which are currently being discussed and

developed in many different ways, is not exactly foreseeable.

But these developments show that the potential of

nuclear energy technology is far from exhausted – here we

are more at the beginning than at the end of the journey.

Christopher Weßelmann

– Editor in Chief –

359

EDITORIAL

Editorial

Figures, Facts, Assessments and Two Remarkable Conclusions


atw Vol. 63 (2018) | Issue 6/7 ı June/July

EDITORIAL 360

Kernenergiebilanzen 2017 | 2018:

Zahlen, Fakten, Einschätzungen und zwei beachtenswerte Feststellungen

Die Jahresmitte ist mit den vorliegenden Daten sicherlich ein geeigneter Zeitpunkt für ein Resümee und eine Einschätzung

der Entwicklung der Kernenergie im Vorjahresverlauf. Doch vorab aus gegebenem Anlass zwei aktuelle und sicherlich

beachtenswerte Feststellungen.

• Mit der Inbetriebnahme des 1.086 MWe (brutto) Kernkraftwerksblocks

Yangjiang 5 in China ist nicht nur das vierzigste

Kernkraftwerk des Landes in Betrieb gegangen, die

Gesamtzahl der in Betrieb befindlichen Kernkraftwerke

weltweit hat damit auch einen weiteren historischen

Höchststand erreicht. Nachdem im Jahresverlauf 2018 in

Russland die Blöcke Leningrad 2-1 (1.199 MWe) und

­Rostov 4 (1.070 MWe) den Betrieb aufgenommen hatten,

sind derzeit weltweit 451 Kernkraftwerke in Betrieb zu

verzeichnen. Mit 421.553 MWe Bruttoleistung sowie

398.094 MWe Nettoleistung steht weltweit auch so viel

Kernenergiekapazität zur Verfügung wie nie zuvor.

• Zudem steigt der Beitrag der Kernkraftwerke, je nach

Struktur der Erzeugung in den einzelnen Ländern differenziert,

in Bezug auf Versorgungssicherheit und Systemstabilität.

Kernkraftwerke weisen heraus ragende Eigenschaften

bei der Lastfolgefähigkeit auf. Mit hohen möglichen

Laständerungsgeschwindig keiten sowie großen

Leistungshüben, also der Differenz zwischen Minimal- und

Maximalleistung, sind die Kernkraftwerke gemeinsam mit

weiteren konventio nellen Kraftwerken nicht Notnagel oder

Füller in Stromversorgungssystemen, sondern Grundlage

für Netz stabilität und Integration der volatilen Einspeisung

aus Erneuerbaren. Kernkraftwerke und konventionelle

Erzeugung liefern dann Strom, wenn die Erneuerbaren

nicht verfügbar sind und sie machen durch ihre Regelfähigkeit

den Weg für erneuerbaren Strom dann frei, wenn

dieser aufgrund der Launen der Natur einmal mit hoher

Leistung verfügbar ist.

Doch nun zurück zur Jahresbilanz: Mit 449 Kernkraft werken

war Ende 2017 ein Block weniger in Betrieb als ein Jahr zuvor.

Im Einzelnen sind vier Blöcke kritisch geworden und wurden

erstmals mit dem Stromnetz synchronisiert: Fuqing 4 und

Tianwan 3 in China, Chasnupp 4 in Pakistan und Rostov 4 in

Russland. Fünf Kernkraftwerksblöcke stellten ihren Betrieb

ein: In Deutschland gemäß der politischen Entscheidung aus

dem Jahr 2011 nach 33 Jahren erfolgreichem Betrieb das

Kernkraftwerk Gund remmingen B; in Japan der proto typische

Schnelle Brutreaktor Monju; in der Republik Korea das erste

Kernkraftwerk des Landes, Kori 1; in Schweden der Block

Oskarshamn 1 und in Spanien hat der Betreiber des Kernkraftwerks

Santa Maria de Garona nach fünf Jahren erfolgloser

Beantragungsphase für eine Laufzeitverlängerung die endgültige

Stilllegung beschlossen.

Bei den Stromerzeugungskapazitäten lag die Brutto leistung

der Kernenergie weltweit mit 420.383 MWe deutlich über

der Marke von 400.000 MW und blieb aufgrund der hohen

Leistung der Neuanlagen quasi auf Vorjahresniveau von

420.534 MW. Die Nettoleistung erreichte 397.193 MWe und

lag damit sogar höher als der Vorjahreswert von 397.003 MW.

Ein erneut gutes Ergebnis kann die Kernenergie auch bei

der Stromerzeugung verzeichnen. Mit einer Nettoerzeugung

von 2.490 TWh lag diese rund 1,0 % höher als im Vorjahr mit

2.477 TWh. Aufgrund von seit 2011 weiterhin nicht in Betrieb

befindlichen 35 Kernkraftwerken in Japan ist diese aber noch

niedriger als vor dem Erdbeben mit Tsunami und Unfall in

Fukushima. Allerdings sind in Japan in 2017 mit Takahama 3

und Takahama 4 weitere zwei Kernkraftwerke wieder in

Betrieb gegangen. Somit sind seit 2011 insgesamt sieben

Anlagen in Japan wieder in Betrieb genommen worden.

Der Anteil an der gesamten weltweiten Strompro duktion lag

weiterhin bei 11 %; der Anteil der Kernenergie an der gesamten

weltweiten Energieversorgung bei rund 4,5 % – dies sind zwei

sicherlich weitere bemerkenswerte Zahlen: Die rund 414

derzeit aktiven Kernkraftwerke sind in der Lage, jeden zehnten

Menschen weltweit mit Strom zu versorgen oder jeder zwanzigste

Mensch weltweit deckt seinen Energiebedarf komplett

mit Kernenergie. Regional und in den einzelnen Kernenergie

nutzenden Ländern liegt der Anteil der Kern energie an der

Stromerzeugung in einer Spannbreite von inzwischen 4 % in

China – eine Verdoppelung innerhalb von 5 Jahren – bis fast

72 % in Frankreich. 13 Staaten decken mehr als 30 % ihrer

Stromerzeugung nuklear. Europa ist weiterhin mit 182 Reaktoren

die bedeutendste Kernenergie nutzende Region. In ihr

wird mit einem Anteil von rund 27 % rund jede vierte

verbrauchte Kilowattstunde Strom in Kernkraftwerken erzeugt.

Bei den neu begonnenen Projekten sind für das Jahr 2017

drei Vorhaben zu verzeichnen: Im Newcomer-Land Bangladesh

wurde gemeinsam mit dem russischen Partner mit dem Bau des

ersten Blocks am Standort Rooppur begonnen, Indien hat die

Errichtung des dritten Blocks in Kudankulam in Angriff

enommen und in der Republik Korea startete das Projekt

Shin-Kori 5. Damit waren weltweit 55 Kernkraftwerksblöcke

mit 59.872 MWe Brutto- und 56.642 MWe Nettoleistung in

Bau; aufgrund der Neuinbetriebnahmen zwei weniger als ein

Jahr zuvor. Darüber hinaus sind rund 125 Neubauprojekte zu

ver zeichnen, die sich im konkreten Planungsstadium befinden.

Viele dieser Projekte werden zudem in Ländern geplant, die

neu in die Kernenergie einsteigen wollen. Für weitere 100 Kernkraftwerksblöcke

bestehen Vorplanungen.

Diese Kernenergiezahlen spiegeln eine im Wesent lichen

unspektakuläre weitere absehbare Entwicklung für die Kernenergie

weltweit wider, mit eher geringfügigen Verän derungen.

Ein Blick auf die Details, sowohl global als auch der einzelnen

Länder, zeigt, dass Kernenergie durchaus zukünftig ihre

wichtige und steigende Rolle bei der weltweiten Energieversorgung

ausbauen kann: Zu begründen ist dies einerseits

sicherlich mit den avisierten Laufzeiten von bestehenden

Reaktoren. Heute sind 60 Jahre Laufzeit für die Anlagen

technisch- wirtschaftlich Realität und 80 Jahre in der Vorbereitung,

sicherheitstechnisch ohne Abstriche umsetzbar und

damit für viele heute ältere Anlagen in der Vorbereitungs- bzw.

Umsetzungs phase. Die kerntechnische „Alterspyramide“, mit

vielen Anlagen im Bereich von 25 bis 40 Betriebsjahren wird

daher in den nächsten Jahrzehnten kaum einen Einfluss haben.

Auch zeigt sich im regula torischen und politischen Umfeld,

dass diese Strategie akzeptiert oder gar mit Argumenten wie

Ressourcen schonung, Klimaschutz, günstige und stabile Kosten

sowie Versorgungssicherheit und Netzstabilität beim Umbau

des Stromversorgungs systems mit mehr volatilen Quellen

Unterstützung erfährt. Welchen Beitrag innovative, tech nische

Entwicklungen, wie unter anderem die derzeit in vielfältiger

Weise diskutierten und in Entwicklung befind lichen Kleinen-

Modularen Reaktoren (Small Modular Reactors) liefern

können, ist nicht konkret absehbar, zeigt aber, dass das Entwicklungspotenzial

der Kernenergie technik bei Weitem nicht

ausgeschöpft ist – hier befinden wir uns eher am Beginn, als am

Ende der Reise.

Christopher Weßelmann

– Chefredakteur –

Editorial

Summary: Nuclear Power 2017/2018: Figures, Facts, Assessments and Two Remarkable Conclusions


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atw Vol. 63 (2018) | Issue 6/7 ı June/July

362

Issue 6/7

June/July

CONTENTS

369

AMNT 2018:

Opening Address

| | View of TU Munich´s research site in Garching with the FRM I in the foreground and the Forschungs-Neutronenquelle

Heinz Maier-Leibnitz (FRM II) in the background. The 20 MWth research neutron source FRM II is in operation, the FRM I

under decommissioning after 43 years of successful operation for various research projects (Courtesy: TUM)

Editorial

Summary: Nuclear Power 2017 | 2018 . . . . . . . . 359

Kernenergiebilanzen 2017 | 2018 . . . . . . . . . . 360

Abstracts | English . . . . . . . . . . . . . . . . . . . 364

Abstracts | German . . . . . . . . . . . . . . . . . . . 365

Inside Nuclear with NucNet

Poland Faces Delays and Decisions

as It Makes Ambitious Plans to To Nuclear . . . . . 366

AMNT 2018

49 th Annual Meeting on Nuclear Technology

(AMNT 2018): Opening Address . . . . . . . . . . . 369

Ralf Güldner

374

| | Crystallographic length change of UO 2 /PuO 2 -fuel relative to

displacement per atom (dpa).

Best Paper

TESPA-ROD Code Prediction of the Fuel Rod

Behaviour During Long-term Storage . . . . . . . . 374

Heinz G. Sonnenburg

DAtF Notes. . . . . . . . . . . . . . . . . . . . . .377

369

| | AMNT 2018, Opening speech, Dr. Ralf Güldner, President, DAtF.

Spotlight on Nuclear Law

New Build Projects Abroad – A Challenge

for Regulation . . . . . . . . . . . . . . . . . . . . . . 378

Neubauprojekte im Ausland – eine

Herausforderung für die Regulierung. . . . . . . . 378

Christian Raetzke

Contents


atw Vol. 63 (2018) | Issue 6/7 ı June/July

Research and Innovation

Safety Assessment of the Research Reactors

FRM II and FR MZ After the Fukushima Event . . . 379

Axel Pichlmaier, Heiko Gerstenberg, Anton Kastenmüller, Christian

Krokowski, Ulrich Lichnovsky, Roland Schätzlein, Michael Schmidt,

Christopher Geppert, Klaus Eberhardt and Sergei Karpuk

|379

383

| | Overall view of the FRM II (foreground), the neutron guide hall (middle)

and the FRM I (”atomic egg”, now under decommissioning).

Decommissioning of Germany’s

First Nuclear Reactor . . . . . . . . . . . . . . . . . . 383

Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and

Anton Kastenmüller

| Visualisation of the unfiltered ventilations system.

|392

404

| | Pyramide.

Operation and New Build

Further Development of

a Thermal- Hydraulics Two-Phase Flow Tool . . . . 401

Verónica Jáuregui Chávez, Uwe Imke, Javier Jiménez and V.H.

Sánchez-Espinoza

Decommissioning and Waste Management

Special Features of Measurement

for the Radiological Characterization

of High-level Radioactive Waste . . . . . . . . . . . 404

Besonderheiten bei Messungen

zur radiologischen Charakterisierung

hochradioaktiver Abfälle . . . . . . . . . . . . . . . . 404

Marina Sokcic-Kostic and Roland Schultheis

| Zelle für den fernhantierten Umgang mit radioaktiven Stoffen.

363

CONTENTS

Energy Policy, Economy and Law

While You Were Sleeping:

The Unnoticed Loss of Carbon-free Generation

in the United States . . . . . . . . . . . . . . . . . . . 389

KTG Inside . . . . . . . . . . . . . . . . . . . . . . 408

News . . . . . . . . . . . . . . . . . . . . . . . . . 410

Chris Vlahoplus, Ed Baker, Sean Lawrie, Paul Quinlan and

Benjamin Lozier

German Secretarial Management ISO/TC 85/SC 6

Reactor Technology . . . . . . . . . . . . . . . . . . . 392

Deutsche Sekretariatsführung ISO/TC 85/SC 6

Reactor-Technology . . . . . . . . . . . . . . . . . . . 392

Janine Winkler and Michael Petri

Nuclear Today

Confidence in Nuclear Safeguards at Risk

as Trump Quits One Deal to Pursue Another. . . . 422

John Shepherd

Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . 388

Environment and Safety

Thermal Hydraulic Analysis of the

Convective Heat Transfer of an Air-cooled

BWR Spent Fuel Assembly . . . . . . . . . . . . . . . 397

AMNT 2019: Call for Papers . . . . . . . . . . . . . Insert

Christine Partmann, Christoph Schuster and Antonio Hurtado

Contents


atw Vol. 63 (2018) | Issue 6/7 ı June/July

364

ABSTRACTS | ENGLISH

Poland Faces Delays and Decisions as It

Makes Ambitious Plans to Go Nuclear

NucNet | Page 366

Poland has drawn up ambitious plans to build up to

6,000 MW of nuclear generating capacity, potentially

at two sites, by the late 2030s or early 2040s.

But the government is yet to take a final decision

and the deadline has pushed back several times,

with plans hampered by changes in government,

problems putting in place the right domestic legislation

and the need to find the right financing model.

Poland needs nuclear because of its low carbon

footprint and as a way to decrease the country’s

carbon emissions. Poland is no newcomer to nuclear

technology. The Polish National Centre for Nuclear

Research (NCBJ) has operated a research reactor at

Swierk. In 1971 the government made its first

binding decision to build a nuclear plant. The

project was formally scrapped in 1990.

49 th Annual Meeting on Nuclear Technology

(AMNT 2018): Opening Address

Ralf Güldner | Page 369

As in other years, DAtF and the German Nuclear

Society (KTG), offer a comprehensive program

with their 49 th Annual Meeting on Nuclear Technology,

giving insights into many aspects of nuclear

technology and contributing to the international

exchange of knowledge and experience in industry,

research, politics and administration. In keeping

with long-standing tradition, even in the year

preceding its 50th anniversary, on 7 and 8 May 2019

in Berlin, the AMNT remains the only conference

in Germany, and one of the few in Europe, that

combines all the issues surrounding nuclear

technology under one roof and is dedicated to every

sector of the industry.

TESPA-ROD Code Prediction of the Fuel Rod

Behaviour During Long-term Storage

Heinz G. Sonnenburg | Page 374

The fuel rod code TESPA-ROD is applicable to LOCA

transients and RIA transients. Recently, code

models have been implemented in order to predict

the transitional fuel rod behaviour during longterm

storage. In particular modelling for both

long-term fuel swelling and associated helium gas

release have been implemented. First TESPA-ROD

code predictions for the long-term transient,

including wet storage, drying procedure and dry

storage indicate gap closure between fuel and

cladding. Thus, stress level in the cladding may

depend on both the internal fission gas pressure and

the fuel/cladding mechanical interaction.

New Build Projects Abroad – A Challenge

for Regulation

Christian Raetzke | Page 378

Numerous reactors are under construction or in the

planning stage worldwide. However, compared to

the situation a few decades ago, when the majority

of the plants in operation today were built,

the project models and boundary conditions are

much more diverse, so that traditional models of

regulation, approval and supervision (regulation)

sometimes reach their limits. The article provides

examples of new challenges. Regulation must find

new answers to the challenges. However, it must

not ignore the proven principles and instruments in

order to ensure nuclear safety.

Safety Assessment of the Research Reactors

FRM II And FR MZ After the Fukushima

Event

Axel Pichlmaier, Heiko Gerstenberg, Anton

Kastenmüller, Christian Krokowski, Ulrich Lichnovsky,

Roland Schätzlein, Michael Schmidt,

Christopher Geppert, Klaus Eberhardt

and Sergei Karpuk | Page 379

After the events at the Fukushima-I nuclear power

plant (NPP) in 2011 the Reaktorsicherheitskommission

(RSK) has carried out an overall assessment

of the German nuclear fleet with respect to extreme

(beyond design base) events. This paper deals only

with the research reactors (RR) FRM II (Garching)

and FR MZ (Mainz). The findings of the RSK, its

recommendations and their status of implementation

will be presented.

Decommissioning of Germany’s First

Nuclear Reactor

Ulrich Lichnovsky, Julia Rehberger,

Axel Pichlmaier and Anton Kastenmüller | Page 383

FRM started operating in 1957 as the first nuclear

reactor in Germany. Reactor operation ended in

2000. Licensing procedures for the deconstruction

and dismantling of the reactor started in 1998. In

2014 the Technical University of Munich (TUM) was

granted the license to decommission the reactor.

The article describes the (long) way to the license

for dismantling of the reactor and gives a short overview

of the current state of the decommissioning

project. Results of the (pre-)licensing stage are

presented: disposal of spent nuclear fuel (SNF) and

preparation of the safety report containing details

on fire protection, radiological characterization

(neutron activation and contamination), waste

management and safety analysis. With regard to

the current state of the project we will discuss:

clearance of material and current obstacles.

While You were Sleeping:

The Unnoticed Loss of Carbon-free

Generation in the United States

Chris Vlahoplus, Ed Baker, Sean Lawrie,

Paul Quinlan and Benjamin Lozier | Page 389

The United States has embarked on actions to

combat climate change by putting a focus on

lowering the carbon emissions from the electric

generation sector. A pillar of this approach is to

promote the greater use of renewable resources,

such as wind and solar. The past decade has seen

significant growth in carbon-free energy from wind

and solar. Generation from these resources reached

333,000 GWh in 2017. However, unbeknownst to

many who care about climate change, most of the

progress made to date through renewables is at

significant risk due to the loss or potential loss of

more than 228,000 GWh of nuclear carbon-free

generation.

German Secretarial Management ISO/TC

85/SC 6 Reactor Technology

Janine Winkler and Michael Petri | Page 392

On behalf of the Federal Ministry for the Environment,

Nature Conservation and Nuclear Safety

(BMU) and represented by the Office of the Nuclear

Safety Standards Commission (KTA), DIN has taken

over the Secretariat of ISO/TC 85/SC 6 Reactor

technology in conjunction with China in 2018. The

new role provides an opportunity to increase

German participation and influence in the field

of International Standardization, for instance

via conversion of German Industrial and KTA

Standards into International Standards. This

demonstrates that Germany is willing to actively

participate in the ongoing efforts to increase

Nuclear Safety in the peaceful use of Nuclear

Energy.

Thermal Hydraulic Analysis

of the Convective Heat Transfer of an

Air-cooled BWR Spent Fuel Assembly

Christine Partmann, Christoph Schuster

and Antonio Hurtado | Page 397

Since the reactor accident in Fukushima Daiichi,

the vulnerability of spent fuel pools (SFP) is more

focused in nuclear safety research. This paper

presents the experimental findings about the

convective heat transfer of a boiling water reactor

(BWR) spent FA under the absence of water. These

studies are performed within the joint project

SINABEL that is funded by the German Federal

Ministry of Education and Research to investigate

the thermal hydraulics of selected accident

scenarios in SFP experimentally and numerically.

Further Development of a Thermal-

Hydraulics Two-Phase Flow Tool

Verónica Jáuregui Chávez, Uwe Imke,

Javier Jiménez and V.H. Sánchez-Espinoza | Page 401

The numerical simulation tool TWOPORFLOW is

under development at the Institute for Neutron

Physics and Reactor Technology (INR) of the

Karlsruhe Institute of Technology (KIT). TWOPOR-

FLOW is a thermal-hydraulics code that is able to

simulate single- and two-phase flow in a structured

or unstructured porous medium using a flexible 3-D

Cartesian geometry. The main purpose of this work

is the extension, improvement and validation of

TWOPORFLOW in order to simulate the thermalhydraulic

behavior of Boiling Water Reactor (BWR)

cores.

Special Features of Measurement

for the Radiological Characterization

of High-level Radioactive Waste

Marina Sokcic-Kostic and

Roland Schultheis | Page 404

In nuclear power plants occasionally highly radioactive

waste is produced, such as fragments of

defective fuel elements or filters from hot cells.

NUKEM Technologies Engineering Services has

designed and implemented waste treatment options

for such waste in projects that characterise highlevel

radioactive waste and condition it in accordance

with the requirements for long-term storage.

This also includes a volume reduction to minimize

future storage costs. The focus of this article is on

the measurement of highly active waste and its

implications.

Confidence in Nuclear Safeguards at Risk as

Trump Quits One Deal to Pursue Another

John Shepherd | Page 422

Confidence in nuclear safeguards at risk as Trump

quits one deal to pursue another. By the time you sit

down to read this article, Donald Trump and Kim

Jong Un may have had an historic sit-down of their

own – in fact the first meeting between a sitting US

president and a leader of North Korea.

Abstracts | English


atw Vol. 63 (2018) | Issue 6/7 ı June/July

Polen steht mit seinen ehrgeizigen Plänen

zur Nutzung der Kernenergie vor

Verzögerungen und Entscheidungen

NucNet | Seite 366

Polen hat ehrgeizige Pläne formuliert, bis Ende

der 2030er oder Anfang der 2040er Jahre bis zu

6.000 MW Kernenergiekapazität an zwei Standorten

zu errichten. Dazu muss die Regierung eine

endgültige Entscheidung treffen, die mehrmals

verschoben wurde, wobei der Zeitplan durch

Regierungswechsel, Probleme bei der Umsetzung

der nationalen Gesetzgebung und der Notwendigkeit,

das geeignete Finanzierungsmodell zu finden,

behindert wurden. Polen benötigt Kernenergie

unter andrem wegen seiner günstigen CO 2 -Bilanz.

Polen ist kein Newcomer in der Nukleartechnik. Das

Polnische Nationale Zentrum für Kernforschung

(NCBJ) betreibt seit 1972 einen Forschungsreaktor

in Swierk. 1971 traf die Regierung ihren ersten

Beschluss zum Bau eines Kernkraftwerks. Das

Projekt wurde 1990 formell eingestellt.

49. Jahrestagung Kerntechnik (AMNT 2018):

Eröffnungsansprache

Ralf Güldner | Seite 369

Wie in den vergangenen Jahren bieten DAtF und

KTG mit der 49. Jahrestagung Kerntechnik ein

umfangreiches Programm, das Einblicke in viele

Aspekte der Kerntechnik gibt und zum internationalen

Wissens- und Erfahrungsaustausch in

Industrie, Forschung, Politik und Verwaltung

beiträgt. Auch im Jahr vor ihrem 50-jährigen

Jubiläum, am 7. und 8. Mai 2019 in Berlin, ist das

AMNT die einzige Konferenz in Deutschland und

eine der wenigen in Europa, die alle Themen rund

um die Kerntechnik unter einem Dach vereint und

sich allen Bereichen der Branche widmet.

TESPA-ROD Codeberechnungen zum Brennelementverhalten

bei Langzeitlagerung

Heinz G. Sonnenburg | Seite 374

Das Brennstab-Rechenprogramm TESPA-ROD

berechnet für Kühlmittelverluststörfälle als auch

für Reaktivitätsstörfälle das transiente Brenn stab-

Verhalten. Dieses Programm wurde jetzt erweitert,

um auch das Verhalten während der Langzeitlagerung

berechnen zu können. Insbesondere

wurden Modelle für das Langzeit-Brennstoffschwellen

und der damit verbundenen Helium-Freisetzung

implementiert. Erste TESPA-ROD Analysen

zeigen, dass für Langzeittransienten, die die Nasslagerung,

den Trocknungsprozess und die Trockenlagerung

umfassen, ein Spaltschluss zwischen

Brennstoff und Brennstabhülle möglich ist. Somit

kann die Hüllrohrspannung vom Spaltgasinnendruck

als auch von der mechanischen Interaktion

zwischen Hülle und Brennstoff bestimmt werden.

Neubauprojekte im Ausland – eine

Herausforderung für die Regulierung

Christian Raetzke | Seite 378

Weltweit sind zahlreiche Reaktoren in Bau oder in

konkreter Planung. Im Vergleich zur Situation vor

einigen Jahrzehnten, als das Gros der heute betriebenen

Anlagen errichtet wurde, sind die Projekt modelle

und Randbedingungen jedoch deutlich vielfältiger, so

dass überlieferte Modelle der Regelsetzung, Genehmigung

und Aufsicht ( Regulierung) zum Teil an ihre

Grenzen stoßen. Der Artikel liefert Beispiele für neue

Herausfor derungen. Auf die Herausforderungen

muss die Regulierung neue Antworten finden. Sie

darf allerdings dabei auch die bewährten Grundsätze

und Instrumente nicht außer Acht lassen, damit die

nukleare Sicherheit gewährleistet bleibt.

Sicherheitsbewertung der Forschungsreaktoren

FRM II und FR MZ nach dem

Fukushima-Ereignis

Axel Pichlmaier, Heiko Gerstenberg,

Anton Kastenmüller, Christian Krokowski,

Ulrich Lichnovsky, Roland Schätzlein,

Michael Schmidt, Christopher Geppert,

Klaus Eberhardt und Sergei Karpuk | Seite 379

Nach den Ereignissen im Kernkraftwerk Fukushima-I

im Jahr 2011 hat die Reaktor-Sicherheitskommission

(RSK) eine Gesamtbewertung der

deutschen Reaktoren im Hinblick auf extreme (über

die Auslegung hinausgehende) Ereignisse durchgeführt.

Diese Arbeit beschäftigt sich ausschließlich

mit den Forschungsreaktoren (RR) FRM II

( Garching) und FR MZ (Mainz). Die Ergebnisse der

RSK, ihre Empfehlungen und ihr Umsetzungsstand

werden vorgestellt.

Stilllegung des ersten Kernreaktors

Deutschlands

Ulrich Lichnovsky, Julia Rehberger,

Axel Pichlmaier und Anton Kastenmüller | Seite 383

Der FRM wurde 1957 als erster Kernreaktor in

Deutschland in Betrieb genommen. Der Reaktorbetrieb

wurde im Jahr 2000 eingestellt. Die Genehmigungsverfahren

für Stilllegung und Rückbau des

Reaktors begannen 1998. Im Jahr 2014 erhielt die

Technische Universität München (TUM) die Genehmigung

zur Stilllegung des Reaktors. Der Beitrag

beschreibt den (langen) Weg zur Genehmigung für

den Rückbau des Reaktors und gibt einen kurzen

Überblick über den aktuellen Stand des Stilllegungsprojekts.

Ergebnisse der (Vor-)Genehmigungsphase

werden vorgestellt: Entsorgung abgebrannter

Brennelemente und Erstellung des Sicherheitsberichts

mit Angaben zum Brandschutz, zur

radiologischen Charakterisierung (Neutronenaktivierung

und -kontamination), zur Abfallwirtschaft

und zur Sicherheitsanalyse. Im Hinblick

auf den aktuellen Stand des Projektes werden

diskutiert: Räumung von Material und aktuellen

Herausforderungen.

Ganz im Stillen: Der unbemerkte Verlust der

CO 2 -freien Stromerzeugung in den USA

Chris Vlahoplus, Ed Baker, Sean Lawrie,

Paul Quinlan und Benjamin Lozier | Page 389

Die Vereinigten Staaten von Amerika haben

Maßnahmen für den Klimaschutz eingeleitet. Ein

Schwerpunkt liegt auf der Minderung von

CO 2 -Emissionen bei der Stromerzeugung. Eine

Säule dieses Ansatzes ist die Förderung der verstärkten

Nutzung erneuerbarer Energien, wie Wind

und Sonne. In den letzten zehn Jahren hat die CO 2 -

freie Erzeugung aus Wind und Sonne auch in den

USA stark zugenommen. Die Erzeugung aus diesen

Ressourcen erreichte im Jahr 2017 333.000 GWh.

Für viele, die sich für das Thema Klimaschutz

interessieren, ist jedoch unbekannt, dass in Summe

die Fortschritte jedoch durch den Verlust oder

potenziellen Verlust von mehr als 228.000 GWh

kohlenstofffreier Stromerzeugung durch Kernenergie

quasi zunichte gemacht werden.

Deutsche Sekretariatsführung ISO/TC 85/SC

6 Reactor technology

Janine Winkler und Michael Petri | Seite 392

Im Auftrag des Bundesministeriums für Umwelt,

Naturschutz und nukleare Sicherheit (BMU), vertreten

durch die KTA-Geschäftsstelle, hat DIN ab

2018 die Sekretariatsführung des ISO/TC 85/SC 6

Reactor technology zusammen mit China übernommen.

Ziel ist hierbei, den deutschen Einfluss in

der internationalen Normung zu erhöhen und die

Möglichkeit wahrzunehmen, deutsche Normen und

KTA-Regeln in die internationalen Normen zu überführen.

Damit möchte Deutschland international

seinen Beitrag für die nukleare Sicherheit bei der

friedlichen Nutzung der Kernenergie leisten.

Thermohydraulische Analyse des

konvektiven Wärmeübergangs eines

luftgekühlten SWR-Brennelementes

Christine Partmann, Christoph Schuster

und Antonio Hurtado | Seite 397

Seit dem Reaktorunfall in Fukushima Daiichi ist die

Forschung zur Sicherheit abgebrannter Brennelemente

stärker ausgeprägt. Dieser Beitrag stellt

die experimentellen Ergebnisse zur konvektiven

Wärme übertragung eines Siedewasserreaktor-

Brennelements unter Abwesenheit von Wasser vor.

Diese Untersuchungen werden im Rahmen des vom

Bundesministerium für Bildung und Forschung

geförderten Verbundprojektes SINABEL durchgeführt,

um die thermische Hydraulik ausgewählter

Unfallszenarien im Brennelementlagerbecken

experimentell und numerisch zu untersuchen.

Weiterentwicklung eines thermohydraulischen

Zwei-Phasen-Strömungsmodells

Verónica Jáuregui Chávez, Uwe Imke,

Javier Jiménez und V.H. Sánchez-Espinoza | Seite 401

Das numerische Simulationstool TWOPORFLOW

wird am Institut für Neutronenphysik und Reaktortechnik

(INR) des Karlsruher Instituts für Technologie

(KIT) entwickelt. TWOPORFLOW ist ein thermohydraulischer

Code, der in der Lage ist, ein- und

zweiphasige Strömungen in einem strukturierten

oder unstrukturierten porösen Medium unter Verwendung

einer flexiblen 3-D kartesischen Geometrie

zu simulieren. Das Hauptziel dieser Arbeit ist die Erweiterung,

Verbesserung und Validierung von TWO-

PORFLOW, um das thermisch- hydraulische Verhalten

in Siedewasserreaktor kernen zu simulieren.

Besonderheiten bei Messungen

zur radiologischen Charakterisierung

hochradioaktiver Abfälle

Marina Sokcic-Kostic und

Roland Schultheis | Seite 404

Beim Betrieb von Kernkraftwerken fallen gelegentlich

hochradioaktive Abfälle an, wie zum Beispiel

Bruchstücke defekter Brennelemente oder Filter

von heißen Zellen. Für solche Abfälle hat die

NUKEM Technologies Engineering Services Abfallbehandlungsmöglichkeiten

konzipiert und in

Projekten umgesetzt, die hochradioaktive Abfälle

charakterisieren und entsprechend den Anforderungen

für die Langzeitlagerung konditionieren.

Dies schließt auch eine Volumenverminderung ein,

um so die künftigen Lagerkosten zu minimieren.

Der Schwerpunkt dieses Artikels liegt in der

Messung von hochaktivem Abfall und seinen

Implikationen.

Trump gefährdet das Vertrauen

in Safeguards mit der Aufgabe eines

Abkommens zugunsten anderer Ziele

John Shepherd | Seite 422

Das Vertrauen in das Safeguards-System gerät in

Gefahr, da der U.S.-amerikanische Präsident Trump

ein Abkommen aufgibt – mit dem Iran –, um ein

anderes zu verfolgen. Bis zur Veröffentlichung

dieses Beitrags hatten Donald Trump und Kim Jong

Un vielleicht ein historisches Treffen – das erste

Treffen zwischen einem amtierenden US-Präsidenten

und einem Führer Nordkoreas.

365

ABSTRACTS | GERMAN

Abstracts | German


atw Vol. 63 (2018) | Issue 6/7 ı June/July

366

INSIDE NUCLEAR WITH NUCNET

Poland Faces Delays and Decisions as It

Makes Ambitious Plans to Go Nuclear

Poland has drawn up ambitious plans to build up to 6 GW of nuclear generating capacity, potentially at

two sites, by the late 2030s or early 2040s. But the government is yet to take a final decision and the deadline

has pushed back several times, with plans hampered by changes in government, problems putting in place

the right domestic legislation and the need to find the right financing model.

Poland is no newcomer to nuclear technology.

The Polish National Centre for Nuclear Research (NCBJ) has

operated a research reactor at Swierk, on the outskirts of

the capital Warsaw, since the mid-1970s. According to the

International Atomic Energy Agency (IAEA), the first plans

to launch nuclear power were drawn up in 1956. Nuclear

energy was seen as a tool that would enable reductions in

internal coal consumption, on which the whole Polish

energy sector was based. With nuclear, it would have been

possible to save precious natural resources or to export

them.

In 1971 the government made its first binding decision

to build a nuclear plant. A year later it designated

Zarnowiec, close to Poland’s Baltic coast in the northern

province of Pomerania, as the site. After a decade of planning,

construction of the four-unit Zarnowiec station began

in 1982. The station was intended to have four Sovietdesigned

VVER-440 pressurised water reactors, which

were to be manufactured by Skoda in Czechoslovakia. But

public opposition to the project arose in 1986 in the aftermath

of the Chernobyl disaster in neighbouring Soviet

Ukraine and work on Zarnowiec was suspended in 1989.

The project was formally scrapped in 1990.

Fast forward nearly two decades and Poland revived

its nuclear ambitions in 2009 as part of a drive to find

alternatives to ageing coal-fired electricity generation,

reduce greenhouse gas emissions and boost energy supply

security.

Poland, the largest economy among the EU’s eastern

members, generated 80 % of its electricity from coal in

2016, according to the International Energy Agency (IEA).

The remaining 20 % came from wind, hydro, other

fossil fuels and biofuels. This reliance on coal is a cheap

way to produce energy but provides an enormous headache

for any government trying to maintain economic

growth and meet ever-stricter EU greenhouse gas emission

targets.

Nuclear, which provides reliable baseload supply and

has overall lifecycle emissions that compare to wind power,

seemed the logical choice. In 2014 the Polish ministry of

economy adopted the Polish Nuclear Power Programme

(PPEJ), which included the construction of up to 6 GW of

capacity by 2035. According to the PPEJ, construction of a

first unit should be completed by the end of 2024, and a

second by 2029. The programme said a second nuclear

station should be ready before 2035.

In 2010, PGE EJ1 (Polska Grupa Energetyczna Energia

Jądrowa 1) was set up and charged with the planning and

construction of the first nuclear station with a capacity of

up to 3,750 MW. PGE EJ1’s job would include site search,

environmental impact assessment, various licensing

procedures, construction and subsequent operation. The

company is 70 % owned by Polish state-controlled utility

PGE. The remaining 30 % share is equally spread between

three state-run companies, Enea, KGHM Polska Miedz and

Tauron Polska Energia.

Under the proposed schedule in the PPEJ, site selection

and a tender process should have been concluded by the

end of 2016 and licensing by the end of 2018. In reality,

development of the nuclear programme has not gone to

schedule and these deadlines have become increasingly

unrealistic.

Delays . . . And some Progress

According to Professor Grzegorz Wrochna of the NCBJ one of

the problems was that the PPEJ had assumed the project

would take 10 years from the investment decision to

operation of the first reactor. This schedule was based on

International Atomic Energy Agency (IAEA) documents,

which were in turn based on the experience of other nuclear

countries. However, it soon became clear that Poland would

need 16 years to get its nuclear programme operational

because existing Polish laws did not allow many of the

regulatory processes to run in parallel to each other.

Delays also arose when, in December 2014, PGE EJ1

cancelled a contract with Woorley Parsons over slow

progress on site characterisation and licensing work.

PGE EJ1 awarded the contract to Woorley Parsons in 2013

with 2016 set as a deadline. PGE EJ1 had to take over the

contractor’s tasks.

However, there has been progress. In April 2017,

PGE EJ1 began environmental and site selection surveys at

two locations: Lubiatowo-Kopalino in the municipality of

Choczewo and Żarnowiec in the municipality of Krokowa,

both in northern Pomerania, west of the city of Gdansk.

The studies aim to determine the potential impact of the

project on both the environment and local residents. This

work is expected to be completed by 2020.

In October 2015 general elections in Poland brought

the conservative Law and Justice party to power and

responsibility for the nascent nuclear programme was

transferred to a newly formed ministry of energy. The new

energy minister Krzysztof Tchorzewski confirmed work on

the programme would continue, but that it needed to be

reviewed. The review was expected to be completed by

mid-2017, but that deadline was later extended to the end

of 2017 and was never met.

There have been few official updates from PGE EJ1

or the energy ministry on the status of the nuclear

programme.

In January 2017, however, Mr Tchorzewski said he had

received a mandate from the cabinet to present a new

financing model for the project. A few months later he said

the government had abandoned plans to finance the

nuclear project by way of contracts for difference – similar

to those set up for the Hinkley Point C nuclear station under

construction in England – because it was too costly for

consumers. He said the government would like to find

funding on a commercial basis, without state guarantees

on loans or electricity prices.

PGE EJ1 has stopped short of confirming a financing

scheme and has said all options remain on the drawing

Inside Nuclear with NucNet

Poland Faces Delays and Decisions as It Makes Ambitious Plans to Go Nuclear ı June/July


atw Vol. 63 (2018) | Issue 6/7 ı June/July

board. Recently, the Polish media has speculated about

possible government plans to involve state-owned energy

companies in the nuclear project. Officials from Orlen,

Poland’s largest oil refiner and petrol retailer, recently

hinted to journalists that the company would be interested

in cooperating with PGE EJ1 on the nuclear project.

In another twist, in September 2017, Jozef Sobolewski,

director of the Polish ministry of energy’s nuclear energy

department, told a parliamentary committee on nuclear

energy that the government was considering using

“ domestic” financing for construction of the first station.

He said the government did not want to have its decision

about the project dominated by financial markets. “It's not

a financial project, it’s an energy project”, he said.

Mr Sobolewski estimated the cost of building 1 GW of

nuclear capacity at € 2.8 bn to € 3.25 bn, based on the

assumption that a 3-GW station would be built.

Rafał Zasun, an editor at the specialised energy portal

Wysokie Napiecie, told NucNet that the main reason for

delaying the final decision on the nuclear programme is

the government’s inability to agree on its financing. “The

idea of building a nuclear power station has strong

opponents in government and in state-owned energy

companies”, he said.

According to Aleksandra Gawlikowska-Fyk, head of the

international economic relations and energy policy

programme at the Polish Institute of International Affairs,

the energy ministry is in the process of revising the national

energy policy framework for the first time since 2009.

Because there are many overlapping aspects, the review

impacts the nuclear programme’s schedule, she said.

A decision is now expected by mid-2018, according to

the latest reports.

Why Does Poland Need Nuclear?

Poland needs nuclear because of its low carbon footprint

and as a way to decrease the country’s carbon emissions,

said energy minister Krzysztof Tchorzewski. He told a

recent conference that Poland’s ongoing large-scale

investment in three new coal-fired power plants may be

the country’s last fossil fuel venture, indicating a possible

energy shift in the country’s revived plans to embrace

nuclear power.

There are other factors that point to the need for

nuclear. Poland signed up to the EU’s target to reduce

greenhouse gas emissions by 20 % from 1990 levels by

2020. It has had one of the fastest growing economies in

the EU for the past decade and electricity demand is

expected to grow by about 36 % by 2030.

“Poland needs to decrease emissions and nuclear offers

that”, Ms Gawlikowska-Fyk said. “This argument has been

used for years”.

“It is no secret that the European Commission expects a

vision of the future energy mix in the country and nuclear

is showcased by Poland as a way to reduce emissions”, she

said.

According to Ms Gawlikowska-Fyk, smog is “the

elephant in the room” and a significant influence on Polish

public opinion. A report by the World Health Organisation

(WHO) says that out of the 50 European cities most

affected by smog, 33 are in Poland. The WHO estimates

that around 50,000 Poles die every year due to illness

caused by air pollution.

But the prominence of coal mining and coal-related

industries in Poland presents challenges to every government

when it comes to energy sector reforms. Poland is the

second largest coal mining country in Europe, second to

Germany, and the coal industry employs 100,000 people.

However, experts have warned for years that the

cheapest sources of coal in the Silesian Basin are nearly

depleted and that the country’s mining sector will have to

prepare for higher costs in the future.

Meanwhile, work continues on the choice of a

technology for the project. The PPEJ did not shortlist a

technology; the only requirement is for reactors to be of

the Generation III/III+ design because of their improved

safety and 60-year design lifespan. The number of units

that will be built and their site configuration will depend

on the technology choice.

In December 2015 PGE EJ1 said five companies had

expressed an interest in supplying reactor technology. They

were GE-Hitachi Nuclear Energy Americas, Korea Electric

Power Corporation, SNC-Lavalin Nuclear Inc, Westinghouse

Electric Company and Areva (now Framatome). PGE EJ1

said at the time that preliminary discussions had been held

with all five.

A public tender for the construction of the first nuclear

power station was scheduled to be announced in late 2017

or early 2018, but this now seems unlikely.

Author

NucNet

The Independent Global Nuclear News Agency

Editor responsible for this story: Kamen Kraev

Avenue des Arts 56

1000 Brussels, Belgium

www.nucnet.org

INSIDE NUCLEAR WITH NUCNET 367

Inside Nuclear with NucNet

Poland Faces Delays and Decisions as It Makes Ambitious Plans to Go Nuclear ı June/July


atw Vol. 63 (2018) | Issue 6/7 ı June/July

CALENDAR 368

Calendar

2018

16.07.-19.07.2018

International Conference on Quality, Leadership

and Management in the Nuclear Industry –

15 th IAEA-FORATOM Management Systems Workshop.

Ottawa, Canada, organized in cooperation

between FORATOM and the International Atomic

Energy Agency, www.foratom.org, www.iaea.org

29.07.-02.08.2018

International Nuclear Physics Conference 2019.

Glasgow, United Kingdom, www.iop.org

30.07.-03.08.2018

14 th Joint ICTP-IAEA School on Nuclear

Knowledge Management. Trieste, Italy,

International Atomic Energy Agency (IAEA) and

Abdus Salam International Centre for Theoretical

Physics (ICTP), www.iaea.org

12.08.-17.08.2018

GOLDSCHMIDT Conference. Boston, USA,

Geochemical Society and the European Association

of Geochemistry, www.goldschmidt.info/2018

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

02.09.-06.09.2018

19 th International Nuclear Graphite Specialists

Meeting (INGSM-19). Shanghai Institute of Applied

Physics, Shanghai, China, ingsm.csp.escience.cn

03.09.-06.09.2018

Jahrestagung des Fachverbandes Strahlenschutz.

Dresden, Germany, Fachverband für Strahlenschutz

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

04.09.-05.09.2018.

8. Symposium Lagerung und Transport radioaktiver

Stoffe. Hannover, Germany, TÜV NORD

Akademie, www.tuev-nord.de

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

12.09.-14.09.2018

SaltMech IX – 9 th Conference on the Mechanical

Behavior of Salt. Hannover, Germany, Federal

Institute for Geosciences and Natural Resources

(BGR) Hannover, the Institute of Geomechanics (IfG)

Leipzig and the Technical University of Clausthal

(TUC), www.saltmech.com

16.09.-20.09.2018

55 th Annual Meeting on Hot Laboratories and

Remote Handling – HOTLAB 2018. Helsinki,

Finland, VTT and International Atomic Energy

Agency (IAEA), www.vtt.fi/sites/hotlab2018/

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

19.09.-21.09.2018

Workshop Sicherheitskonzepte Endlagerung.

Grimsel, Switzerland. Fachverband für Strahlenschutz

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

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.-03.10.2018

Fifteenth NEA Information Exchange Meeting on

ctinide and Fission Product Partitioning and

Transmutation. Manchester Hall, Manchester, UK,

OECD Nuclear Energy Agency (NEA), National

Nuclear Laboratory (NNL) in co‐operation with the

International Atomic Energy Agency (IAEA),

www.oecd-nea.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

08.10.-11.10.2018

World Energy Week. World Energy Council Council’s

Italian Member Committee, www.worldenergy.org

09.10.-11.10.2018

8 th International Conference on Simulation

Methods in Nuclear Science and Engineering.

Ottawa, Ontario, Canada, Canadian Nuclear Society

(CNS), www.cns-snc.ca

10.10.-11.10.2018

IGSC Symposium 2018 – Integrated Group for the

Safety Case; Current Understanding and Future

Direction for the Geological Disposal of Radioactive

Waste. Rotterdam, The Netherlands, OECD

Nuclear Energy Agency (NEA), www.oecd-nea.org

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

16.10.2018

The next steps for nuclear energy projects in the

UK. London, United Kingdom, Westminster Energy,

Environment & Transport Forum,

www.westminsterforumprojects.co.uk

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 Plants. 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

13.11.-15.11.2018

24 th QUENCH Workshop 2018. Karlsruhe, Germany,

Karlsruhe Institute of Technology in cooperation with

the International Atomic Energy Agency (IAEA),

quench.forschung.kit.edu

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

06.12.2018

Nuclear 2018. London, United Kingdom, Nuclear

Industry Association (NIA), www.niauk.org

2019

25.02.-26.02.2019

Symposium Anlagensicherung. Hamburg,

Germany, TÜV NORD Akademie, www.tuev-nord.de

10.03.-15.03.2019

83. Annual Meeting of DPG and DPG Spring

Meeting of the Atomic, Molecular, Plasma Physics

and Quantum Optics Section (SAMOP), incl.

Working Group on Energy. Rostock, Germany,

Deutsche Physikalische Gesellschaft e.V.,

www.dpg-physik.de

10.03.-14.03.2019

The 9 th International Symposium On

Supercritical- Water-Cooled Reactors (ISSCWR-9).

Vancouver Marriott Hotel, Vancouver, British

Columbia, Canada, Canadian Nuclear Society (CNS),

www.cns-snc.ca

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 – Save the Date!

Calendar


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49 th Annual Meeting on Nuclear Technology

(AMNT 2018): Opening Address

29 to 30 May 2018, Berlin

Ralf Güldner

Ladies and Gentlemen, on behalf of the DAtF and the German Nuclear Society (KTG), welcome to our

49 th ­Annual Meeting on Nuclear Technology in Berlin. As in other years, we offer a comprehensive program, giving

insights into many aspects of nuclear technology and contributing to the international exchange of knowledge and

experience in industry, research, politics and administration.

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AMNT 2018

Ladies and Gentlemen,

In keeping with long-standing tradition, even in the year

preceding its 50 th anniversary, which I’d like to invite you

to attend in Berlin on 7 and 8 May 2019, the AMNT remains

the only conference in Germany, and one of the few in

Europe, that combines all the issues surrounding nuclear

technology under one roof and is dedicated to every sector

of our industry.

Waste management – old challenges,

new structures

It is not only Germany as a venue but also the relevance

and complexity of the dismantling and waste management

issues that again make them one of the focal points of our

conference. The reorganization of nuclear waste management

in Germany, which has been negotiated and put into

legal form over the last few years, is as such broadly

complete following approval by the European Union and

the transfer of around EUR 24.1 billion to the state waste

management fund by the operators on 3 June 2017.

However, the transfer of responsibility for the site-based

interim storage facilities to the Federal Company for Interim

Storage (BGZ Gesellschaft für Zwischenlagerung mbH) is still

pending. But let me give you an overview of the status quo

based on the new structure.

The search for a final repository

is slowly gaining momentum

The Federal Office for the Safety of Nuclear Waste Management

(Bundesamt für kerntechnische Entsorgungssicherheit

– BfE) is, in its own words, “a growing authority

with growing responsibilities”. In the public perception,

according to the Site Selection Act (StandAG), the BfE is

currently complying with the second part of the task

assigned to it by the legislator: “The Federal Office for the

Safety of Nuclear Waste Management is the body responsible

for organising public participation in the site selection process.

It informs the public comprehensively and system atically

about the site selection process.” The breadth of issues dealt

with by the BfE’s publications is evidence of this.

At this point, I would also like to mention the work of

the national advisory committee which recently presented

its first report on the selection process for a final repository

site. Broad public participation in the work of the advisory

committee is essential for performing the tasks assigned to

it and is therefore desirable.

Other key tasks of the BfE according to the Site Selection

Act (StandAG) are: specification of the exploration programmes,

examination of the project developer’s proposals

and the preparation of well-founded recom mendations

based on them as well as monitoring of the site selection

process. However, this initially requires pre liminary work

that is currently in process, including

the selection of site regions and the

sites to be explored.

The Federal Company for Radioactive

Waste Disposal (Bundesgesellschaft

für Endlagerung – BGE) is responsible

for this and Dr. Jörg Tietze, Acting

Head of Site Selection within the BGE,

will tell us about its work today. As the

project developer, the BGE has to

“ carry out the site selection process”

according to the will of the legislator.

The BGE which is also responsible, of course, for the

Konrad final repository has now announced a concrete

date for completion of the final repository in the first six

months of 2027. Measured against the original objective,

that is to say 2013, the criticism regarding the extent of the

delay, particularly on the part of the public authorities, for

example in Baden-Württemberg or Schleswig-Holstein, is

entirely comprehensible. However, we now see this fixing

of a date as a kind of voluntary commitment by which it

must be measured.

On 1 August of last year, the Company for Interim

Storage (BGZ Gesellschaft für Zwischenlagerung) became

the property of the federal government and took over the

responsibility for the central interim storage facilities in

Ahaus and Gorleben. From 1 January 2019, the company

will take over the decentralised interim storage facilities

with heat-generating waste and, from 1 January 2020, the

interim storage facilities for waste with negligible heat

generation.

The BGZ will also need the skills it has acquired in the

process for the construction of a central receiving store for

Konrad. This important task and the responsibility for all

interim storage facilities will broaden the BGZ’s range of

responsibilities and strengthen public perception of the

company.

Good and trusting collaboration

between state actors and private operators

is indispensable for public acceptance

A smooth working relationship between state agencies and

private operators as part of this reorganization is crucial

not only for safety and efficiency but also for public acceptance

of the newly created structures and regulations. The

operators are fully committed to the obligations agreed and

we also expect this from our contractual partners.

Dismantling: operators meeting

their obligations without any ifs or buts

The operators are committed, without any ifs or buts, to

their obligation to drive forward safe and efficient

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dismantling and have already received a whole series of

licenses for this. Unfortunately, challenges are arising that

are neither objectively justifiable nor covered by regulatory

requirements. This applies in particular to the release of

rubble from nuclear power plants and taking it to public

landfill sites. Some of the cited public criticism about

alleged health risks is inappropriate and does not

contribute to solving this task for the whole of society.

The rubble released from dismantling

is harmless to health

Unsettling the local residents of the landfill sites concerned

with unfounded and alarmist claims unnecessarily complicates

a safe and internationally recognised disposal path

and considerably hinders dismantling. It is gratifying that

socially recognised institutions, such as the German

Medical Association (Bundesärztekammer), are successfully

resisting the attempt to exploit them for the purpose of a

discrediting campaign. We would sincerely like to thank

the Board of the German Medical Association and its president,

Prof. Dr. Frank Ulrich Montgomery, for confirming

that the resolution taken by the 120 th German Medical

Assembly, which critically questions the 10 microsievert

concept, is scientifically untenable. It is very important and

very positive to see authorities, operators and policymakers

working together on this issue. Here I would like to

quote Franz Untersteller, Baden-Württemberg’s Minister

for the Environment: “The rubble which we have now cleared

for landfill is harmless to health.”

No technically unfounded, regulatory

tightening for phase-out operation

It would be desirable and important for us to continue this

broad consensus on other issues too, such as the smooth

return of waste from reprocessing in France and the United

Kingdom, and above all on politically trouble-free phaseout

operation of nuclear power plants.

We welcome the federal government’s intention to

implement the judgment of the Federal German Constitutional

Court of 6.12.2016 with the 16 th Atomic Energy

Amendment Act and thus to pacify long-standing disputes

and protect employees affected by the structural change.

The decision to convert the electricity quantities agreed in

2000/2001 into electricity within the remaining operating

time will have a positive effect not only on the federal

budget but also on CO 2 -emissions.

The current draft bill still needs some clarification to

avoid further disputes. It has to be made clear that there

must be not only a “fair” but a complete settlement in money,

that is for all non-convertable quantities within the

group and not just for the quantities that are not transferred.

The values of the electricity quantities existing at

the time of the compensation-triggering event must be taken

as a basis and the appropriate interest paid. Payments

already made for a transfer must be taken into account.

Only in this way can constitutionality be restored. Any limit

imposed on compensation claims must not result in the

power generation deficit determined by the German Federal

Constitutional Court not being fully compensated.

Isolated positive signals in the coalition

agreement – the implementation is what

is important

Overall, it will be exciting to see how things progress with

nuclear power in particular and nuclear technology in

general in Germany. Karsten Möring, Member of the German

Bundestag, rapporteur for nuclear energy in the CDU/CSU

parliamentary group, will give us an insight into the future

of both immediately following this speech. However, if we

try to gain an impression ourselves based on the current

coalition agreement, then it is basically positive.

The following was agreed in the wording: “We stand for

speedy implementation in the search for a final repository for

highly active waste in accordance with the Repository Site

Selection Act. We are adhering to the statutory goal of

­establishing the site for a final repository by 2031.” It remains

true that this milestone is already ambitious. According to

the relevant experts, planning approval and construction

are clearly likely to continue into the second half of the

century. It remains to be seen how the implementation will

be carried out, especially against the backdrop of extensive

public participation and the legal remedies available.

It was also agreed to develop “...a concept for the

perspective preservation of specialist knowledge and staff for

operation, dismantling and safety issues in nuclear facilities as

well as for interim and final disposal.” It was stated absolutely

correctly: “Anyone who wants to have a say in safety issues

must also be able to do so. The preservation of expertise is indispensable

for this.” For this reason too, the topic of preserving

expertise is top of the agenda in the work of the DAtF.

Sites in Gronau and Lingen must be

maintained – Nuclear competence can only be

preserved by means of further development

To preserve expertise, it is necessary to further develop

and apply nuclear skills and knowledge and therefore to

continue to operate production facilities. Only in this way

can Germany continue to have a say and also make

decisions at international level. Fortunately, this realization

is also gaining acceptance among political decisionmakers.

Armin Laschet, Minister President of North-Rhine

Westphalia, was completely right when he asserted in his

speech at the state parliament on 1 March 2018: “If we close

Gronau, if we close Lingen, then this will mean that Germany

is saying goodbye to this area of production. We will then no

longer be a member of the International Atomic Energy

Agency. [...] Gronau will therefore remain, […].” In his

speech, Minister President Laschet meant the loss of a

permanent seat on the IAEA Board of Governors.

Drawing up a master plan for the further

development of nuclear competence

In view of the need to further develop nuclear competence

and skills in order to tackle the upcoming tasks in Germany

and to preserve the ability to have a say internationally, it is

appropriate to ask here when the federal government will

come up with a master plan for the further development of

nuclear competence? A master plan that will allow Germany

to adequately assess international development, whether in

operation, regulation or research, in 10, 20 or even 30 years’

time. And what will this master plan look like?

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Many countries continue to rely on

nuclear power as part of a clean, low-carbon,

sustainable electricity mix

The worldwide development of nuclear energy emphasizes

the topicality and importance of this question. With the

start of construction on the first block of the Akkuyu NPP in

Turkey on 3 April this year, which will be followed by three

more blocks, we see the implementation of a project whose

origins date back to the 1970s. EDF’s construction of the

new Hinkley Point C in the United Kingdom is being

boosted further by the 3,000 strong workforce working on

it daily. The first block constructed by South Korea in

Barakah in the United Arab Emirates is scheduled to go

online this year. And with the run-up of Ohi-3 on 14 March

and Genkai-3 on 23 March 2018, Japan is still on its way

back to becoming a leading nuclear nation. These are just

a few examples that refute the frequently implied decline

of nuclear energy. With initial criticality and first feed-in to

the grid of the fifth nuclear power plant block at the

Chinese site in Yangjiang, there are currently 451 nuclear

power plants in operation worldwide – more than ever

before in the over 60-year history of nuclear energy. At

421 GW gross and almost 400 GW net, 399.8 GW to be

exact, the output of the plants also hits record figures.

The other 57 nuclear power plants currently under

construction worldwide clearly show that many countries

continue to choose nuclear power as part of a clean,

low-carbon, sustainable electricity mix of the future. A

joint report by the International Energy Agency, the

International Renewable Agency, UNO organizations and

the World Bank of 2 May 2018 predicts an increase in the

share of nuclear power in global power generation to a

total of 15 percent, from 10 percent at present. German

companies and Germany-based companies, with their

acknowledged expertise, particularly in nuclear safety

issues, have great potential to participate in this development.

However, this requires reliable and ideology-free

support for export activities. This would also secure the

urgently needed development of competence in the

nuclear technology field in the long term.

Nuclear energy research continues

to be promoted internationally

An important milestone in the ITER project was reached in

November of last year when 50 percent of the total output

on the way to the first plasma was generated. Despite some

scepticism about the project, including among our own

ranks, we are hopeful about the outcome of this exemplary

international collaboration.

Development in the field of SMRs also remains exciting.

The announcement by Rolls Royce of bringing electricity

costs to the level of offshore wind power and the creation

of a new technical working group at IAEA dedicated to

SMRs give a new boost to the development of compact

small reactors as does the transport, which commenced on

28 April 2018, of the world’s first floating nuclear power

plant from St. Petersburg to its site of operation.

Nuclear technology is more than

nuclear power

Against this background, we do not intend to challenge

Germany’s phasing out of nuclear power. The phase out,

however, does not mean that Germany should be allowed

to become a nuclear technology-free zone. Nuclear

technology, as we know, is more than just power generation.

This is why, as the DAtF, we are increasingly devoting

ourselves to other topics. We must oppose the efforts of

some political forces to phase out the use of nuclear

technology in other areas such as medicine, agriculture

and industry.

This means that not only top-level research, such as the

Garching research reactor FRM II, which holds the world’s

best ratio of thermal output to neutron flux and is thus one

of the most effective and modern neutron sources in the

world, but also everyday applications must also be maintained.

We need to raise public awareness of the fact that

nuclear technology associated with

• medical applications such as X-rays, computed

tomography, radiation treatment and diagnostic

applications,

• the killing of germs in the food industry and in

medicine,

• the treatment of seeds and the development of new

plant species,

• the non-destructive testing of materials and joints in

the aviation and automotive industries

is part of our everyday life. The aim is to raise public

awareness of the fundamental importance of maintaining

nuclear research and applications in Germany.

Successful AMNT as the annual meeting of the

whole industry

Ladies and Gentlemen,

With your commitment, you all make an important

con tribution to the development of nuclear competence,

not only in Germany but worldwide. It is you who actually

make the AMNT, as an international platform for

knowledge and dialogue, possible and who fill it with life;

who plan and are responsible for the programme, give

presentations and enrich our Annual Meeting with your

participation. For this you have my heartfelt thanks.

I would also like to sincerely thank our many partners

who are providing an exceptional display at the completely

sold-out industry exhibition. I am also pleased to welcome

our British partners. Our intensive discussions as part of

the last “Energy in Dialogue” event as well as the personal

exchange between the DAtF and British government

representatives on the topic of Brexit are currently arousing

justified hope that this collaboration will continue largely

undisturbed. The fact that, in addition to nuclear communities

from the United Kingdom and the Czech Republic,

we are also welcoming many well-known but new exhibitors

from home and abroad strengthens this hope. The

breaks are a good opportunity for obtaining information

and exchanging ideas and opinions.

The DAtF reception, to which you are all cordially

invited, will take place immediately after the second part

of the plenary session. Following this, we can look forward

to the traditional social evening which our exhibitors and

sponsors warmly invite you to attend.

Ladies and Gentlemen,

I wish everybody a successful meeting with fruitful and

interesting discussions and exceptional insights.

Author

Dr. Ralf Güldner

President of the DAtF

(German Atomic Forum)

Robert-Koch-Platz 4

10115 Berlin, Germany

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TESPA-ROD Code Prediction of the Fuel

Rod Behaviour During Long-term Storage

Heinz G. Sonnenburg

| | AMNT 2018: Best Paper Award, awarded by Dr. Erwin Fischer (left) to

Dr. Heinz G. Sonnenburg (right).

| | Fig. 1.

Crystallographic length change of UO 2 /PuO 2 -fuel relative to displacement per atom (dpa) [RAY 15].

The paper “TESPA-

ROD Code Prediction

of the Fuel Rod

Behaviour During

Long-term Storage” by

Heinz G. Sonnenburg

has been awarded as

Best Paper of the

49 th Annual Meeting

on Nuclear Technology

(AMNT 2018), Berlin,

29 and 30 May 2018.

Introduction The TESPA-ROD code is applicable to both LOCA and RIA transients. Recently, the code’s models

have been extended in order to predict the transitional fuel rod behaviour during long-term storage [SON 17].

Due to permanent α-decay of actinides

in the fuel during long-term storage,

both fuel swelling and helium release

continue and generate an impact on

the fuel rod behaviour. Therefore, the

TESPA-ROD code extension requires

particular modelling of fuel swelling

and modelling of the associated

helium gas release. These processes

and their modelling have significant

impact on the prediction of cladding’s

stress level.

Continued fuel swelling reduces

the gap between fuel and cladding

which reduces the fuel rod fission

gas volume and might increase

the fuel rod inner pressure by that.

Simultaneously, the release of

helium tends to keep the rod

internal pressure high, thus the

gap between fuel and cladding could

be enlarged. If fuel swelling is the

dominating process, as in case of

MOX fuel, even gap closure might

occur which leads to pellet- cladding

interaction which finally enhances

significantly the stress level in the

cladding. A priori, which effect

dominates cannot be estimated with

simple engineering judgment. Therefore,

a code prediction is inevitable

in order to get reliable estimates about

dominating processes.

Fuel swelling

Fuel in a fuel rod accumulates fission

gases in the fuel matrix during normal

operation. E.g., small gas bubbles of

micrometer size appear within the

fuel grain at higher burn-up levels.

Because the accumulation of fission

gas in the fuel matrix is limited, some

quantity of fission gas will get released

from fuel.

There is a well-known interlinkage

between the accumulation of fission

gas and swelling of the pellet. The

more the fuel accumulates fission gas,

the more the fuel swells.

The same mechanism is true for

the long-term storage, but here helium

is accumulated instead of fission

gases. This helium stems from the

decay of α-emitting actinides.

Patrick Raynaud [RAY 15] from

US.NRC has compiled fuel swelling

correlations for UO 2 fuel and PuO 2

fuel which refer to the α-decay in the

fuel (Figure 1). Correlating parameter

is dpa (displacement per atom). This

compilation reveals a swelling mechanism

which saturates at a certain

maximal swelling level. Consequently,

the swelling can be expressed as

exponential function:

upper bounding values (1)

and

mean values (2)

where ∆a is the change of lattice

parameter, a 0 is the undeformed

lattice parameter.

The parameter dpa correlates with

time. Raynaud /RAY 15/ provides for

60 GWd/t UO 2 fuel the relation

dpa(t) =0.01172 t 0.72246 , where t is

measured in years. In case of MOX

fuel, this relation can be multiplied by

3, because MOX fuel has 3-times more

α-decays, see figure 5.3 on page 54 in

[SON 17].

The swelling mechanism, as correlated

above, refers mainly to the production

of Frenkel pairs and helium

atoms at interstitial positions in the

crystal structure of UO 2 . The effect

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| | Fig. 2.

Evolution of fuel rod temperature (all curves FGR=10 %…40 % overlap).

| | Fig. 3.

Evolution of internal gas pressure.

| | Fig. 4.

Evolution of gap size.

| | Fig. 5.

Evolution of hoop stress.

of helium bubble formation due to

accumulation of α-particles in bubbles

is presumably underrepresented in

the experimental investigations mentioned

in figure 1. According to Wiss

et al. [WIS 14] this accumulation

effect would dominate in case of very

high alpha doses. Thus fuel swelling

saturation could also be observed

at 0.6 % instead of 0.4642 % as

mentioned in Figure 1. Therefore, the

characterization “upper bounding

values” in Figure 1 has to be taken

with caution.

Helium gas release

The characteristic saturation of the

fuel length change ∆a/a 0 is an

indication that the accumulation of

helium within the fuel matrix is

limited. Therefore, the shape of the

curve above is taken in order to

quantify the fraction of produced

helium which gets released from fuel

matrix. The more the saturation in

Figure 1 is reached the larger the

fraction of produced helium atoms

is. E.g., this interlinkage can be

expressed with:

(3)


where HeF released (t) is considered as fraction of produced helium mol rate at

time t that gets released from fuel matrix. Equation (3) refers to the upper

bounding swelling correlation given with equation (1). Consequently, the

helium mol rate produced and retained in fuel matrix is:

(4)

Whereas the helium mol rate released from fuel matrix is:

(5)

The helium production rate from α-decay Ḣe produced (t) within the UO 2 fuel

matrix can be approximated with:

(6)

Therefore, the TESPA-ROD modelling

approach for helium release follows

the concept of an athermal helium

release, because a) annealing tests

reveal a minimum temperature of

800 K for the start of thermal helium

release [WIS 14] and b) high burn-up

structure with fuel grain size below

1 µm offers a huge intergranular

surface thus α-particles can easily

escape from these grains without

thermal controlled diffusion.

TESPA-ROD predictions

for long-term storage

The TESPA-ROD prediction for longterm

storage significantly depends on

the normal operating condition of the

fuel rod just before reactor shut down.

If the fuel rod burn-up is e.g. at

70 GWd/t and the fission gas release

(FGR) due to normal operation is low

(e.g. 10 %) due to not demanding

normal operation, the pellet-cladding

gap would be closed and simultaneously

the hoop stress in cladding

would be rather low, which is a consequence

of cladding irradiation creep

during normal operation (stretch-out

operation).

Under these conditions (70 GWd/t,

10 % FGR, low cladding stress and

gap closure at normal operation), the

reactor shut down would consequently

lead to an opening of the gap

between fuel and cladding. This gap

AMNT 2018

TESPA-ROD Code Prediction of the Fuel Rod Behaviour During Long-term Storage ı Heinz G. Sonnenburg


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AMNT 2018

376

| | Fig. 6.

Comparison of internal gas pressure (upper vs mean).

| | Fig. 7.

Comparison of gap size between pellet and cladding (upper vs mean).

| | Fig. 8.

Comparison of cladding hoop stress (upper vs mean).

| | Fig. 9.

Comparison of helium gas release (upper vs mean).

| | Fig. 10.

Comparison of pellet outer diameter (upper vs mean).

opening would be in the range of

20 µm.

The question is whether the gap

gets smaller due to pellet swelling or

will the gap get larger due to the

internal gas pressure affected by

helium gas release.

Of course, various mechanisms

have an impact on the gap size during

long-term storage. The decay heat

power drops and the fuel rod cools

down, thus the internal pressure

decreases and both cladding and pellet

contract due to this cooldown. On the

basis of these obvious mechanisms, it

is easy to estimate the evolution of the

cladding hoop stress because the stress

level depends on the dropping internal

gas pressure only.

The additional consideration of the

mechanisms (α-decay related fuel

swelling and helium gas release) leads

to a more complex situation. Especially,

if the gap gets closed during the

long-term storage, the evolution of

the hoop stress depends no longer on

the internal gas pressure.

Figures 2 through 5 illustrate

the TESPA-ROD prediction for the

evolution of temperature, pressure,

gap size and hoop stress for the periods

of a) reactor shut down, b) wet- storage

(4 years), c) drying procedure and

d) subsequent dry- storage over several

decades. The prediction makes use

of the upper bounding values of the

swelling correlation as defined by

Raynaud [RAY 15].

In particular, the result for cases

with fission gas release (FGR) below

25 %, gap closure (gap closure = gap

size at 3 µm, which corresponds to

roughness of clad and pellet) occurs a

few years after beginning of wetstorage.

For FGR at 40 %, no gap

closure occurs and the hoop stress

depends on the internal gas pressure

only.

Due to helium release from fuel,

Figure 3 reveals an internal gas pressure

minimum during the period of

dry-storage. The occurrence of

minimum depends on FGR. E.g. at

10 % FGR the minimum occurs at

about 60 years after start of drystorage.

These results require further investigations.

Therefore, the TESPA-ROD

prediction using the pellet swelling

model for both the upper bounding

value and the mean value according

to Raynaud [RAY 15] has been compared.

As before, the pellet burn-up of

70 GWd/t is at focus. The initial hoop

stress level during normal operation

(no stretch-out operation) is adjusted

to 90 MPa which leads to a gap

opening after reactor shut down to

roughly 15 µm.

The evolution of pressure, gap

size, hoop stress, helium gas release

and pellet outer diameter for both

model predictions (upper vs mean) is

shown in Figures 6 through 10.

In these figures, fission gas release

is set to 10 %. The temperature

evolution is the same as shown in

Figure 2 above.

Figure 7 shows an early gap

closure for the “upper” case and a late

gap closure for the “mean” case after

about 40 years. Consequently, the

stress level increases at start of

dry-storage for the “upper” case while

this increase, although less pronounced,

occurs after 40 years for

the “mean” case.

The significant increase of hoop

stress after gap closure is a consequence

of a shrinking cladding when

it cools down and when its shrinkage

gets blocked by the pellet. The pellet

outer diameter evolution is shown

in figure 10. This figure reveals

an almost stagnant outer diameter

evolution for the “mean” case. That is,

AMNT 2018

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pellet shrinkage due to cooldown

is roughly compensated by pellet

swelling.

Conclusion

A comparative study with TESPA-ROD

for long-term storage transient under

boundary conditions described above

illustrates that both pellet swelling

models as defined by Raynaud (“upper”

case and “mean” case) show up

with gap closure during wet-storage.

There is a significant delay in gap

closure by 40 years when the pellet

swelling model according Raynaud

relies on “mean” values. Furthermore,

the TESPA-ROD predictions reveal

that the impact of pellet swelling on

cladding’s hoop stress is dominating

above the impact of helium gas release.

Nevertheless, both effects need

to be taken into account, because they

affect the cladding’s hoop stresses.

And cladding hoop stress is an important

parameter for subsequent safety

evaluation of the long-term storage

fuel rod behaviour.

Literature

[SON 17] Sonnenburg, H.G.; Boldt, F.: Brennstabverhalten

im Normalbetrieb,

bei Störfällen und bei Langzeitlagerung.

GRS-Bericht: GRS – 464,

ISBN 978-3-946607-47-2, August

2017.

[RAY 15] Raynaud, P.; Einziger, R.: Cladding

stress during extended storage of

high burnup spent nuclear fuel.

Journal of Nuclear Materials 464,

pp. 304–312, 2015.

[WIS 14] Wiss, T.; et al.: Evolution of spent

nuclear fuel in dry storage conditions

for millennia and beyond.

Journal of Nuclear Materials 451,

pp. 198–206, 2014.

Author

Dr. Heinz G. Sonnenburg

Gesellschaft für Anlagen- und

Reaktorsicherheit (GRS) gGmbH

Boltzmannstr. 14

85748 Garching, Germany

DATF EDITORIAL NOTES

377

Notes

Global Outlook on Nuclear Energy

From a global perspective, nuclear energy is a requested energy

source. It covers about 11 % of the worldwide demand for electricity,

ensured by 451 nuclear reactors with a net capacity of 395 GW e .

Currently, there are 58 nuclear reactors with a net capacity of

nearly 60 GW e under construction. In addition, more than

150 reactors in about 25 countries are in the planning phase.

Nuclear power plants under construction

According to forecasts, the demand for nuclear energy will rise,

especially in Asia and Central Europe as well as Eastern Europe.

In its study “Sustainable Development Scenario”, the International

Energy Agency projects a share of nuclear energy of 15 % in electricity

generation worldwide until the year 2040.

For further details

please contact:

Nicolas Wendler

DAtF

Robert-Koch-Platz 4

10115 Berlin

Germany

E-mail: presse@

kernenergie.de

www.kernenergie.de

Russia: 6

USA: 2

Brazil: 1

Finland: 1 Belarus: 2

China: 17

Slovakia: 2 Ukraine: 2

France: 1

Pakistan: 2

South Korea: 4

Turkey: 1

Japan: 2

Taiwan, China: 2

United Arab Emirates: 4

Bangladesh: 1

India: 7

Argentina: 1

Source: Weltenergierat Deutschland e. V.: Energie für Deutschland; IAEA; Date: 15.06.18

DAtF Notes


atw Vol. 63 (2018) | Issue 6/7 ı June/July

378

Neubauprojekte im Ausland – eine Herausforderung

für die Regulierung

Christian Raetzke

SPOTLIGHT ON NUCLEAR LAW

Weltweit sind gegenwärtig 57 Reaktoren in 16 Ländern in Bau; 154 Reaktoren befinden sich in 24 Ländern in – mehr

oder weniger – konkreter Planung (Zahlen nach World Nuclear Association, www.world-nuclear.org). Ähnlich große

Zahlen gab es ja schon einmal vor einigen Jahrzehnten, als das Gros der heute laufenden Anlagen errichtet wurde. Die

Lage ist trotzdem heute anders, auch in juristischer und genehmigungstechnischer Hinsicht. Aufgrund der Vielfalt

der für die Projekte in den verschiedenen Ländern verwendeten strukturellen Ansätze ist es zunehmend schwerer,

bewährte Modelle von Regelsetzung, Genehmigung und Aufsicht (alles zusammen kann im Englischen ja mit dem

Begriff „Regulation“ bezeichnet werden) weiterzuführen.

Wie war es früher? Länder, die sich für die Kernenergie

entschieden, bauten eine Infrastruktur auf; sie entwickelten

eine eigene Technologie oder übernahmen sie

von woanders (z.B. aus den USA), entwickelten sie dann

aber oft eigenständig weiter (wie Deutschland und

Frankreich). Erst wurden Versuchsreaktoren gebaut, dann

kamen kleine kommerzielle Kernkraftwerke, dann große.

Das nationale Atomrecht wuchs mit bzw. war immer einen

Schritt voraus; dasselbe galt für die Behörden und ggf.

Gutachterorganisationen, die sich in Größe und Erfahrung

organisch entwickelten.

Für viele Neubauländer trifft dieses Modell nicht mehr

zu. China und Indien haben vielleicht einen grob vergleichbaren

Weg gewählt; in anderen Ländern gibt es aber ganz

andere Szenarien.

Eines der interessantesten und bislang erfolgreichsten

Beispiele liefern sicherlich die Vereinigten Arabischen

Emirate mit dem Barakah-Projekt. Für viel Geld wurden

Experten aus zahlreichen Ländern an den Golf geholt. Sie

haben in kurzer Zeit Bemerkenswertes geleistet; allerdings

fügten sich die unterschiedlichen Kulturen auch nicht

immer bruchlos zusammen. Der Unterschied zu einem

organisch gewachsenen Nuklearprogramm ist deutlich

erkennbar bei der noch lückenhaft vorhandenen lokalen

fachlichen Kompetenz, sowohl beim Betreiber als auch bei

den Behörden.

In rechtlicher Hinsicht wurde ein nationaler regulatorischer

Rahmen geschaffen, der sich an den Leitlinien

der Internationalen Atomenergieorganisation (IAEO) orientiert,

aber auch Elemente aus Ländern übernimmt, die den

jeweiligen Entwurfsverfassern (meist externe Berater)

vertraut waren oder mit denen eine Zusammenarbeit

besteht. Das hat, soweit man das überblicken kann, in den

Emiraten gut funktioniert. Allgemein ergibt sich jedoch in

solchen Konstellationen durchaus die Gefahr, dass IAEO-

Texte mehr oder weniger mit „copy-and-paste“ übernommen

und mit Versatzstücken aus den Gesetzen und

Regelwerken anderer Länder ergänzt werden, ohne dass

die Adressaten notwendigerweise die Texte vollständig

durchdringen; auch besteht die Gefahr von Wider sprüchen

und Redundanzen und einer mangelnden Ankopplung an

das vorhandene nationale Recht. Und was ist, wenn die

(juristischen) Berater wieder abziehen? Gibt es dann

eine fundierte Kompetenz in Ministerien und Behörden,

mit den Atomrechtsinstrumenten umzugehen? Die

Ertüchtigung lokaler Kräfte ist auch hier ein langwieriger,

aber unumgänglicher Prozess.

Ein ganz anderes Modell kann am Beispiel des Projekts

Akkuyu in der Türkei angesprochen werden. Hier bauen

die Russen nicht nur das Kraftwerk, sondern sie finanzieren

es auch (abgesichert durch einen Langzeit-

Stromliefervertrag) und werden es auch betreiben. Man

spricht hier von einem BOO(build-own-operate)-Modell.

Das ist neuartig in der Kerntechnik und die Herausforderung

ist nicht zu übersehen, dass das „Gastland“

möglicherweise gar nicht so sehr daran interessiert ist,

eine regulatorische Infrastruktur „in voller Schönheit“

aufzubauen. In der Türkei stellt sich diese Frage letztlich

nicht ernsthaft, da ja weitere Kernkraftwerke mit anderen

Technologien und Betreibern geplant sind und daher ein

vollständiger regulatorischer Rahmen unentbehrlich ist.

Dessen Aufbau hat sich jedoch bislang als schleppend

erwiesen. Und gerade für Akkuyu ist auch keinesfalls

geklärt, was passiert, wenn die Genehmigungsbehörde

z.B. die Erteilung einer Genehmigung verweigert, wenn

aus ihrer Sicht die Voraussetzungen nicht vorliegen, die

russischen Vertragspartner aber auf die zwischenstaatlichen

Vereinbarungen pochen und Erfüllung verlangen.

Solche Konstellationen werden noch deutlicher werden

in Ländern, die nur ein Kernkraftwerk mit einem Lieferanten

(der gleichzeitig oft das Geld mitbringt) planen.

Eine ganz parallele Herausforderung stellt die

möglicherweise bevorstehende internationale Verbreitung

von SMR (Small Modular Reactors) dar. Die meisten

Modelle und Konzepte hierfür haben den Ansatz

gemeinsam, dass die Anlagen insgesamt oder in Modulen

in Fabriken montiert und zum Einsatzort gebracht werden;

idealerweise werden sie dort nur „hingestellt“ oder

„ zusammengebaut“ und mit dem Stromnetz verbunden

(„plug-and-play“) und womöglich samt abgebrannten

Brennelementen irgendwann wieder abgeholt und ins

Ursprungsland zurückgebracht. Ist es realistisch, vom

„Gastland“ zu erwarten, dass es eigens hierfür eine

vollständige regulatorische Infrastruktur mit eigener

Expertise aufbaut? Andererseits kann ein Reaktor in einem

Land ohne ein vollständiges Atomgesetz und eine fachlich

gut besetzte Behörde wohl kaum betrieben werden. Hier

sind Konflikte vorprogrammiert, aber auch hoffentlich

gute Lösungen zu erwarten.

Für die regulatorische Bewältigung dieser Herausforderungen

gilt wie so oft: neue, komplexe Ent wicklungen

fordern teils neue Lösungen. In der Kerntechnik aber gilt

ebenso, dass man auf Bewährtem aufbauen muss und

neue Lösungen – nicht nur technische, sondern auch

regulatorische – nicht einfach mal „ausprobiert“ werden

können, sondern der sehr sorgfältigen Rechtfertigung

bedürfen, dass sie die erforderliche Sicherheit gewährleisten.

Author

Rechtsanwalt Dr. Christian Raetzke

CONLAR Consulting on Nuclear Law and Regulation

Beethovenstr. 19

04107 Leipzig, Germany

Spotlight on Nuclear Law

New Build Projects Abroad – A Challenge for Regulation ı Christian Raetzke


atw Vol. 63 (2018) | Issue 6/7 ı June/July

Safety Assessment of the

Research Reactors FRM II and FR MZ

After the Fukushima Event

Axel Pichlmaier, Heiko Gerstenberg, Anton Kastenmüller, Christian Krokowski, Ulrich Lichnovsky,

Roland Schätzlein, Michael Schmidt, Christopher Geppert, Klaus Eberhardt and Sergei Karpuk

After the events at the Fukushima-I nuclear power plant (NPP) in 2011 the Reaktorsicherheitskommission (RSK) has

carried out an overall assessment of the German nuclear fleet with respect to extreme (beyond design base) events. The

RSK is an expert group of operators, technical support organizations (TSO) and scientists that consults the German

Federal Ministry of the Environment (BMUB) in questions concerning reactor safety. This paper deals only with the

research reactors (RR) FRM II (Garching) and FR MZ (Mainz). The findings of the RSK, its recommendations and their

status of implementation will be presented.

1 Introduction

Upon request of the German Federal

Government the FRM II, the FR MZ and

the research reactor BER II at the

Helmholtz Zentrum Berlin, like every

other nuclear facility exceeding 50 kW

thermal power, underwent a so-called

stress test by the Reactor Safety

Commission (RSK – Reaktor-Sicherheitskommission).

Special emphasis

was put on seismic events, flooding

and other natural events, superposition

of such events and manmade

hazards like aircraft crashes. Additionally,

independent event sequences

relevant for research reactors have

been postulated and analysed, even

under aggravated conditions. Following

these analyses the RSK has deduced

recommendations with respect

to the robustness of these facilities

under such circumstances. The RSK

findings summarized in this paper are

based on [1] and [2].

The following main aspects have

been evaluated in detail for the three

still operational German research

reactors FRM II, FR MZ and BER II:

vital safety functions of the RR and

their behaviour at seismic events,

flooding, other natural events, postulated

events (like long lasting station

blackout (SBO) with emergency

power supply requirements, complete

loss of ancillary cooling systems),

robustness of emergency preparations

for safety measures even under

aggravated conditions due to external

events; consequences of the release

of burnable or toxic gas.

This article focuses on the research

reactors FRM II and FR MZ.

The FRM II in Garching is a tank in

pool reactor with 20 MW thermal

power. A single fuel element, containing

113 fuel plates with highly

enriched Uranium, is cooled by light

| | Fig. 1.

Overall view of the FRM II (foreground), the neutron guide hall (middle) and the FRM I (”atomic egg”, now under decommissioning).

water and placed in a moderator

tank filled with heavy water. This

setup yields an unperturbed thermal

equivalent flux of 8 × 10 14 n/cm 2 /s

over a cycle of 60 days. Generally, the

reactor is operated for up to four

cycles per year. The FRM II has

reached criticality for the first time on

March 2 nd , 2004. It is, therefore, the

most modern research reactor in

Germany.

The main purpose of the FRM II

is scientific research in beam tube

experiments. However, it is also used

for radioisotope production; it operates

a silicon doping facility and an

installation for medical treatment.

Details can be found e. g. in [3]. A

sketch of the overall FRM II design is

given in Figure 1.

The Forschungsreaktor TRIGA

Mainz (FR MZ) at the Johannes

Gutenberg University is a classical

light water cooled swimming pool

reactor with up to 100 kW thermal

power in steady state operation mode.

Furthermore, the FR MZ provides

neutron pulses with energies of typically

10 MJ and pulse lengths of

30 ms. Currently the core is equipped

with 76 fuel elements applying low

enriched uranium in a zirconiumhydrid

matrix. This fuel configuration

has a negative temperature coefficient

which provides an inherent safety

mechanism: the moderation of neutrons

is automatically suppressed with

increasing temperature on a chemicalphysical

basis. Thus the chain reaction

is stopped, as soon as the fuel reaches

temperature of about 150 °C or above.

This mechanism neither requires the

availability of personnel nor any

infrastructure as electricity or further

control systems and works faster than

any engineered device. The FR MZ is

one of the most intensively utilized

TRIGA reactors worldwide with in

379

RESEARCH AND INNOVATION

Research and Innovation

Safety Assessment of the Research Reactors FRM II and FR MZ After the Fukushima Event

ı Axel Pichlmaier, Heiko Gerstenberg, Anton Kastenmüller, Christian Krokowski, Ulrich Lichnovsky, Roland Schätzlein, Michael Schmidt, Christopher Geppert, Klaus Eberhardt and Sergei Karpuk


atw Vol. 63 (2018) | Issue 6/7 ı June/July

RESEARCH AND INNOVATION 380

average 200 days of operation per

annum. It has reached its first

criticality on 3 rd of August 1965. The

FR MZ has three main pillars of

utilization: the most operation time is

consumed for fundamental research

such as nuclear spectroscopy and

physics with ultra-cold neutrons.

Besides, the reactor is deployed for

applied sciences, e.g. for neutron

activation analysis or radioactive

tracer production. Apart from this

pure scientific utilization, the FR MZ

contributes to the education of

students and maintenance of nuclear

competence by various lectures, lab

courses and one-week lasting rector

operation courses.

2 Procedure of the

evaluation

Objective was to evaluate whether the

fundamental safety requirements

• to control reactivity

• to cool the fuel assemblies and

• to limit the release of radioactive

material

could still be met under more difficult

conditions than taken into account

during the licensing process. For

example, larger scale external

destruction has been assumed and

additional event sequences were

postulated, most prominently the

non-availability of the electric grid to

supply safety-relevant installations.

The conditions investigated were:

• Seismic events

• Flooding

• Other naturally occurring adverse

conditions

• Postulated events like long lasting

(> 2 h) station blackout

• Robustness of preventive

measures

• Airplane crash

• Release of gas (explosion, other

effects of burnable gas, toxic gas)

• Terrorist attacks

The established site specific emergency

measures, even under extreme

conditions like core melt-down, have

been re-evaluated in view of large

scale destruction of the relevant

infrastructure also in the surroundings

of the affected RR.

Because these evaluations are

based on requirements for NPP, a

graded approach has been taken

bearing in mind that the risk associated

with a RR is much lower than

that of a NPP. This is due to the fact

that radioactive inventory of a RR is

typically several orders of magnitude

smaller than that of a NPP and the

processes involved are in general

more benign.

The RR operators followed two

different systematic approaches of

objective evidence. In most cases,

distinct expertise has been produced

to the individual conditions listed

above. For the FR MZ several scenarios

have been covered by a scenario,

which is assumed as all-embracing

event. This case is described by an

airplane crash with maximal damage

to the reactor as described in section

3.2.8.

Following the analysis in [3]

several recommendations have been

made by the RSK.

3 Recommendations

3.1 Generic recommendations

for RR

The RSK has recommended generic

measures for the German RR:

• Every RR should work out a site

specific emergency concept for

internal preventive and mitigating

emergency measures based on the

risk associated with the respective

RR. This concept should be based

upon recommendations given for

NPP in [4].

• Adverse environmental conditions

should be taken into account when

planning such measures.

• Methods to cope with beyond

design base LOCA-type accidents

should be considered in the emergency

planning.

• For beyond design base scenarios

when standard instrumentation to

monitor reactor and radiation

parameters might fail, sufficient

backup is to be made available.

• In the event of a core melt-down a

concept to minimize the release of

radioactivity should be available.

3.2 Specific evaluation

of the FRM II

3.2.1 Specific evaluation

Immediately following the events at

the Fukushima NPP the evaluation of

the FRM II by the RSK, based on

information provided by the licensee

and other available information, gave

the following results [3]:

3.2.2 Seismic events

Cornerstone of the FRM II safety

concept is the integrity of the reactor

pool and related structures. The

design requirement for the FRM II is

robustness against an earthquake of

magnitude VI ½ (MSK). Although

strong hints towards the robustness

of the FRM II in general and in particular

the reactor pool even against a

magnitude VIII quake existed, no conclusive

prove could be provided by the

licensee in 2012. The RSK therefore

concluded that further investigations

should be carried out and be evaluated

by the TSO.

The FRM II immediately started

calculations on the robustness of the

reactor pool and related structures.

The calculations were double-checked

by the TSO. Overall it could be proven

that even a seismic event of the

magnitude VIII ½ would not damage

the integrity of the reactor pool and

consequently no water (for cooling/

shielding) would be lost. This was

completed already before the 2017

RSK assessment.

3.2.3 Flooding

The FRM II is designed to withstand

a flood that is to occur statistically

every 10, 000 years. Even more severe

flooding, however, would not do any

damage that might endanger the vital

safety functions of the FRM II. Therefore

the RSK gave the FRM II the best

grade (“level 3”) regarding flooding

and did not request any further measures.

3.2.4 Other naturally occurring

adverse conditions

No such conditions could be identified

that would require further action.

3.2.5 Postulated events

The only relevant event is the station

black out (SBO). Because of the

diesel/battery buffering the safety

functions in case of SBO are guaranteed

for at least two hours. Additionally,

in the framework of the

licensing process it could be shown

that even a total loss of all active core

cooling components would not lead to

fuel damage. According to the RSK the

required criteria are met, no further

improvement is necessary.

3.2.6 Robustness of preventive

measures

The robustness of a suite of preventive

measures has been analysed by the

RSK:

• Measures against fire: the RSK

concludes that fire cannot endanger

the vital safety functions of

the FRM II.

• Measures against blocked cooling

channels (beyond design base):

these are mainly based on passive

measures like several grids to

stop migration of small particles

in the primary cooling loop. Even

a failure of these preventive

measures would not lead to

Research and Innovation

Safety Assessment of the Research Reactors FRM II and FR MZ After the Fukushima Event

ı Axel Pichlmaier, Heiko Gerstenberg, Anton Kastenmüller, Christian Krokowski, Ulrich Lichnovsky, Roland Schätzlein, Michael Schmidt, Christopher Geppert, Klaus Eberhardt and Sergei Karpuk


atw Vol. 63 (2018) | Issue 6/7 ı June/July

radiologically required evacuation

of the general public in the surroundings

of the FRM II.

• Measures against loss of the integrity

of the reactor pool leading to

loss of pool water: the concept of –

at least – double barriers has been

used throughout. Additionally,

heavy lifts in the vicinity of the

pool or delicate installations like

the cold source with its D 2 contents

are only allowed after additional

measures are in place (e. g. the

reactor is shut down and the D 2

removed).

• Internal flooding: water is drained

in such a way that safety relevant

functions cannot be affected. The

RSK considers the required criteria

as more than met.

• Measures against improper reactivity

changes: the overall reactivity

coefficients of the FRM II are

negative with increase in temperature.

A postulated release of 3 $

reactivity has been investigated in

the process of the FRM II licensing.

No need for additional measures

could be deduced.

3.2.7 Aggravated boundary

conditions

Several emergency measures (draining

of the D 2 O moderator, sealing

of the reactor building ventilation

systems against the environment,

measures to maintain the pool- waterlevel

and emergency fuel unloading,

installation of a backup 400 V electric

power supply) are described in the

FRM II operating manual (BHB).

There is an emergency control room

and sufficient room for emergency

first responders. The functioning of

communication lines under such

conditions could not be verified by

the RSK. The existing instrumentation

is robust against seismic events and

airplane crashes. Some measures,

however, require access to the reactor

hall. The RSK recommends implementing

measures that do not require

such access since it might no longer be

possible under some circumstances.

Additional emergency drills and the

availability of the required personnel

in case of such events should be

verified.

3.2.8 Airplane crash

No additional measures are required

to withstand the impact of a military

or a large civilian aircraft.

3.2.9 Release of gas

The effects of explosions are covered

by the robustness of the FRM II

towards seismic events and the crash

of even a large aircraft.

In the vicinity of the FRM II no

significant supply of burnable gas

exists, therefore no additional measures

are required (but could be

handled regardless by the design of

the FRM II site).

Toxic gas might affect the availability

of personnel but not compromise

the vital safety functions of the

FRM II.

3.3 Specific evaluation

of the FR MZ

3.3.1 Seismic events

For the FR MZ no detailed seismic

studies, investigating the influence of

the integrity of the reactor under

various earthquakes levels, have been

initiated. Instead, the damage impact

and radioactive release initiated by

the airplane crash scenario (described

in section 3.3.4) covers the consequences

triggered by an earthquake.

The RSK followed the evaluation of

the local authorities (Ministerium für

Umwelt, Energie, Ernährung und

Forsten Rheinland-Pfalz MUEEF) of the

FR MZ and accepted the robustness

level 2 as fulfilled.

3.3.2 Flooding and other

naturally occurring adverse

conditions

Several assessments for rain and

snow, based on long-term weather

data provided by the meteorological

institute of the Johannes Gutenberg

University in Mainz, have been taken

into account. In addition, static investigations

against the highest storm

level experienced in the last decades

in the region have been conducted.

From that no unexpected naturally

occurring conditions could be identified

that would require further action.

The MUEEF accepts for these scenarios

a robustness level of the highest grade

(level 3).

3.3.3 Release of gas and

protection against

explosions

The effects of gas-induced explosions

are covered by the scenario of an

airplane crash. A formerly existing

nearby underground gas pipe is

shutdown, so that within 350 m radius

of the reactor hall, no significant

supply of burnable gas exists. The

handling of burnable gas for shortterm

scientific purposes is strictly

regulated at the FR MZ. No burnable

or flammable gas bottles are allowed

to be stored in the reactor hall. Neither

burnable gas nor toxic gas compromise

the inherent safety mechanism

of the FR MZ. As a result, no additional

measures other than the standard

safety provisions for workers in the reactor

hall are required.

The emergency diesel generator of

the FR MZ has an attached 600 L diesel

reservoir. The reactor building is

shielded against the generator container

by the larger cooling tower

installation.

As a result, the RSK assesses this

scenario with a degree of protection

level 3 and has no further demand for

continuative investigations.

3.3.4 Airplane crash

The impact and consequences of an

airplane crash on the FR MZ is the allembracing

damage scenario for the

FR MZ. In the scenario, an airplane

crashes on the reactor hall and

destroys the roof of the hall. In a first

phase debris of the plane destroys the

reactor vessel in such way, that all

water is instantaneously and completely

leaking out of the reactor pool

and furthermore decapitating every

single fuel element. Afterwards the

reactor vessel is sealed in such a way,

that airplane fuel can accumulate

under the now dry reactor core and

start to burn, heating the fuel elements

to a temperature of 1,100 °C. Calculations

of the TSO (TÜV Rheinland)

study the emission of radioactive

isotopes through the open reactor roof

including the scenario with and without

rain. Taking into account the

emergency reference level of the

German Commission of Radiological

Protection (Strahlenschutzkomission),

the radiological levels reached remain

below 30 % of the permissible values.

Therefore, except for the typical ban

area inherent to a conventional airplane

crash, no further immediate

measures have to be initiated.

The MUEEF sees the degree of

protection level 2, the highest possible

scale for this scenario, as fulfilled. The

RSK confirms this assessment.

3.3.5 Robustness of preventive

measures

The FR MZ has four radial beam tubes,

which provide access for experiments,

which are placed close to the reactor

core. The airplane crash scenario

covers a sudden loss of beam tube integrity

and subsequent loss of water

inside the core. Furthermore the influence

of an explosion inside the beam

tubes, e. g. through the D 2 and H 2 gas

applied for ultra-cold neutron experiments

have been evaluated by an

RESEARCH AND INNOVATION 381

Research and Innovation

Safety Assessment of the Research Reactors FRM II and FR MZ After the Fukushima Event

ı Axel Pichlmaier, Heiko Gerstenberg, Anton Kastenmüller, Christian Krokowski, Ulrich Lichnovsky, Roland Schätzlein, Michael Schmidt, Christopher Geppert, Klaus Eberhardt and Sergei Karpuk


atw Vol. 63 (2018) | Issue 6/7 ı June/July

RESEARCH AND INNOVATION 382

external TSO. Due to a double-barrier

concept of the experiments, the TSO

verifies that even in case of D 2 or H 2

ignition, neither the beam tube

integrity is affected nor any damage to

the core is expected.

In the assessment of the MUEEF,

the FR MZ fulfils a protection level 3

and the RSK sees no further need for

robustness analysis on this subject.

In case of a SBO the FR MZ is

equipped with an emergency-power

supply which consists of a combination

of battery buffer, which drives all

necessary reactor control units and

radiation surveillance systems for at

least one hour, and an emergency

diesel generator which starts within a

few minutes after the power blackout.

With a permanent diesel reservoir of

about 600 L, the diesel generator

supplies electric power to the FR MZ

infrastructure for up to 40 hours.

4 Evaluation of the

measures

As a follow-up, the points mentioned

above have been re-evaluated by the

RSK in 2017 [2]. The conclusions can

be summarized as follows:

4.1 Evaluation of the measures

taken by FRM II until 2017

4.1.1 Emergency drills

The FRM II has significantly revised

its emergency concept and mostly

implemented the RSK recommendations.

Some recommendations have

not been addressed in full detail yet:

The RSK recommends that the FRM II

should enlarge its concept of emergency

drills. The internal emergency

organisation as a whole should train at

least once yearly, the relevant external

authorities should be included in these

exercises at least every five years.

At the time of writing, however,

the internal emergency exercise concept

is fully functional and even an

external exercise has been done.

These measures, though, have not

been evaluated by the RSK yet.

4.1.2 Emergency measures

to supply water to the

reactor pool

The RSK recommends having a system

in place to supply water to the reactor

pool in case of a failure of the relevant

barriers against loss of pool water.

While this recommendation has not

been addressed explicitly by the

FRM II yet, at FRM II already now with

existing measures or minor changes it

would be possible to supply water to

the pool in case of emergency without

access to the reactor hall. Since no

explicit evidence has been provided by

FRM II yet there is also no evaluation

of the RSK.

4.1.3 Robustness of the

emergency data

acquisition systems

The RSK recommends an analysis on

the availability of the relevant DAQ

systems in case of beyond design base

accidents, since emergency measures

require reliable information e. g. on

the pool water level, temperature,

neutron flux and radiation levels.

While such information – especially

pool level and temperature – can be

acquired easily by rather primitive

means the recommended prove has

not yet been provided by the FRM II.

4.1.4 Emergency communication

The FRM II is equipped with several

independent and diverse communication

channels (e. g. landlines and

GSM mobile phones). On top of that,

the RSK recommends the FRM II

emergency communication should

have priority over other’s communication

needs. This recommendation has

not yet been implemented. However,

the relevant communication channels

(e. g. land line telephone service)

have large reserves and therefore the

safety gain through priority might be

negligible.

4.1.5 Seismic robustness/

implementation of an

additional system to

maintain long term

undercriticality

Additional very detailed and thorough

analysis confirmed that the earlier

only assumed robustness of the

reactor building and the reactor pool

even towards magnitude VIII ½

(MSK) earth quakes. This has been

confirmed by the TSO.

Such a beyond design base event

might impede the proper functioning

of the primary (control rod) and

secondary (four out of five shut down

rods) shut down system. Therefore the

implementation of an additional

system to maintain long term undercriticality

is recommended by the RSK.

The FRM II is exploring several

options to implement such a system.

Ideas include diluting the D 2 O with

H 2 O in the moderator or adding Boron

to the primary cooling loop or the D 2 O

moderator. Calculations show that

even small amounts of such impurities

would already lead to the required

long term undercriticality. No final

design has been drawn up yet.

4.2 Measures taken by the FR

MZ resulting from the RSK

analysis

4.2.1 Emergency communication

Although the FR MZ infrastructure

contains several communication systems,

the RSK suggests, similar to

section 4.1.4 for the FRM II, the

prioritization of the mobile phones in

the public network. The request to the

telephone network provider is under

progress.

4.2.2 Emergency drills

The RSK recognizes that the emergency

management of the FR MZ is

upgraded by creating two new safetydedicated

reactor staff positions. It

furthermore appreciates the idea of

triannual exercises with external

forces and under the involvement of

the MUEEF. In addition to that, the

RSK request to implement annual

internal drills, including the complete

reactor crisis management, into the

FR MZ emergency drill concept.

Preparations for the establishment of

the triannual exercises are currently

ongoing.

4.2.3 Earthquake

Based on the all-embracing event of

an airplane crash, the RSK confirms

the MUEEF’s evaluation to robustness

level 2. Additionally the RSK suggests

describing measures how to shut

down the reactor manually following

an earthquake with a subsequent

malfunction of the control rods. This

description should be integrated in

the reactor operation regulations.

5 Conclusion

After the events in the Fukushima-I

NPP the RSK has analysed the robustness

of the German nuclear reactors in

general and also the FRM II and the

FR MZ with respect to beyond design

base accidents. Already the analysis in

2012 [3] had given a positive result

and only few recommendations to

even further improve the overall

safety of the research reactors in

Germany were presented.

In its 2017 re-analysis [1] the RSK

confirmed that most recommendations

were met by the FRM II. The

FRM II is working to answer the last

open points. For the FR MZ the RSK

confirmed the Mainz MUEEF’s assessment

of the TRIGA research reactor.

No open questions remained from the

2017 assessment of the FR MZ. Both

facilities are working on reaching full

compliance with all the RSK recommendations

in the near future.

Research and Innovation

Safety Assessment of the Research Reactors FRM II and FR MZ After the Fukushima Event

ı Axel Pichlmaier, Heiko Gerstenberg, Anton Kastenmüller, Christian Krokowski, Ulrich Lichnovsky, Roland Schätzlein, Michael Schmidt, Christopher Geppert, Klaus Eberhardt and Sergei Karpuk


atw Vol. 63 (2018) | Issue 6/7 ı June/July

References

[1] RSK-Stellungnahme, 492. Sitzung der

Reaktor-Sicherheitskommission (RSK)

am 22.03.2017.

[2] Stellungnahme der RSK „Anlagenspezifische

Sicherheitsüberprüfung

(RSK-SÜ) deutscher Forschungsreaktoren

unter Berücksichtigung der

Ereignisse in Fukushima-I (Japan)“,

Anlage 1 zum Ergebnisprotokoll der

447. Sitzung der Reaktor-Sicherheitskommission

(RSK) am 03.05.2012.

[3] FRM II description, http://www.frm2.

tum.de/en/the-neutron-source/.

[4] „Rahmenempfehlungen für die

Planungen von Notfallschutzmaßnahmen

durch die Betreiber von

Kernkraftwerken“, Empfehlungen der

Strahlenschutzkommission und der

Reaktorsicherheitskommission,

gebilligt in der 244. Sitzung der

Strahlenschutz-kommission am

03. November 2010, zum Zeitpunkt der

RSK-SÜ-FR gültig, mittlerweile abgelöst

durch [5].

[5] „Rahmenempfehlungen für die

Planung von Notfallschutzmaßnahmen

durch Betreiber von Kernkraftwerken“,

Empfehlung der Strahlenschutzkommission

und der Reaktor-

Sicherheitskommission, verabschiedet

in der 242. Sitzung der Strahlenschutzkommission

am 01./02. Juli 2010, gebilligt

in der 244. Sitzung der Strahlenschutz-kommission

am 03. November

2010, verabschiedet in der 429. Sitzung

der Reaktor-Sicherheitskommission am

14. Oktober 2010. Ergänzung verabschiedet

in der 468. Sitzung der RSK

am 04.September 2014 und in der 271.

Sitzung der SSK am 21.Oktober 2014.

Abbreviations

BMUB

DAQ

KSB

LOCA

MSK

MUEEF

NPP

OBe

RSK

RR

SBO

TSO

Authors

German Federal Ministry of the

Environment

data acquisition

Kerntechnischer Sicherheitsbeauftragter

(nuclear safety

officer)

loss of coolant accident

Medwedew-Sponheuer-Karnik-

Scale for the magnitude of

earthquakes (IXII)

Ministerium für Umwelt, Energie,

Ernährung und Forsten

Rheinland-Pfalz

Nuclear power plant

Objektsicherungsbeauftragter

(security officer)

Reaktorsicherheitskommission

(Reactor safety commission that

advises the BMUB)

Research Reactor

Station Black Out

Technical Support Organisation

Dr. Axel Pichlmaier

Fachbereichsleiter Reaktorbetrieb

FRM II

Dr. Heiko Gerstenberg

stellv. Technischer Direktor und

Fachbereichsleiter Bestrahlung und

Quellen FRM II

Dr. Anton Kastenmüller

Technischer Direktor FRM II

Dr. Christian Krokowski

Fachbereichsleiter Reaktorweiterentwicklung

FRM II

M. Sc. Ulrich Lichnovsky

Fachbereichsleiter Stilllegung und Rückbau

FRM(alt)

Dipl. Phys. Roland Schätzlein

stellv. Technischer Direktor und Fachbereichsleiter

Elektro- und Leittechnik FRM II

Dipl. Ing. Michael Schmidt

Fachbereichsleiter Reaktor überwachung FRM

II

FRM II, Forschungs-Neutronen quelle Heinz

Maier-Leibnitz

Technische Universität München

Lichtenbergstr. 1

85748 Garching, Germany

Dr. Christopher Geppert

Betriebsleitung FR MZ

Dr. Klaus Eberhardt

Betriebsleitung FR MZ

Dr. Sergei Karpuk

KSB und OBe FR MZ

FR MZ, Forschungsreaktor TRIGA Mainz

Fritz-Straßmann-Weg 2

55128 Mainz; Germany

RESEARCH AND INNOVATION 383

Decommissioning of Germany’s

First Nuclear Reactor

Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and Anton Kastenmüller

FRM started operating in 1957 as the first nuclear reactor in Germany. Reactor operation ended in 2000. Licensing

procedures for the deconstruction and dismantling of the reactor started in 1998. In 2014 the Technical University of

Munich (TUM) was granted the license to decommission the reactor.

In this article we describe our

(long) way to the license for dismantling

of the reactor and give a short

overview of the current state of the

decommissioning project.

We present the results of the (pre-)

licensing stage: disposal of spent

nuclear fuel (SNF) and preparation of

the safety report containing details

on fire protection, radiological characterization

(neutron activation and

contamination), waste management

and safety analysis.

With regard to the current state of

the project we will discuss: clearance

of material and current obstacles.

1 Introduction 1

1.1 Construction, licensing and

commissioning

In the 1950s the Bavarian government

gave impulses for the transformation

of Bavaria from an agricultural to an

industrial country. The worldwide

euphoria towards the peaceful use of

nuclear energy was shared by the

leading political parties in Germany,

left-wing as well as right-wing parties.

It took an incredibly short time –

compared to today’s time consuming

licensing processes – from the political

decision to build a nuclear research

reactor near Munich to the first

criticality of the research reactor

Forschungsreaktor München (FRM):

In 1956 the Bavarian government

decided to buy a research reactor. A

few days after the political decision,

Professor Maier-Leibnitz (Physics Depart

ment of the Technical University of

Munich) was sent to the USA with the

task of buying a nuclear reactor.

When the reactor was bought from

AMF, there was no federal legislation

regarding nuclear installations in

Germany. The construction of the

FRM started in 1956, without the

legislation that would be necessary for

1) The main contents

of the introduction

were collected from

the articles in [1].

Research and Innovation

Decommissioning of Germany’s First Nuclear Reactor ı Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and Anton Kastenmüller


atw Vol. 63 (2018) | Issue 6/7 ı June/July

RESEARCH AND INNOVATION 384

| | Fig. 1.

Reactor pool of the FRM.

reactor operation. With the approval

of the Bavarian state government the

construction of the FRM started – still

without any federal nuclear law. In

1957 the Bavarian government passed

a nuclear law of their own, before the

federal government had an appropriate

nuclear law for the Federal

Republic of Germany. The Bavarian

law was the basis for the license of

Germany’s first nuclear reactor.

In October 1957 first criticality was

achieved, thus making the FRM the

first nuclear installation in Germany.

Operation of FRM laid the foundation

for many decades of successful

and peacefull nuclear research and

industry in Germany.

1.2 Technical overview

The FRM was designed as neutron

source for science and material test

reactor.

It was a light-water-moderated

open pool reactor (see Figure 1).

From 1957 to 1960 the fuel was

20% enriched uranium (U-Al-alloy).

With a reactor power of 1 MW a maximal

neutron flux of 6.6 * 10 12 n/cm 2 /s

was achieved.

In order to further increase neutron

flux, the use of 90 % enriched

uranium (HEU) started in 1960. Six

years later, in 1966, the thermal

reactor power was increased to

2.5 MW and in 1968 to 4 MW. For

further increase of the neutron flux in

1982 the core design was completely

refurbished: the operating team

added Beryllium and Graphite reflector

elements and was able to increase

the neutron flux to 8 * 10 13 n/cm 2 /s.

In order to keep up with the scientific

demands other facilities were

added to the FRM. In 1962 a system to

irradiate samples near liquid Helium

temperature (TTB) was installed, with

the nozzle of the irradiation facility

inside the reactor core. Additionally

in 1995 a cold neutron source was

installed.

A pneumatic tube system made it

possible to irradiate and change

scientific probes while the reactor is

running.

The reactor pool is situated inside

the reactor hall. The reactor hall is a

spheroid with 30 m height and 30 m

diameter – giving the FRM its characteristic

egg-shaped form (Atom-Ei; see

Figure 2). The open reactor pool is

built of 450 m 3 baryte and normal

concrete containing 55 metric tonnes

of steel as reinforcement. The reactor

pool holds 270 m 3 of deionized light

water as moderator and primary

coolant.

1.3 Scientific highlights

The FRM always was a reactor for

scientific purposes and the scientific

use of neutrons. Power generation

was not of interest.

The main fields of science at

FRM were: nuclear physics, neutron

physics, solid-state physics, irradiation

techniques and radiochemistry.

Ultracold neutrons (UCN) were first

detected at the FRM by A. Steyerl et al.

[2].

Many techniques that are still used

today at neutron source type research

reactors were developed or greatly

refined at FRM including:

spectroscopy, mass spectroscopy

of fission products, interferometry,

neutron guides, small angle scattering,

fast pneumatic tube system and

gravity refractometry.

1.4 End of operation

When it was decided to build a new

high flux neutron source next to FRM,

it was also decided to decommission

the old reactor. Licensing for decommissioning

started in 1998.

The last reactor operation took

place in the year 2000. Since 2002

FRM has had no fuel elements on site.

The license for decommissioning

was granted in 2014.

2 Licensing for

decommissioning

2.1 General legislation

In order to decommission a nuclear

installation in Germany the licensee of

the running installation has to request

a license for decommissioning according

to § 7 of the German Atomic

Energy Act AtG.

The licensee’s request has to

contain a detailed safety report. The

safety report has to cover the following

items:

• geographic location

• neighbouring infrastructure (residential,

industrial and agricultural

areas)

• technical description of the nuclear

installation

• description of the planned procedure

for practical decommissioning

• radiation safety

• waste management

• fire protection

• possible incidents/accidents

• personnel organisation of the

operator

After receiving the safety report, the

licensing authority asks a technical

safety organisation (TSO) for their

expert opinion regarding the safety

report and its contents. The TSO

writes a final report commenting on

each point of the safety report.

The licensing authority grants the

license for decommissioning, if it is

convinced that the decommissioning

can be done safely for people and the

environment.

2.2 Licensing at FRM

FRM requested a license for decommissioning

in 1998. A new environmental

impact assessment (Umweltverträglichkeitsprüfung

UVP) with

participation of the public was not

required due to the fact that only

already licensed values were re quested

for permissible emissions ( only 1 E6

Bq aerosol-bound activity per year).

Shortly after the first request, all

licensing efforts for decommissioning

were put on halt, as all resources from

FRM, authority and TSO were needed

for the licensing and commissioning

of the new high flux neutron source

FRM II.

Finally, in 2010 FRM negotiated a

contract with a nuclear consulting

company, to get the licensing done. In

this process the contractor prepared

all the necessary documents and

discussed them with the operator, the

TSO and the licensing authority – the

Bavarian state ministry of the environment,

health and consumer protection

(StMUV).

The TSO and the licensing authority

were involved during the whole

process of preparing the safety report.

Both, the licensing authority and the

TSO, already knew the FRM form the

operating time in great detail, which

facilitated the unavoidable technical

discussions.

Research and Innovation

Decommissioning of Germany’s First Nuclear Reactor ı Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and Anton Kastenmüller


atw Vol. 63 (2018) | Issue 6/7 ı June/July

| | Fig. 2.

Historic picture of the FRM site.

Year

Activities towards license for decommissioning

1957-2000 Reactor operation of FRM under operating license

1998 Operator (FRM) requests license for decommissioning

1998-2010 FRM, TSO and authority are busy with licensing and commissioning

of the new high flux research reactor FRM II

2010-2012 FRM working with a contractor to finalize the licensing documents

2010-2014 FRM, contractor, TSO and authority discuss details of the licensing

documents and prepare additional documents

2014 License for decommissioning is granted by the authority

| | Tab. 1.

FRM’s way from reactor operation to decommissioning.

After the submission of the safety

report by FRM, the TSO reviewed the

report and wrote a detailed report

themselves. The overall result was

positive. Still, the TSO requested some

conditions to be met before the start of

dismantling.

The licensing authority granted

the license for decommissioning in

2014 [3]. Basis of the license are:

• the safety report presented by FRM

• the technical report presented by

the TSO (conditions, written down

by the TSO, did not necessarily become

requirements of the licensing

authority)

• federal and national legislation

• additional national regulations

The whole process from reactor operation

to license for decommissioning

is summed up in Table 1.

3 Description of the

(pre-)licensing stage

This chapter describes the main

challenges during the pre-licensing

stage: the disposal of spent nuclear

fuel (chapter 3.1) and writing up the

safety report (chapters 3.2 to 3.5).

The parts of the safety report (see

chapter 2.1), that are of general

interest, are described in detail.

Figure 2 shows the “atomic egg” –

with circumferential rooms – from

the outside.

The licensed area of the FRM

consists of the reactor hall – containing

the open pool (see Figure 1)

– and additional rooms around the

“atomic egg”. The adjacent rooms

contain storage rooms, laboratories

and rooms for technical installations

such as ventilation system and waste

water treatment.

3.1 Disposal of spent nuclear

fuel

The main part of the spent nuclear

fuel from FRM was highly enriched

uranium as described above. Supporting

the international effort of minimizing

civilian HEU stockpiles, the

FRM participated in the repatriation

of the spent nuclear fuel back to the

United States in 2002. The packaging

of the SNF took place still under the

operating license.

This step reduced the radioactive

inventory at the FRM drastically. The

typical protection objectives of a

nuclear facility: control of reactivity

and heat removal, are not necessary

after the removal of all the nuclear

fuel from the FRM reactor hall.

3.2 Fire protection

Fire protection is gaining more and

more focus in international and German

regulations. This is especially the

case after tragic accidents in industrial

[4] and private building complexes.

The German Nuclear Safety Standards

Commission (KTA) published

regulations concerning fire protection

in nuclear power plants (i.e. reactors

commercially producing electricity)

under the national regulation KTA

2101 in 1985. This regulation is being

updated on a regular basis.

The fire protection at the research

reactor FRM has always been inspired

– but not governed – by this KTA rule.

Especially because of the age and

the structural design of the reactor

hall it is not always possible to fully

comply with KTA-rule 2101. There is

consensus that because of the low risk

arising from the decommissioning of

the FRM this is tolerable.

The main outline of the fire protection

concept is as follows:

• use of burnable substances and

burnable structural materials is

minimized at FRM – especially in

controlled areas

• utilization units (e.g. ventilation

system, reactor hall, emergency

battery room) are separated by

walls and doors with a fireresistance

rating that guarantees a

resistance time of 90 minutes

( following DIN 4102-5) during a

conventional fire

• every room is monitored by fire/

smoke detectors

• there are hand held fire extinguishers

for firefighting by the FRM

personnel (only small fires)

• the combination of fire-resistant

structural material and detector

systems gives the Werkfeuerwehr

(fire department of the TUM

located at the campus site [5])

enough time to arrive for firefighting

• there are enough outdoor and

indoor hydrants for the fire

department to do their work.

All the installations necessary for fire

protection and the inventory of burnable

substances on the site are checked

on a regular basis by the operator.

Additionally there are yearly site

inspections by the TSO, which also

include a detailed review of the

operator’s inspection protocols.

3.3 Radiological

characterisation

The reactor hall and most of the

reactor systems were known to be very

clean – with regard to radioactivity

RESEARCH AND INNOVATION 385

Research and Innovation

Decommissioning of Germany’s First Nuclear Reactor ı Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and Anton Kastenmüller


atw Vol. 63 (2018) | Issue 6/7 ı June/July

RESEARCH AND INNOVATION 386

2) The results

presented are

mainly from work

done by the

contractor [6]

Measured spot

Heat Exchanger

Pneumatic Tube System

Contaminated parts from the Reactor Pool

Ventilation System

Reactor Hall – concrete wall

| | Tab. 2.

Results from gamma spectroscopy.

– throughout the decades of reactor

operation. Still, there was one known

contamination event with Eu-152/

Eu-154 in the pneumatic rabbit system

that contaminated the tube system,

the reactor hall and the ventilation

system. Additionally there is a Americium

and Plutonium contamination

in at least one pneumatic tube.

In order to show that safe decommissioning

is possible with respect to

radioactive waste and safety of the

working personnel, it was necessary

to characterize the radiological situation

at the FRM – after removal of SNF.

The methodology and the main results

of the radiological characterisation

are described in the following. 2

Gamma dose rate measurements:

Gamma dose rate measurements were

executed for a first overview in the

reactor hall, the pump room – containing

the primary circuit, the heat

exchanger and the water cleaning

system - and on the reactor platform.

The dose rate levels were very low at

every place (< 5 µSv/h).

No new contamination was found.

Only places with known contamination

could be confirmed: pneumatic rabbit

system, heat exchanger, waste storage

areas and some scientific experiments.

The dose rate from activated core

components is negli gible as long as the

open pool is filled with water.

Gamma spectrometry:

In order to determine the most important

nuclides in-situ measurements

with a gamma spectrometer were

performed at spots of increased dose

rate. The summarized results are

presented in Table 2.

Relevant isotopes

Co-60, Ag-108m, Cs-137, Eu-152, Eu-154

Co-60, Ag-108m, Cs-137, Eu-152, Eu-154

Co-60, Cs-137

Co-60, Cs-137, Eu-152, Eu-154

Only naturally occurring nuclides

Nuclide Total activity in concrete [Bq] Total activity in steel [Bq]

H-3 3.1 E07

C-14 2.0 E04 1.5 E04

Fe-55 4.3 E06 1.0 E09

Co-60 2.5 E06 1.3 E07

Ba-133 1.3 E08 –

Eu-152 3.9 E06 –

Eu-154 2.7 E05 –

| | Tab. 3.

Total activity of the wall of the reactor pool.

Contamination measurements:

To determine the contamination of

the reactor hall and reactor systems,

swipes were taken and measurements

with hand held contamination monitoring

devices were performed. The

three main results were:

• there is no relevant contamination

– with one exception: inside the

pneumatic tube system

• Co-60 and Cs-137 are the main

contaminants

• there are no relevant isotopes

( excluding the pneumatic tube

system), which are difficult to

measure (e.g. H-3, Alpha emitters)

Nuclide

Mn-54

Fe-55

Co-60

Ni-59

Ni-63

Zn-65

Total activity

in aluminium [Bq]

5.2 E07

| | Tab. 4.

Total activity of the aluminum components

close to the core.

Nuclide

Mn-54

Fe-55

Co-60

Ni-59

Ni-63

1.1 E12

5.4 E11

2.5 E09

3.3 E11

5.9 E12

Total activity

in steel [Bq]

1.1 E07

3.5 E11

7.7 E09

1.7 E09

2.2 E11

| | Tab. 5.

Total activity of the steel components close to

the core.

Neutron activation calculations:

Neutron activation calculations for

the components close to the reactor

core and close to the neutron beam

tubes were performed to determine

the remaining main radioactive

inventory (see Table 3 to Table 5).

Summary:

The main part of the remaining

radioactivity in the reactor hall is

activation of the components close to

the core. The handling of the highly

activated components is not trivial

(dose rates up to > 1 Sv/h for components

very close to the reactor) but

possible. It is possible to comply

with the acceptance criteria for

final disposal. There are only a few

contaminated systems (pneumatic

rabbit tubes, primary circuit with

heat exchanger). The decommissioning

of these systems will be technically

possible.

3.4 Waste management

Clearance/free release of material

from controlled areas:

Most of the material from the controlled

areas of the FRM can be

released without any restriction. For

this clearance measurements have to

be performed by FRM and those

measurements have to be confirmed

by an independent third party (at

FRM: the Bavarian State Agency for the

Environment (LfU)). After that the

material is cleared for unrestricted use

or restricted conventional dis posal.

Conventional pollutants:

Due to the age of the FRM there are

several conventional pollutants that

have to be taken into account when

planning the disposal.

In order to shield some of the neutron

beam guide tubes going through

the reactor hall, concrete tubs containing

water were built around the

neutron guide tubes. The concrete

shielding tubs were constructed from

concrete bars that were joint by

material containing PCB (sum PCB

9.2 g/kg). The concrete bars were

covered with insulating material also

containing PCB and lead (sum PCB

2 g/kg; Lead 11 g/kg).

There are two possible ways of

disposing of the concrete contaminated

with PCB for the FRM:

• breaking up the concrete into

small parts (size of a human fist),

burning every single piece,

disposing of the pieces conventionally

or

• disposing of the concrete bars

in appropriate containers at a

licensed underground disposal

site.

Research and Innovation

Decommissioning of Germany’s First Nuclear Reactor ı Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and Anton Kastenmüller


atw Vol. 63 (2018) | Issue 6/7 ı June/July

The disposal route preferred by the

competent state disposal company is

the underground disposal site.

No asbestos could be found so far.

Disposal of radioactive waste:

The activated components and some

contaminated pieces have to be prepared

for disposal in a federal final

repository. At the moment FRM has to

prepare the radioactive waste in such

conditions (stable, packaging, limit for

activity, limit for conventional

pollutants) that it complies with the

acceptance criteria of Schacht Konrad

[7].

There are wastes at the FRM site

that cannot be brought in such a form,

that they comply with the acceptance

criteria. The problematic wastes are:

• beryllium reflector elements and

• graphite reflector elements.

The disposal of the FRM reflector

elements is an unresolved problem.

The TUM Institute for Radiochemistry

already presented information regarding

the disposal of the beryllium

elements [8].

For the licensed final repository

Schacht Konrad the H-3-inventory

(radioactivity) and the beryllium

(conventional pollutant) inventory

are too big. This is most likely also the

case for the C-14-inventory of the

graphite elements. The responsible

federal authority (former Federal

Agency for Radiation Protection (BfS –

Bundesamt für Strahlenschutz) now

Federal Office for the Safety of Nuclear

Waste Management (BfE – Bundesamt

für kerntechnische Entsorgungssicherheit))

is legally obliged to provide

a final repository for this kind of

wastes.

Until when the packaged wastes

will be safely stored in a final repository,

FRM remains the owner of

the radioactive wastes, thus also

respon sible (especially financially) to

ensure safe storage of the radioactive

material.

As stated above the spent nuclear

fuel was already repatriated to the

United States.

• destruction of the reactor building

with release of radioactivity (e.g.

after an earthquake)

• outage of important systems

For the calculation of a possible

release of radioactivity into the environment,

the inventory of radioactivity

at the FRM site was taken from the

radiological characterisation described

above (see chapter 3.3). The

possible release of activity was calculated

conservatively following national

and international guidelines.

The German Radiation Protection

Ordinance (Strahlenschutzverordnung

– StrlSchV) requires that the effective

radiation dose which follows the worst

accident conditions has to be below

50 mSv. The result of the safety analysis

was: the possible radiation dose for

every group of age is well below the

acceptable limits (also including radiation

doses to the organs).

4 Status of the project

4.1 Clearance of material

Clearance measurements are performed

routinely by the FRM radiation

protection personnel. The preferred

method of such measurements

is a circa 1 m 3 box (FMA) with gammasensitive

detectors on every side.

Clearance for unrestricted use of

material that is measured in the FMA

can be achieved within a few weeks –

including the time the independent

expert takes to verify the FRM measurements.

Material that can’t be measured in

the FMA – due to its size – can take a

much longer time to achieve the clearance

for unrestricted use (roughly half

a year). Clearance measurements for

this kind of material are performed

with in-situ gamma spectrometry and/

or hand-held contamination monitors

with additional material probes.

Clearance for restricted use (i.e.

conventional disposal sites or conventional

incineration) can take a much

longer time (up to a few years). There

are a few problems for this way of

clearance:

• hard to find appropriate sites

• hard to find sites willing to take

material from a nuclear site

• conventional authority and

nuclear/radiation protection

author ity have to communicate

• operator has to coordinate

everything

From the year 2011 until the beginning

of 2018 almost 170 metric tonnes

of material have been cleared for free

release. Clearance for restricted use

(incineration) has been achieved for

almost 100 kg.

4.2 Ventilation system

The license for decommissioning

issued by the responsible authority

described the present ventilation

system as appropriate for the time of

decommissioning.

The visualisation scheme of the

present ventilation system is shown in

Figure 3.

Because of the lack of filtration

under normal operation (this was also

the case during reactor operation) the

TSO was able to come up with a

scenario in which, because of the

requested low permissible yearly

emissions (1E6 Bq per year), the

emission limit for one year could be

exhausted into the air – undetected by

the operator – in a few hours. Although

RESEARCH AND INNOVATION 387

3.5 Safety analysis of possible

accidents

The safety analysis of possible accidents

and the possible resulting radiation

dose to the public was done in

accordance with German legislation

and regulations.

For the safety analysis a conservative

spectrum of possible incidents

was selected. They can be summed up

as follows:

• long lasting fire in the reactor hall

| | Fig. 3.

Visualisation of the unfiltered ventilations system.

Research and Innovation

Decommissioning of Germany’s First Nuclear Reactor ı Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and Anton Kastenmüller


atw Vol. 63 (2018) | Issue 6/7 ı June/July

RESEARCH AND INNOVATION 388

the scenario described by the TSO

seems to be extremely unlikely, FRM

decided to install a new ventilation

system – with filtered exhaust air – for

the time of decommissioning and the

use of the reactor hall after decommissioning.

As long as the ventilation system

is not accepted by the licensing

authority, FRM is not allowed to use

tools in the reactor hall that could

cause emission of aerosols into the air,

thus rendering impossible every

significant step in physical decommissioning.

4.3 Radioactive Waste

Most of the radioactive material is still

stored in the reactor hall. The components

that were close to the reactor

core are still in the reactor pool

covered with water.

The disposal of radioactive waste is

not going on at the moment. This is

mainly due to the following:

The radioactive material can’t be

processed or separated in the reactor

hall because of the state of the ventilation

system described in chapter 4.2.

Additionally, appropriate financial

resources and personnel have not yet

been allocated for the disposal of

radioactive waste.

4.4 Financing of the project

FRM is part of FRM II which is a central

scientific facility of TUM. In order to

pay for projects necessary for the

decommissioning (for example new

construction of the ventilation system

and disposal of radioactive waste)

FRM has to ask the Board of the TUM

for financial resources. The TUM

Board asks the responsible Ministry of

Science for money which again

requests their financial means from

the Bavarian Parliament. If the

Bavarian Parliament decides to

finance the necessary projects, the

Ministry – responsible for publicsector

construction – assigns their

Building and Construction authority

to plan the construction project (in

terms of public finance every step of

the decommissioning project is a

construction project).

This planning is than done by FRM

(and of course contractors working

for FRM) and presented to the construction

committee of the Bavarian

Parliament by the Building and Construction

authority.

If the construction committee sees

the project feasible, the Building

and Construction authority has

access to financial resources. But still

FRM is the operator responsible for

the nuclear installation. So in very

close cooperation the Building and

Construction authority and FRM are

acting as contracting authorities for

the decommissioning projects.

5 Conclusion

The history and the current state of

the FRM have been described. Once

the remaining issues will be resolved,

the decommissioning will proceed as

licensed. The complex financing

procedure described in chapter 4.4 is

adding an additional level of complexity

to the already challenging

decommissioning project.

References

[1] Festschrift der Technischen Universität

München, „40 Jahre Atom-Ei Garching“.

[2] A. Steyerl, 1969, Physics Letters 29B

33-5.

[3] Bayerische Staatsministerium für

Umwelt und Verbraucherschutz

StMUV, 03.04.2014, „Genehmigung

nach § 7 Atomgesetz (AtG) zum Abbau

der Reaktoranlage des Forschungsreaktors

München FRM in Garching“.

[4] Der Spiegel, 15.12.1999, „Chronologie:

Die Brandkatastrophe am Düsseldorfer

Flughafen.“

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Research and Innovation

Decommissioning of Germany’s First Nuclear Reactor ı Ulrich Lichnovsky, Julia Rehberger, Axel Pichlmaier and Anton Kastenmüller


atw Vol. 63 (2018) | Issue 6/7 ı June/July

[5] History of the Werkfeuerwehr, Technical

University of Munich, Garching,

https://www.feuerwehr.tum.de/index.

php?id=30.

[6] S. Thierfeldt et. al. (Brenk Systemplanung),

March 2013, „Sicherheitsbericht

zur Stilllegung des Forschungsreaktors

München (FRM) der Technischen

Universität München”.

[7] K. Kugel, K. Möller (BfS), 20.2.2017,

„Anforderungen an endzulagernde

radioaktive Abfälle (Endlagerungsbedingungen,

Stand: Februar 2017) –

Endlager Konrad –”.

[8] C. Lierse et. al., 28.12.2010, Abschlussbericht

FKZ02S7951: „Entsorgung von

Beryllium/Berylliumoxid und Cadmium

aus Forschungsreaktoren“.

Abbreviations

FMA

HEU

KTA

PCB

SNF

Freimessanlage (Release

Measurement Facility)

Highly Enriched Uranium

Kerntechnischer Ausschuss (Nuclear

Safety Standards Commission)

Polychlorinated biphenyl

Spent Nuclear Fuel

StMUV Staatsministerium für Umwelt,

Gesundheit und Verbraucherschutz

(Bavarian state ministry of the

environment, health and consumer

protection)

TSO

Technical Safety Organisation

Authors

While You Were Sleeping:

The Unnoticed Loss of Carbon-free

Generation in the United States

Chris Vlahoplus, Ed Baker, Sean Lawrie, Paul Quinlan and Benjamin Lozier

M. Sc. Ulrich Lichnovsky

Fachbereichsleiter Stilllegung und

Rückbau FRM(alt)

B. Sc. Julia Rehberger

Ingenieurin Dokumentation und

Änderungsdienst FRM(alt)

Dr. Axel Pichlmaier

Fachbereichsleiter Reaktorbetrieb

FRM II

Dr. Anton Kastenmüller

Technischer Direktor FRM II

FRM II, Forschungs-Neutronenquelle

Heinz Maier-Leibnitz

Technische Universität München

Lichtenbergstr. 1

85748 Garching, Germany

389

ENERGY POLICY, ECONOMY AND LAW

The United States has embarked on actions to combat climate change by putting a focus on lowering the carbon

emissions from the electric generation sector. A pillar of this approach is to promote the greater use of renewable

resources, such as wind and solar. The past decade has seen significant growth in carbon-free energy from wind and

solar. Generation from these resources reached 333,000 GWh in 2017. However, unbeknownst to many who care about

climate change, most of the progress made to date through renewables is at significant risk due to the loss or potential

loss of more than 228,000 GWh of nuclear carbon-free generation.

Renewables growth:

Investment in carbon-free

generation

Over the past decade, wind and solar

have grown in large part due to

policies such as renewable portfolio

standards, federal tax incentives, and

in some cases state tax incentives. Few

would argue that the addition of

renewable generation is a critical

element of a comprehensive carbonreduction

strategy.

Since 2008, the policy focus on

renewables has attracted hundreds of

billions of dollars of investment for

the development of wind and solar.

The results have been significant – in

the past decade 90 % of the current

operating wind and solar capacity was

added, roughly 75 GW of wind and

52 GW solar. [1] Another result of

these investments has been to help

wind and solar drive down the cost

curve reaching a more competitive

position. The policies promoting

renewables have clearly contributed

to the addition of a meaningful

amount of carbon-free electricity as

well as to jump-starting an industry in

the United States.

Early retirement: Nuclear

generators face challenges

In the same timeframe, natural gas

prices have driven down power prices,

causing difficulties for both renewables

and existing generation. The

nuclear industry in particular has

been challenged by low natural gas

prices and the lack of overall policy

support for its zero-carbon attributes.

As a result, the nuclear industry has

faced a wave of actual and announced

retirements. The most vulnerable

nuclear plants have been small, singleunit

plants and merchant facilities in

deregulated markets with low energy

and capacity values. Under these

conditions, existing nuclear plants are

having difficulty competing in bidbased

markets and in some regulated

as well. Some states have recognized

this issue and have explored zerocarbon

incentives to keep plants open

that would otherwise have shut down.

However, these incentives are being

challenged and still make these plants,

while technically “reprieved,” what

we categorize as “at risk.”

In 2016, the New York Public Service

Commission approved a Clean Energy

Standard (CES), which supported

the continuation of more than 3 GW

of nuclear capacity (i.e., Fitzpatrick,

Ginna, and Nine Mile Point nuclear

plants). [2] In the same year, Illinois

passed The Future Energy Jobs Bill

that provides nuclear plants with

$ 0.01/kWh, saving almost 3 GW of

nuclear capacity (i.e., Clinton and

Quad Cities nuclear plants). [3] The

actions in New York and Illinois

sustained more than 50,000 GWh

of carbon-free generation per year.

Meanwhile, Connecticut recently

estimated it would cost roughly

$ 5.5 billion to replace the carbon-free

generation from Dominion Energy’s

Millstone station with renewables. [4]

To understand the potential for

loss of carbon-free generation, Scott-

Madden identified four categories of

“at-risk” nuclear assets. Each nuclear

plant operating in 2008 (a date that

coincides with the rapid growth in

renewables) was reviewed and, if

applicable, placed into one of the

following “at-risk” categories:

• Retired – Any nuclear plant that

has ceased operations since 2008.

Some plants on the list had physical

Energy Policy, Economy and Law

While You Were Sleeping: The Unnoticed Loss of Carbon-free Generation in the United States ı Chris Vlahoplus, Ed Baker, Sean Lawrie, Paul Quinlan and Benjamin Lozier


atw Vol. 63 (2018) | Issue 6/7 ı June/July

ENERGY POLICY, ECONOMY AND LAW 390

Category Nuclear plants Capacity

(MW)

Retired

Announced

Crystal River, Fort Calhoun, Kewaunee, San Onofre, and

Vermont Yankee

Beaver Valley, Davis-Besse, Diablo Canyon, Indian Point,

Oyster Creek, Palisades, Perry, Pilgrim, and Three Mile Island

issues driving retirement, but they

may have continued to operate

under different economic circumstances,

including markets valuing

carbon-free generation

• Announced – Any nuclear plants

where the owner has announced

plans to cease operations early

• In Jeopardy – Any nuclear plant

where the owner has indicated

the plant may close if market

conditions do not improve

• Reprieved – Any nuclear plant

that has received state support to

remain open. These were on the

cusp of closure, and absent followthrough

on these programs, the

plants will likely close

For each “at-risk” category, we calculated

total capacity and annual generation.

[5] As seen in the Table 1

below, more than 28,000 MW of

Generation

(GWh)

4,674 37,795

11,109 89,818

In Jeopardy Duane Arnold, Hope Creek, Millstone, and Salem 6,189 50,044

Reprieved Clinton, Fitzpatrick, Ginna, Nine Mile Point, and Quad Cities 6,232 50,388

“At-risk” nuclear total: 28,204 228,045

| | Tab. 1.

Nuclear capacities retired or facing early retirement in the U.S.

| | Fig. 1.

Change in U.S. Carbon-Free Generation.

| | Fig. 2.

In-State Renewable Generation vs. “At-Risk” U.S. Nuclear Generation.

nuclear capacity has retired or

is facing early retirement. The

228,045 GWh of nuclear generation

retired since 2008 or at risk of early

retirement represents 5.6 % of total

U.S. net generation in 2016.

Carbon impact: Early

retirement of nuclear

diminishes renewable gains

To understand the potential impact on

carbon-free energy, ScottMadden compared

“at-risk” nuclear assets facing

early retirement to all wind and solar

assets operating at the end of 2017. [6]

As discussed previously, there has been

great publicity around the wind and

solar capacity that has been added over

the past decade. If compared on this

popular measure, nuclear capacity at

risk of early retirement only accounts

for 20 % of the total 2017 renewable

capacity. If at first glance, it is not

alarming, it is because capacity is not

the right measure to show impact on

carbon. To understand that, we must

compare on electric output, or energy.

When compared on energy output,

the potential loss of nuclear presents a

greater concern. With capacity factors

greater than 90 %, losing a smaller

amount of nuclear can produce outsized

impacts on carbon-free generation

compared to the low-capacity

factor of wind and solar (35 % to

22 %). [7] In 2017, wind and

solar produced a combined total of

333,000 GWh of carbon-free generation

(see Figure 1). This gain has the

potential to be reduced by 68 % or

228,045 GWh through the early

retirement of nuclear capacity. In fact,

the United States has lost 11 % of the

renewable generation from plants

already retired.

In the states that host “at-risk”

nuclear assets, the potential lost

carbon- free generation from nuclear

energy exceeds total in-state renew able

energy generation (see Figure 2). [8]

This represents a sig nificant barrier to

achieving near-term state- level reductions

in greenhouse gas emissions.

A further potential challenge is

the relicensing of nuclear plants.

Those plants not currently at risk of

early retirement must renew their operating

licenses with the U.S. Nuclear

Regulatory Commission in the next

20 years. If these plants do not renew

their licenses, even more carbon- free

generation would be lost. In fact, wind

and solar output would need to more

than double just to break even on the

loss of carbon-free generation from

the retirement of the entire nuclear

fleet (see Figure 3).

Germany: A cautionary tale

Many have pointed to Germany as a

shining example of a country that has

led the way in deploying renewables.

In 2000, the Renewable Energy Act

established feed-in tariffs and priority

grid access for renewables. The action

represents a key milestone in the

Energiewende or transition to a lowcarbon

economy based on renewable

resources. Since then, the country has

spent roughly $ 222 billion on renewable

subsidies. [9] The result is renewable

energy as a percentage of gross

electricity generation increasing from

6.2 % in 2000 to 31. 3% in 2015. [10]

At the same time however, Germany

has embarked on a strategy of shuttering

its nuclear plants. Roughly 40 %

of the country’s nuclear capacity

was shut down in 2011, following the

Energy Policy, Economy and Law

While You Were Sleeping: The Unnoticed Loss of Carbon-free Generation in the United States ı Chris Vlahoplus, Ed Baker, Sean Lawrie, Paul Quinlan and Benjamin Lozier


atw Vol. 63 (2018) | Issue 6/7 ı June/July

| | Fig. 3.

Total U.S. Renewable Generation vs. Total U.S. Nuclear Generation.

| | Fig. 4.

Annual greenhouse gas emissions in German electricity sector.

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top 20 energy utilities. We have performed

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based.

ENERGY POLICY, ECONOMY AND LAW 391

Fukushima nuclear accident. [11] As

a result, despite the addition of significant

renewable resources, there is

limited progress in reducing total

carbon emissions in the electricity

sector due to the early retirement of

nuclear plants. In fact, greenhouse gas

emission from the electricity sector has

only decreased 3 % from 2000 to 2015

(see Figure 4). [12]

Conclusion: Rapid and deep

carbon reductions require

nuclear assets

Investments in renewables have made

a significant contribution to emissionfree

electricity generation. For those

concerned with climate change, this

represents a meaningful step in the

right direction. The early retirement of

“at-risk” nuclear, however, puts the

United States in danger of “giving

back” an amount equivalent to twothirds

of the overall carbon-free generation

supplied from wind and solar. In

states with these nuclear assets, the

loss represents a significantly larger

impact. The losses could become even

greater if more nuclear plants do not

renew operating relicenses.

However, a glimmer of hope

emerges as states, such as New York

and Illinois, are developing policies

to value the carbon-free generation

provided by nuclear plants. Even

environ mentalists are beginning to

offer support for nuclear energy. In

Illinois, the Union of Concerned

Scientists called the Future Energy

Jobs Bill “one of the most comprehensive

state energy bills ever crafted

and is the most important climate

bill in Illinois history.” [13] In addition,

an open letter signed by more

than 70 ecologists and conservation

researchers stated that wind and solar

are promising, but “nuclear power –

being by far the most compact and

energy dense of sources – could also

make a major, and perhaps leading,

contribution” to carbon emission

reductions. [14]

If nuclear plants are not saved in

the near term, it will put the entire

industry at risk. For once a nuclear

plant shuts down, it will not come

back. If enough nuclear plants shut

down, a tipping point may be reached

for the entire industry in the United

States, and we will lose forever that

carbon-free generation. [15] While

one might argue that in the long run,

this nuclear hole may be filled with

renewables and other evolving clean

technologies, in the near term it is

certain that a rapid and deep carbon

reduction will require these nuclear

assets.

About the Authors

Chris Vlahoplus is a partner and leads

the firm’s Clean Tech & Sustainability

practice, Ed Baker is a partner and

co-leads the firm’s nuclear and gas

practices, Sean Lawrie is a partner and

co-leads the firm’s nuclear practice,

Paul Quinlan is a clean tech manager,

and Benjamin Lozier is an energy

research analyst.

This report is one of a series of

ScottMadden white papers on clean

energy technologies and is based on

our independent analysis. The

contents have been updated from the

initial version to reflect the recent

change of FirstEnergy Solutions units

from “in jeopardy” to “announced.”

References

[1] Data obtained from Bloomberg New

Energy Finance’s 2018 Sustainable

Energy in America Factbook. Wind

capacity is reported in AC; solar

capacity is reported in DC.

[2] New York State Department of Public

Service, Governor Cuomo Announces

Establishment of Clean Energy Standard

that Mandates 50 Percent Renewables

by 2030.

[3] Forbes, Illinois Sees The Light – Retains

Nuclear Power. December 4, 2016.

Energy Policy, Economy and Law

While You Were Sleeping: The Unnoticed Loss of Carbon-free Generation in the United States ı Chris Vlahoplus, Ed Baker, Sean Lawrie, Paul Quinlan and Benjamin Lozier


atw Vol. 63 (2018) | Issue 6/7 ı June/July

ENERGY POLICY, ECONOMY AND LAW 392

[4] Connecticut Department of Energy &

Environmental Protection Connecticut

Public Utilities Regulatory Authority,

Resource Assessment of Millstone

Pursuant to Executive Order No. 59 and

Public Act 17-3; Determination

Pursuant to Public Act 17-3. February 1,

2018.

[5] Capacity was calculated using net

summer peak capacity obtained from

SNL Financial. Generation was

calculated using 92.3 % capacity factor,

which represents the average capacity

factor for the U.S. nuclear fleet in 2016

as reported by the Energy Information

Administration.

[6] Data obtained from Bloomberg New

Energy Finance’s 2018 Sustainable

Energy in America Factbook. Wind

capacity is reported in AC; solar

capacity is reported in DC.

Deutsche Sekretariatsführung ISO/TC

85/SC 6 Reactor-Technology

Janine Winkler und Michael Petri

[7] Average capacity factor of utility-scale

generators in 2016: nuclear 92.3 %,

wind 34.5 %, utility-scale solar 25.1 %,

and solar thermal 22.2 %. Source:

Energy Information Administration,

Electric Power Annual. Distributed solar

capacity factors are often below 20 %.

[8] These states include California,

Connecticut, Florida, Illinois, Iowa,

Massachusetts, Michigan, Nebraska,

New Jersey, New York, Ohio,

Pennsylvania, Vermont, and Wisconsin.

[9] The New York Times. Germany’s Shift to

Green Power Stalls, Despite Huge

Investments. October 7, 2017.

[10] German Environment Agency on the

basis of Working Group on Renewable

Energy Statistics (AGEE-Stat)

[11] The Economist. Is Germany's Energiewende

Cutting GHG Emissions? March

20, 2017.

[12] United Nations Framework Convention

on Climate Change Data Interface

[13] Union of Concerned Scientists. A Huge

Success in Illinois: Future Energy Jobs Bill

Signed Into Law. December 8, 2016.

[14] The Washington Post. Why Climate

Change Is Forcing Some

Environmentalists to Back Nuclear

Power. December 16, 2014.

Authors

Chris Vlahoplus

Ed Baker

Sean Lawrie

Paul Quinlan

Benjamin Lozier

ScottMadden

2626 Glenwood Ave., Suite 480

Raleigh, NC 27608, USA

Im Auftrag des Bundesministeriums für Umwelt, Naturschutz und nukleare Sicherheit (BMU), vertreten durch die

KTA-Geschäftsstelle, hat DIN ab 2018 die Sekretariatsführung des ISO/TC 85/SC 6 Reactor technology zusammen mit

China übernommen. Ziel ist hierbei, den deutschen Einfluss in der internationalen Normung zu erhöhen und die

Möglichkeit wahrzunehmen, deutsche Normen und KTA-Regeln in die internationalen Normen zu überführen. Damit

möchte Deutschland international seinen Beitrag für die nukleare Sicherheit bei der friedlichen Nutzung der

Kernenergie leisten.

Einleitung

Im Jahr 2017 fanden bei DIN mehrere

Workshops in Zusammenarbeit mit

dem Bundesministerium für Umwelt,

Naturschutz und nukleare Sicherheit

(BMU) und der KTA-Geschäftsstelle

(Kerntechnischen Ausschuss – KTA)

statt, um den aktuellen Stand der

Normung im Bereich der Kerntechnik

zu eruieren und das weitere Vorgehen

über das Jahr 2022 hinaus zu planen.

Die Workshops haben verdeutlicht,

dass der aktuelle Stand der Gesetzgebung

im Verbund mit den nachgeordneten

Regeln in der Kerntechnik

und im Strahlenschutz verlässlich

ist und die deutschen Regeln international

sehr geachtet sind und bereits

in einigen Ländern verwendet

werden. Die Herausforderung ist es

daher, die aufgebaute Kompetenz zu

erhalten und zusammen mit anderen

internationalen Institutionen weiterzuentwickeln.

Zusätzlich sind die veränderten

Anforderungen in Bezug auf den

bevorstehenden Ausstieg Deutschlands

aus der Kerntechnik zu berücksichtigen.

Bei DIN hat dies bereits zu

einer Verlagerung der Aktivitäten

geführt, wobei der Fokus nun stärker

auf dem Bereich der Normen zum

Strahlenschutz liegt. Hierzu zählen

die Normen für z.B. Radionuklidlabore

für die Medizin, den Brandschutz

oder auch Abschirmeinrichtungen.

DIN wird auch weiterhin

sicherstellen, die Normen für den

Leistungsbetrieb auf dem aktuellen

Stand von Wissenschaft und Technik

zu halten.

Im Kerntechnischen Ausschuss

(KTA) wird ebenfalls diskutiert, wie

die Regelwerksarbeit im KTA – über

2022 hinaus – gestaltet werden kann.

In diesem Zusammenhang hat das

BMU einvernehmlich mit dem KTA-

Präsidium beschlossen, dass die KTA-

Geschäftsstelle (KTA-GS) verstärkt die

Koordination und Mitarbeit bei der

internationalen Normung für Kernkraftwerke

übernehmen soll. Dies

impliziert eine aktivere Mitarbeit der

KTA-GS in allen wichtigen internationalen

Normungsgremien, insbesondere

in den Gremien, in denen

dies bisher nur punktuell erfolgt ist

(ISO, CEN, ASME), um die deutschen

Interessen, die im KTA gebündelt

sind (Betreiber, Hersteller, Behörden,

Gutachter), möglichst effektiv auch

international zu vertreten und zu

koordinieren. Wesentliche Gesichtspunkte

bei dieser Entscheidung waren

die enge Verzahnung zwischen

KTA-Regeln und den thematisch zugehörigen

nationalen Normen. Die

zunehmende Bedeutung der Standardisierung

im internationalen Bereich

(insbesondere bei kerntechnischen

„Newcomern“ wie z. B. Bangladesch,

Türkei, VAE) erfordert ein aus

deutscher Sicht ausreichendes Niveau

sowie die Schließung noch vorhandener

Lücken im internationalen Regelwerk.

Außerdem ist festzustellen, dass

aufgrund der begrenzten Laufzeit der

deutschen Kernkraftwerke bis Ende

2022 eine abnehmende Bereitschaft

seitens der Betreiber und Gutachter

absehbar und auch bereits erkennbar

ist, Experten in internationale Regelgremien

zu entsenden. In Abstimmung

zwischen BMU und dem KTA-

Präsidium ist daher vorgesehen, dies

– soweit möglich – durch eine verstärkte

Aktivität von Mitarbeitern der

Energy Policy, Economy and Law

German Secretarial Management ISO/TC 85/SC 6 Reactor Technology ı Janine Winkler and Michael Petri


atw Vol. 63 (2018) | Issue 6/7 ı June/July

KTA-GS und durch Entsendung von

weiteren Experten im Auftrag des

BMU aufzufangen.

1 Zusammenhang

zwischen KTA-Regeln,

internationalen und

nationalen Normen

Der Zusammenhang zwischen technischen

Normen, wie den Regeln des

Kerntechnischen Ausschusses (KTA),

und DIN-Normen und den einschlägigen

Gesetzen und Verordnungen

ist rechtlich begründet (siehe

Abbildung 1). Während das Atomgesetz

[1] und das Strahlenschutzgesetz

[2] Hoheitsaufgaben des

Staates sind, liegt die Erarbeitung von

technischen Normen in der Selbstverwaltung

der Wirtschaft. Dies bedeutet

zum einen, dass die staatlichen Vorgaben

meist einen sehr viel geringeren

Detaillierungsgrad aufweisen und

somit auf weiterführende Dokumente

verweisen müssen (wie z.B. auf DIN-

Normen), und zum anderen, dass,

obwohl der Detaillierungsgrad größer

ist, technische Normen wesentlich

flexibler hinsichtlich der Erarbeitung

und Überarbeitung sind. Die Erarbeitung

von technischen Normen ist ein

offener Prozess (siehe auch Abschnitt

4), wobei die Beteiligung einer

breiteren Öffentlichkeit durch Einspruchsverfahren

erfolgt [3].

Es wird in 22 DIN-Normen auf

KTA-Regeln verwiesen, während andererseits

in 67 KTA-Regeln auf

DIN-Normen verwiesen wird. Diese

Zahlen verdeutlichen die enge Verflechtung

von DIN-Normen und KTA-

Regeln und zeigen auch, dass DIN-

Normen praxisrelevant für den Alltag

in kerntechnischen Anlagen sind. Als

Beispiel kann hier die KTA 3201.4 [4]

aufgeführt werden, in welcher nicht

weniger als 14-mal auf DIN-Normen

verwiesen wird. Formulierungen wie

„nach dem Stand von Wissenschaft

und Technik“ deuten auf weitere Verweise

auf Normen hin, werden aber

nicht explizit angegeben.

Zwischen dem KTA und DIN e. V.

besteht ein Vertrag [5], der genau

regelt, wer welche technischen Regeln

aufstellt, wobei es unter § 2 heißt „das

DIN erarbeitet ferner solche Normen,

die das Regelwerk des KTA außerhalb

des sicherheitstechnischen Bereiches

ergänzt“. Ein für die internationalen

Normungsaktivitäten wichtiger Aspekt

wird unter § 9 der Vereinbarung angesprochen

und besagt „das DIN vertritt

die Fachmeinung des KTA auf internationaler

und europäischer Ebene“.

Die Erarbeitung von neuen KTA-

Regeln und DIN-Normen im Bereich

| | Abb. 1.

Pyramide.

der Kernreaktoren ist allerdings stark

rückläufig, was auch Auswirkungen

auf die Beteiligung in den jeweiligen

Gremien hat. Eine Fortschreibung

bzw. Aktualisierung wird somit immer

schwieriger. Daher muss ein neuer

Weg beschritten werden und der erarbeitete

Stand in andere Normen und

Regeln einfließen, die auch in Zukunft

bearbeitet werden. Wie auch in

anderen Bereichen außerhalb der

Kerntechnik ist daher die Über führung

in die internationale Normung eine

Möglichkeit den erarbeiteten Stand

von Wissenschaft und Technik zu

erhalten und weiterzugeben (Abbildung

2).

2 Internationale

Normungsarbeit

des ISO/TC 85/SC 6

Das ISO/TC 85/SC 6 Reactor technology

ist zuständig für die Normung

im Bereich Kernkraftwerke und

Forschungsreaktoren. Der Geltungsbereich

umfasst dabei Standortauswahl,

Konstruktion, Bau, Betrieb

und Stilllegung. Die Standortauswahl

umfasst alle Arten von kerntechnischen

Anlagen und alle Themen wie

Hochwasser, seismische Gefahren

usw. Forschungsreaktoren umfassen

eine Vielzahl von Einrichtungen:

Erzeugung von Neutronenstrahlen,

Bestrahlung von Proben, Herstellung

von Isotope (insbesondere Produktion

für Nuklear medizin) und Testreaktoren

oder Prototypen neuer Technologien.

Die Normung im Bereich

der Stilllegung beschränkt sich

auf reaktorspezifische technische

Themen.

Die Erarbeitung der internationalen

Normen des ISO/TC 85/SC 6

findet in den einzelnen Arbeitsgruppen

des SC 6 statt (Abbildung

3). Die Abstimmung über Annahme

und weiteres Vorgehen zu den

| | Abb. 2.

Ablösung nationaler Normung durch internationale und europäische

Normen.

internationalen Norm-Entwürfen

findet im SC 6 selbst statt. Das SC 6

arbeitet dabei weitgehend unabhängig

vom übergeordneten ISO/TC

85, mit welchem lediglich eine

Koordinierung der Arbeiten stattfindet.

Bezüglich einer engeren

Zusam menarbeit ist insbesondere das

SC 5 zu nennen, welches teilweise in

gleichen Auf gabengebieten arbeitet,

allerdings mit dem Kernthema des

Brennstoffkreislaufes.

Die Arbeitsgruppe 1 (Working

Group 1 – WG 1) beschäftigt sich

mit der Entwicklung, der Pflege und

Förderung von Normen für Berechnungen,

Analysen und Messungen

zur Unterstützung der Reaktorkernphysik.

Solche internationalen

Normen liefern u. a. Kriterien für

die Auswahl nuklearer Daten und

com putergestützter Methoden; geeignete

Benchmark-Problemspezifikationen

zur Verifizierung der vom

Reaktorkern verwendeten Berechnungs

methoden, Kriterien für die

Bewertung der Genauigkeit und der

Anwendbarkeit von Datenmethoden;

Methoden der Verifikation und der

Abschätzung von Unsicherheiten. Die

WG 1 wird von einem amerikanischen

ENERGY POLICY, ECONOMY AND LAW 393

Energy Policy, Economy and Law

German Secretarial Management ISO/TC 85/SC 6 Reactor Technology ı Janine Winkler and Michael Petri


atw Vol. 63 (2018) | Issue 6/7 ı June/July

ENERGY POLICY, ECONOMY AND LAW 394

| | Abb. 3.

Struktur des Unterkomitees 6 im ISO/TC 85.

Convenor geführt und bearbeitet

derzeit 6 Projekte:

• ISO/NP 19226 Determination of

neutron fluence and displacement

per atom (dpa) in reactor vessel

and internals

• ISO/NP 18077 Reload Startup

Physics Tests for Pressurized Water

Reactors

• ISO/DIS 10979 Identification of

fuel assemblies for nuclear power

reactors

• ISO/AWI 23018 Group-Averaged

Neutron and Gamma-Ray Cross

Sections for Radiation Protection

and Shielding Calculations for

Nuclear Reactors

• ISO/NP 23468 Determination

of heavy water isotopic purity

by Fourier Transform Infrared

Spectroscopy

• ISO/PWI 18156 Technical specification

guide for decay heat computation

codes in nuclear reactors

Die Arbeitsgruppe 2 (WG 2) beschäftigt

sich mit der Entwicklung, Pflege

und Förderung von Normen für

Auslegung, Konstruktion, Betrieb,

Wartung, Nutzung und Stilllegung

von Forschungs- und Testreaktoren.

Die WG 2 wurde bislang ebenfalls

von einem amerikanischen Convenor

geführt, der diese Position letztes Jahr

übernommen hat aber nicht weiter

wahrnehmen kann. Dieses Jahr hat

sich China bereit erklärt die Führung

der Arbeitsgruppe 2 zu übernehmen,

der vorgesehene Fokus der Arbeiten

wird sich dann auf Fusionsreaktoren

sowie Forschungsreaktoren für die

Herstellung von Isotopen für die

Nuklearmedizin richten. Die Arbeitsgruppe

hat allerdings derzeit keine

Projekte.

Die Arbeitsgruppe 3 (WG 3)

beschäftigt sich mit der Entwicklung,

der Pflege und Förderung von Normen,

die sich mit allen Themen zum Standort,

Konstruktion, Betrieb und Stilllegung

von Kernkraftwerken auseinander

setzen. Der Betrieb umfasst

dabei auch die Notaus rüstungen.

Weiterhin werden inter nationale Normen

für nicht- elek trischen Anwendungen

und trans portable Kernreak toren

erarbeitet. Die WG 3 wird durch einen

französischen Convenor geführt und

hat momentan mit 12 Projekten ein

sehr anspruchsvolles Arbeits pensum.

• ISO/DIS 18195 Method for the

justification of fire partitioning in

water cooled NPP

• ISO/DIS 20890-1 In-service inspections

for primary coolant

circuit components of light water

reactors – Part 1: Mechanized

ultrasonic testing

• ISO/DIS 20890-2 In-service inspections

for primary coolant

circuit components of light water

reactors – Part 2: Magnetic particle

and penetrant testing

• ISO/DIS 20890-3 In-service inspections

for primary coolant

circuit components of light water

reactors – Part 3: Hydrostatic

testing

• ISO/DIS 20890-4 In-service inspections

for primary coolant

circuit components of light water

reactors – Part 4: Visual testing

• ISO/DIS 20890-5 In-service inspections

for primary coolant

circuit components of light water

reactors – Part 5: Eddy current

testing of steam generator heating

tubes

• ISO/DIS 20890-6 In-service inspections

for primary coolant

circuit components of light water

reactors – Part 6: Radiographic

testing

• ISO/CD 21146 Classification of

Transients and Accidents for

Pressurized Water Reactor

• ISO/NP 23466 Design criteria for

the thermal insulation of reactor

coolant system main equipments

and pipings of PWR nuclear power

plants

• ISO/NP 23467 Guidance of ice

plug isolation technique for

nuclear power station

• ISO/PWI 18583 Technical Spe cifications

for the connection of Mobile

Equipments for Emergency Intervention

on Nuclear Instal lation

• ISO/PWI 19462 Criteria for Assessing

Atmospheric Effects on the

Ultimate Heat Sink

In diesem Arbeitsprogramm sei

insbesondere auf die Reihe ISO/DIS

20890 hingewiesen, welche der erste

deutsche Vorschlag im SC 6 ist und

der deutschen Normenreihe DIN

25435 in großen Teilen entspricht.

Das vorhandene Arbeitsvolumen

wird voraussichtlich weiter zunehmen,

da ein Großteil der DIN-Normen

und KTA-Regeln, die als Grundlage

für die Erarbeitung entsprechender

internationaler Normen geeignet

sind, in den Aufgabenbereich der

WG 3 fallen.

Bis vor etwa 6 Jahren war das SC 6

eher inaktiv, mit insgesamt 6 veröffentlichten

Normen, von denen eine

zurückgezogen wurde. Diese Normen

wurden vor etwa 20 Jahren erarbeitet.

Vor etwa 6 Jahren, nach den Reaktorunfällen

von Fukushima, kam es zu

einer Wiederbelebung. Das Sekretariat

wurde in diesem Zeitraum durch

die USA (ANSI) geführt. Im Jahr 2017

und 2018 wurden 5 neue Normen erarbeitet

und veröffentlicht, was zu

einer Verdopplung der vorhandenen

Normen im Arbeitsprogramm des SC

6 geführt hat.

Auf seiner Plenarsitzung in Helsinki

hat das SC 6 weiterhin festgelegt,

dass es eine Roadmap für die Ausrichtung

seines zukünftigen Arbeitsprogrammes

erstellen wird. Hierzu wird

es im Rahmen der nächsten Sitzungswoche

im Mai 2019 in Berlin einen

Workshop geben, bei dem die Bedürfnisse

der beteiligten Länder herausgearbeitet

werden sollen. Weiterhin soll

dann auch die Zusammenarbeit mit

dem SC 5 sowie anderen Organisationen,

wie IAEA, Europäische Kommission

untersucht und systematisch

aufgearbeitet werden.

3 Deutsche

Sekretariatsführung

Im August 2017 hat ANSI mitgeteilt,

dass sie das Sekretariat des SC 6 aus

Energy Policy, Economy and Law

German Secretarial Management ISO/TC 85/SC 6 Reactor Technology ı Janine Winkler and Michael Petri


atw Vol. 63 (2018) | Issue 6/7 ı June/July

kapazitiven Gründen abgeben wollen.

Daraufhin hat das ISO/TC 85 eine

Umfrage zur Neuvergabe des Sekretariats

durchgeführt.

Diese Umfrage hat DIN an seine

interessierten Kreise im Fachbeirat

„Kerntechnik und Strahlenschutz“

weitergegeben. Seitens des BMU gab

es großes Interesse, dieses Sekretariat

aus strategischen Gründen zu besetzen.

Eine Zusage der Finanzierung

wurde ebenfalls ausgesprochen. Ziel

des BMU ist es, ein Konzept zur

Wahrung der Kompetenz durch die

Mitarbeit und die Einbringung des

deutschen Standes von Wissenschaft

und Technik in verschiedene internationale

Regelsetzungen umzusetzen.

Die Motivation des BMU ist es,

einen hohen Qualitätsstandard in den

internationalen Regelwerken sicherzu

stellen, der den nationalen sicherheitstechnischen

Anforderungen in

der Anwendung der friedlichen Kerntechnik

und der sonstigen Anwendungen,

wie z. B. Forschung, Strahlenschutz,

Medizin etc. gerecht wird.

Als erster Schritt soll daher die

Sekretariatsführung durch DIN diesen

hohen Standard bei den vorhandenen

Projekten, neuen Vorschlägen für

internationale Normen und den

älteren ISO-Normen des SC 6 unterstützen.

DIN hat sich daher um die

Sekretariatsführung beworben und

dieses nach einem positiven Beschluss

des ISO Technical Management Board

übernommen. Als nächster Schritt ist

vorgesehen, in Zusammenarbeit mit

der Geschäftsstelle des Kerntechnischen

Ausschusses (KTA-GS) fehlende

Standards in den Bereichen Hochwasserschutz,

Seismik und Blitzschutz

einzubringen und als internationale

Normen zu etablieren. Weiterhin

wird das BMU auch Experten

von DIN unterstützen, welche sich in

der internationalen Normung beteiligen

um DIN-Normen im Bereich der

Konstruktion, Betrieb und Konta mina

tions überwachung/Freimessungen

in die internationale Normung zu

überführen.

Ein weiteres Ziel des deutschen

Sekretariats wird es sein, die Teilnahme

der aktiven und passiven Mitglieder

von ISO/TC 85/SC 6 in den

Sitzungen und bei der Erarbeitung

von Normen in den drei Arbeitsgruppen

zu verbessern. Darüber

hinaus soll die Zusammenarbeit

innerhalb des ISO/TC 85 und die

Verbindungen zu Organisationen

außerhalb der ISO-Struktur, wie

z. B. zu IAEA, aktiver gestaltet werden.

Die Zusammenarbeit innerhalb

des ISO/TC 85 mit den anderen

| | Abb. 4.

Sitzung des ISO/TC 85/SC 6/WG 3 im Mai 2018 in Helsinki.

beiden Unterkomitees innerhalb von

ISO/TC 85 und insbesondere mit

ISO TC 85/SC 5, Nuclear installations,

processes and technologies soll intensiviert

werden. Weiterhin hat Deutschland

die Unterkomitees SC 5 und SC 6

2019 zur jährlichen Sitzung nach

Berlin eingeladen, um die Idee einer

engen Zusammenarbeit zu unterstützen

(Abbildung 4).

Das ISO/TMB hat weiterhin entschieden,

dass DIN die Sekretariatsführung

des SC 6 nicht alleine übernimmt,

sondern dem Vorschlag von

DIN folgend, zusammen mit China als

twinning secretariat. Dies bedeutet,

dass ein entwickeltes Land, welches

sehr aktiv bei ISO ist und die ISO-

Regeln beherrscht, ein sich entwickelndes

Land, welches noch nicht

so aktiv bei ISO ist und die ISO-Regeln

nicht im Detail kennt, unterstützt.

Beide Partner innerhalb des twinning

secretariat müssen sich mittels

einer Vereinbarung über die Aufgabenteilung

verständigen. Das primäre

Ziel eines twinning secretariat ist es,

die Fähigkeit des sich entwickelnden

Landes aufzubauen und die Teil nahme

an der ISO-Arbeit zu ver bessern. Die

Ziele bei einem twinning secretariat

sollten daher auch die Ziele des

Entwicklungslandes und seine nationalen

Pläne sowie Strategien umfassen.

Die twinning secretariat-

Vereinbarung sollte daher einen kontinuierlichen

Prozess der Einbindung

und Beteiligung des Entwicklungslandes

in die Sekretariatsführung

abbilden, eine Nachverfolgung der

vereinbarten Ziele enthalten und falls

notwendig, Korrekturmaßnahmen in

der Zusammenarbeit zulassen. Das Ergebnis

des twinning secretariat sollte

es sein, dass das Entwicklungsland

nach der vereinbarten Zeit des twinnings

die Fähigkeit aufgebaut hat, die

Sekretariatsführung selbstständig und

entsprechend den ISO- Regeln korrekt

durchzuführen und hauptverantwortlich

zu übernehmen.

Für das deutsche Sekretariat hat

dies den Vorteil, dass mit Laufe der

Vertragszeit die administrative Last

weniger wird und es sich verstärkt auf

inhaltliche Aspekte und die Qualität

der entstehenden Normen konzentrieren

kann. Weiterhin ist dies ein

hervorragendes Mittel zur Stärkung

einer strategischen Partnerschaft mit

China.

Außerdem erhofft sich Deutschland

durch die gemeinsame Partnerschaft

eine erhöhte Beteiligung asiatischer

Länder in der Erarbeitung von

internationalen Normen im ISO/TC

85/SC 6 sowie deren Akzeptanz und

Anwendung.

Das twinning ist für das ISO/TC

85/SC 6 auf zwei Ebenen vorgesehen,

d. h. es wird neben einer deutschen

Sekretärin (Dipl.-Ing. (FH) Janine

Winkler) einen chinesischen Co-

Sekretär geben und neben dem deutschen

Chairman (Dr. Michael Petri)

eine Vize-Chair aus China. Zeitlich ist

das twinning vorerst auf drei Jahre

beschränkt, wobei eine Verlängerung

um drei Jahre optional ist.

4 Erarbeitung von

internationalen Normen

DIN vertritt die Normungsinteressen

Deutschlands in der International

Organization for Standardization

(ISO). Dabei ist jedem internationalen

Normungsgremium ein nationales

Spiegelgremium bei DIN zugeordnet.

Dieses nationale Spiegelgremium

entscheidet über die aktive Mitarbeit

auf internationaler Ebene, eruiert

die deutsche Meinung und entsendet

Experten in die weltweit tagenden

internationalen Gremien, um die

nationale Position zu vertreten

oder – bei Projekten von besonderem

nationalen Interesse – die ISO-

Projektleitung zu übernehmen. Die

ENERGY POLICY, ECONOMY AND LAW 395

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ENERGY POLICY, ECONOMY AND LAW 396

| | Abb. 5.

Delegationsprinzip.

nationalen Spiegelgremien entscheiden

zusätzlich über die Übernahme

internationaler Normen in das nationale

Normenwerk, welche freiwillig

ist.

Die Erarbeitung der internationalen

Normen vollzieht sich in den

Technischen Komitees und deren

Unterkomitees und Arbeitsgruppen.

Alle ISO-Mitglieder haben das Recht,

in jedem beliebigen Technischen

Komitee (TC) oder Unterkomitee (SC)

ihrer Organisation mitzuarbeiten

( aktive Mitarbeit: P-Mitglieder, Mitwirkung

als Beobachter: O – Mitglieder).

Die Arbeit ist dezentralisiert; die

Sekretariate dieser Gremien werden

jeweils durch ein Mitglied betreut. Die

Zentralsekretariate der ISO in Genf

sind für folgende Aufgaben zuständig:

Allgemeine Verwaltung, Unterstützung

bei der Planung, Koordinierung

der Facharbeit, Durchführung der

Umfrage- und Annahmeverfahren

und Veröffentlichung von internationalen

Normen und anderen Publikationen,

Betreuung der Organe außer

der Technischen Komitees.

Vorschläge für die Erarbeitung

internationaler Normen (oder auch

deren Überarbeitung oder Änderung)

können von verschiedenen Seiten

eingebracht werden. Erhält der Vorschlag

ausreichende Unterstützung,

wird er in das Arbeitsprogramm des

TC oder SC aufgenommen. Zur Aufnahme

ist eine einfache Mehrheit der

P Mitglieder notwendig sowie eine

Verpflichtungserklärung zur aktiven

Mitarbeit von mindestens fünf – dem

Vorschlag zustimmenden – P-Mitgliedern.

Damit ist die Vorschlagsstufe

(Proposal Stage) abgeschlossen.

Ein TC oder SC kann auch vorläufige

Norm-Projekte (Preliminary Work

Items) in sein Arbeitsprogramm aufnehmen,

die für eine Bearbeitung

noch nicht reif sind. Sie werden auf

einer Bereitschaftsstufe (Preliminary

Stage) gehalten und vom Komitee

regelmäßig auf ihre Bearbeitbarkeit

hin überprüft.

Die nächste Bearbeitungsstufe

(Preparatory Stage) umfasst die

Ausarbeitung eines Arbeitsentwurfes

(Working Draft – WD). Dies geschieht

normalerweise auf Arbeitsgruppenebene.

Meist sind mehrere aufeinanderfolgende

WD erforderlich, bis ein

weitestgehend stabiles Arbeitsergebnis

als Entwurfsvorschlag (Committee

Draft – CD) registriert werden und in

die nachfolgende Komiteestufe (Committee

Stage) gehen kann.

Auf der Komiteestufe wird das

Dokument dem TC oder SC schriftlich

zur Kommentierung unterbreitet. Die

zuständige Arbeitsgruppe (Working

Group – WG) sichtet die eingegangenen

Kommentare und erarbeitet

einen Vorschlag für das weitere Vorgehen.

Der Beschluss zur Einreichung

als internationaler Norm-Entwurf ist

im Konsens zu fassen (Konsens gilt als

erreicht, wenn keine Einwände gegen

wesentliche Teile des Dokumentes

aufrechterhalten werden; er bedingt

nicht Einstimmigkeit).

In der Umfragestufe (Enquiry

Stage) wird der Internationale Norm-

Entwurf (Draft International Standard

– DIS) allen Mitgliedern von ISO

zur Prüfung und Abstimmung (Ja,

Nein, oder Enthaltung) innerhalb von

fünf Monaten vorgelegt. Fachliche

Kommentare können eingereicht

werden; ihre Annahme darf bei einer

Ja-Stimme aber nicht zur Bedingung

gemacht werden. Ist der DIS für ein

Mitglied in der vorliegenden Form

nicht annehmbar, muss dieses deshalb

mit Nein stimmen und die fachlichen

Gründe dafür angeben. Die Annahme

| | Abb. 6.

Entstehung einer internationalen Norm.

des DIS erfordert eine Zwei-Drittel-

Mehrheit der P-Mitglieder des zuständigen

TC oder SC und zugleich

eine Drei-Viertel-Mehrheit aller abgegebenen

Stimmen (ohne Enthaltungen).

Nach Ablauf der Umfrage

trifft der Vorsitzende des TC oder SC

in Zusammenarbeit mit dem Sekretär

und in Konsultation mit dem Zentralsekretariat

die Entscheidung über das

weitere Vorgehen und das Sekretariat

erstellt einen Bericht über das Umfrageergebnis

und die eingegangenen

Kommentare. Es muss dabei versucht

werden, auch die mit den Nein-

Stimmen verbundenen Probleme zu

lösen. Im Falle eines positiven Ergebnisses

erarbeitet das Sekretariat mit

der Unterstützung der zuständigen

Arbeitsgruppe die endgültige Fassung.

Für den Fall, dass keine einzige

Nein-Stimme vorliegt, führt die

Umfragestufe direkt zur Veröffentlichungsstufe

(Publication Stage), ansonsten

endet die Umfragestufe mit

der Registrierung des Dokumentes

als Internationaler Schluss-Entwurf

durch das Zentralsekretariat.

In der Annahmestufe (Approval

Stage) wird der Internationale

Schluss- Entwurf (Final Draft International

Standard – FDIS) allen Mitgliedern

von ISO zur Abstimmung

innerhalb von zwei Monaten unterbreitet.

In der Annahmestufe kann der

Schluss-Entwurf sachlich nicht mehr

geändert, sondern nur noch angenommen

oder – mit entsprechender

Begründung – abgelehnt werden.

Die Veröffentlichungsstufe (Publication

Stage) umfasst die abschließende

Korrektur, Veröffentlichung

und Verteilung der Norm durch das

Zentralsekretariat. Die internationale

Norm wird regelmäßig überprüft

(Systematic Review – SR), wobei über

ihren Fortbestand, eine Überarbeitung

oder auch die Zurückziehung

entschieden wird. Bei der ISO erfolgt

dies spätestens nach jeweils fünf

Energy Policy, Economy and Law

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Jahren. Falls die Überprüfung (SR)

ergibt, dass die Norm aktualisiert

werden muss, folgen die Schritte zur

vollständigen Überarbeitung dem vorangehend

geschilderten Arbeitsablauf

für eine neue Norm.

5 Schlussfolgerung

Trotz der politischen Entscheidung

Deutschlands aus der Kerntechnik

auszuscheiden, wird Deutschland

auf Initiative des BMU ein weiteres

internationales Normungsgremium

führend übernehmen. Damit soll auch

in Zukunft ein hohes Niveau der

nuklearen Sicherheit bei der friedlichen

Nutzung der Kernenergie in der

Welt unterstützt und sichergestellt

werden. Im Laufe der nächsten Jahre

wird das BMU – über die Sekretariatsführung

hinaus – die Mitarbeit von

deutschen Experten an internationalen

Projekten unterstützen. Neben

der Fortführung der Regelwerksarbeit

im KTA werden damit auch die dazugehörigen

nationalen und internationalen

Normungsarbeiten unterstützt.

Literatur

[1] Atomgesetz in der Fassung der

Bekanntmachung vom 15. Juli 1985

(BGBl. I S. 1565), das zuletzt durch

Artikel 2 Absatz 2 des Gesetzes vom

20. Juli 2017 (BGBl. I S. 2808) geändert

worden ist

[2] SStrahlenschutzgesetz vom 27. Juni

2017 (BGBl. S. 1966), das durch Artikel

2 G des Gesetzes vom 27. Juni 2017

(BGBl. S. 1966) geändert worden ist.

https://goo.gl/qhvNr5

[3] DIN 820-4:2013-06, Normungsarbeit –

Teil 4: Geschäftsgang

[4] KTA 3201.4, Komponenten des Primärkreises

von Leichtwasserreaktoren Teil

4: Wiederkehrende Prüfungen und

Betriebsüberwachung in der Fassung

6/99

[5] DER KERNTECHNISCHE AUSSCHUSS -

Grundlagen und Verfahren – KTA-

GS-63, Anhang E Vereinbarung

zwischen dem KTA und dem DIN

www.kta-gs.de

Authors

Janine Winkler

Dr. Michael Petri

Senior-Projektmanagerin

DIN-Normenausschuss

Materialprüfung (NMP)

Burggrafenstraße 6

10787 Berlin

397

ENVIRONMENT AND SAFETY

Thermal Hydraulic Analysis of the

Convective Heat Transfer of an

Air-cooled BWR Spent Fuel Assembly

Christine Partmann, Christoph Schuster and Antonio Hurtado

Since the reactor accident in Fukushima Daiichi, the vulnerability of spent fuel pools (SFP) is more focused in nuclear

safety research. In case of a structural damage of the SFP through an external event with a loss of coolant, the coolability

of the spent nuclear fuel is endangered. If the pool is completely drained, the fuel assemblies (FA) are fully uncovered

and only cooled by air. A sufficient decay heat transfer depends on the arising air mass flow from the containment

through the spent FA from bottom to top (chimney effect). Beside analytical approximations from the U.S. Nuclear

Regulatory Commission (NRC) that are partially not publicly accessible only little experimental data about loss of

coolant accidents in SFP exist.

Revised version of a

paper presented at

the Annual Meeting

of Nuclear Technology

(AMNT 2018), Berlin,

Germany.

This paper presents the experimental

findings about the convective heat

transfer of a boiling water reactor

(BWR) spent FA under the absence of

water. These studies are performed

within the joint project SINABEL that

is funded by the German Federal

Ministry of Education and Research to

investigate the thermal hydraulics of

selected accident scenarios in SFP

experimentally and numerically.

For the experimental investigation,

the test facility ALADIN was build up

in 2016. This mock-up is a full-scale

electrically heated 10x10 rod bundle

surrounded by additional heated rods

to simulate the surrounding and to

adjust nearly adiabatic boundary conditions.

The rod bundle is equipped

with a unique 3-D net of thermocouples

that allows the observation of

the cladding temperature distribution

and temperature development combined

with filling level measurement

and video monitoring. Different experiments

at single rod powers

between (20-100) W were conducted.

The results show that the convective

heat transfer of a BWR FA that

is only cooled by air is strongly inhibited.

Inside the FA channel where

axial flow is not appreciable, the entire

heat has to be transferred in radial

direction requiring large temperature

differences.

1 Introduction

Spent nuclear fuel assemblies (FA)

has to be stored in spent fuel pools

(SFP) after their last operation cycle

in the reactor. The decay heat arising

from the decay of the fission products

has to be removed from the SFP by

active systems during the long-term

storage. In case of a system failure

or a loss of integrity of the SFP, a

boil-off or a partial drain-down

scenario can be postulated. Under

these conditions the FA get exposed

to the air of the containment. The conservative

approach is a single FA in the

middle of the SFP surrounded by other

FA of the same temperature. Heat

can be dissipated only in axial

direction. Crossflows are not possible

because of the design of a boiling

water reactor (BWR) FA especially in

case of the application of high-density

racks. The temperature limit up to

which the integrity of the cladding

is guaranteed depends on the atmosphere,

the cladding material, individual

cladding characteristics and the

accident sequences [1, 2].

In publications, a complete loss of

coolant with a free bottom nozzle can

be found as the ultimate benchmark of

the coolability of FA. Partial draindown

scenarios with a blockade at the

bottom play a minor role [3].

The phenomenology and quantity

of arising flow conditions that occur

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ENVIRONMENT AND SAFETY 398

parameter test facility ALADIN SFP (real plant)

pressure 0.1 MPa 0.1 MPa

maximum power 350 W/rod 200 W/rod

(2 d after reactor shutdown)

heated length

of the rods

inside and around the FA during

air-cooling conditions are focus of

this paper. Therefore, experimental

investigations are conducted at the

test facility ALADIN at Technische

Universität Dresden (TUD) within

the joint project SINABEL. The test

facility ALADIN simulates a generic

full scale BWR FA in a SFP under

accident conditions. The experiments

serves for a better understanding of

the heat transport phenomena in

a FA and their dependencies during

air-cooling conditions.

2 Experimental

investigation

For about 30 years, there have

been investigations at the Chair of

Hydrogen and Nuclear Energy, TUD

to study the safety of BWR. Since

2007, the safety of BWR FA inside

SFP is more focused. Vattenfall

Europe Nuclear Energy GmbH initiated

first experiments that were carried

out by TUD at the test facilities

ADELA I (2007-2010) and ADELA II

3,600 mm 3,760 mm

(depending on the manufacturer)

cladding material stainless steel Zircaloy

full height

outer dimensions

4,748 mm

350 mm x 350 mm

(plus 70 mm insulation)

| | Tab. 1.

Technical data of the test facility ALADIN in comparison to a SFP (real plant).

(2010-2013). Beside boil-off experiments,

air- cooling experiments were

conducted.

The test facility ADELA-I simulated

a part of a BWR FA with a 3x3

elec trically heated rod bundle with

one additional heater. The experiments

showed great dependence of

the axial temperature profile on the

entering airflow in the upper region

of the rod bundle. Complex flow

conditions could not be studied in

detail [4].

The enhanced test facility ADELA-II

investigated a quarter of a BWR FA.

It consisted of a 5x5 rod bundle

with 8 additional heaters. The experiments

showed the influence of the

convection flows inside the test facility

in dependence of the depth of the test

facility. Significant cooling effects of

ambient air at the top worsening to

the bottom were observed. Consequently,

heat was transferred in radial

direction to the surrounding by

conduction leading to specific axial

heating profiles [5].

2.1 Test facility ALADIN

The test facility ALADIN was designed

to simulate a full FA inside a SFP by

taking the heat transfer mechanisms

with the surrounding into account.

Therefore, the geometric boundary

conditions of a generic BWR FA were

adopted in nearly original scale.

Experiments were conducted under

atmospheric pressure. Plant conditions

shortly after reactor shutdown

until years afterwards can be simulated

with a continuously adjustable

power up to 350 W per rod. The main

technical data is listed in Table 1.

Main component is a central electrically

heated rod bundle consisting

of 96 heating rods in a square arrangement

positioned in an inner channel.

In the outer channel 44 additional

heating rods represent neighboring

FA in the SFP. All channels are connected

hydraulically to each other

at the bottom and top. The outer

surface of the test facility is thermally

shielded by a microporous insulation

with a total width of 70 mm. Simplified

views of the test facility are given

in Figure 1 and Figure 2.

The power supply of the heating

rods is ensured via 13 individually adjustable

power supply units. Hence, it

is possible to create different radial

power profiles. In recent experiments,

a constant profile was chosen due to

low importance of a specific radial

profile. The embossed axial power

distribution of every rod takes the

burn-up of the fuel rods into account

(Figure 3).

To understand the thermal

hydraulic behavior inside the test

facility during the experiments,

| | Fig. 1.

Simplified sectional view of the test facility ALADIN.

| | Fig. 2.

Simplified longitudinal view of the test facility

ALADIN and axial measuring planes.

Environment and Safety

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measuring

plane (level)

position from bottom

to top in mm

| | Tab. 2.

Arrangement of the instrumentation of ALADIN.

cladding and wall surface temperatures

are measured with 216 thermocouples

on 12 different elevations

thereof 125 in the rod bundle on 10

elevations. Inside the test facility,

thermocouples with a diameter of 0.5

mm are used for minimal- intrusiveness

and good temporal resolution.

On the outside, the use of thermocouples

with a diameter of 1 mm is

sufficient. The probe tips were fixed

under 0.05 mm point welded sheets of

metal. The arrangement of the instrumentation

is listed in Table 2.

The temperature signals were

wired with two solid-state multi plexer

cards with cold junction compensation

and a 26-bit resolution digital

multimeter. Due to the large amount

of measuring points, the time step for

a complete measurement is 6.5 s. The

chosen scan interval is 10 s. The maximum

error including the uncertainty

of the thermocouple is about ±2.5 K.

2.2 Experimental set-up

The heat transfer mechanisms of an

air-cooled FA half a year after reactor

shutdown are investigated by a power

supply of 20 W per rod. Two experiments

are presented:

• Experiment I (exp. 1) simulates a

completely drained SFP. For this

purpose, the test facility is completely

drained.

• Experiment II (exp. 2) represents

the accident scenario of a partialdrain

down with a minimum

blockade at the bottom. Therefore,

the hydraulic connection between

the inner and the outer channel at

the bottom of the test facility is

locked by a water level of 280 mm.

In both scenarios, the test facility is

open at the top and connected to

T cladding

1

T wall_inside 2 T wall_outside

3

1 384 x x x

2 784 x x x

3 1,184 x x x

4 1,584 x x x

5 1,984 x x x

6 2,384 x x x

7 2,784 x x x

8 3,184 x x x

9 3,584 x x x

10 3,984 x x x

11 4,284 x x

12 4,584 x

ambient air. The experiments are

limited to maximum cladding temperatures

of 450 °C due to maximum

operation temperatures of the heating

rods.

3 Results and discussion

3.1 Results of experiment I

In experiment I, the rod bundle heats

up to a maximum cladding temperature

of 427 °C measured at the elevation

of 1,984 mm on measuring plane

5. The experiment was stopped

after 13.5 h hours with a maximum

cladding temperature increase of at

least 3 K/h.

In Figure 4, the axial temperature

profiles of the hottest rod inside the

rod bundle are presented for different

test durations. The dashed line represents

the power profile of a single rod.

The profiles illustrate the heat up of

the cladding and the similarity to the

linear power profile. The comparison

of the temperature profiles with

regard to the power profile shows a

continuously increasing warming in

the lower half of the test facility by

| | Fig. 4.

Axial temperature profiles at different test durations at 20 W/rod (exp. I).

| | Fig. 3.

Axial power distribution of a single rod to simulate the burn-up.

contrast with the upper part. That

indicates a formation of a flow stagnation

region in that area of the rod

bundle. In the upper part, a better

heat dissipation due to convection is

obvious. The air flowing upwards is in

interaction with the air flowing

downwards.

The radial cladding temperature

distributions of the quarter at the

elevation of 2,784 mm (measuring

plane 7) and 3,184 mm (measuring

plane 8) at the end of the experiment

are pictured in Figure 5. Each rod

is instrumented with at least one

thermocouple at each plane whereby

the average temperature distribution

is shown. Both temperature distributions

are non-uniform with a

temperature maximum towards the

center of the test facility. The maximum

temperature difference between

the hottest and coldest rod is about

82 K at level 7 and about 73 K at level

8. It can be concluded that the radial

heat transport is very low and therefore

the coolabilty of the inner rods is

not sufficiently secured.

In Figure 6, axial temperature

differences of the hottest and the

coldest rod inside the bundle after 6 h,

9 h and 13.5 h are compared. The

apparent constant differences per

elevation indicate a constant radial

heat transport inside the rod bundle at

the analyzed test times.

1) surface of the rods

2) surface of the inner

channel inside the

test facility

3) outer surface of the

test facility (plus

insulation)

ENVIRONMENT AND SAFETY 399

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ENVIRONMENT AND SAFETY 400

| | Fig. 5.

Radial cladding temperature distribution after 13.5 h (exp. I).

| | Fig. 6.

Axial temperature differences of the hottest and coldest rod after 6 h, 9 h and 13.5 h at 20 W/rod (exp. I).

3.2 Comparison of the

experimental results

of experiment I with

those of experiment II

For a better understanding of the heat

transfer mechanisms, the experiment

I with the opened bottom connection

is compared to experiment II with the

flow blockade. The cladding temperatures

of the hottest rod inside the rod

bundle and the temperatures of the

outer surface of the inner channel

were analyzed in Figure 7. The rod

bundle is shorter than the inner channel

and ends at a height of 4,035 mm.

Hence, there is no measurement

point above this elevation for the rod

bundle. In the lower and in the upper

region, differences between the temperature

profiles of both experiments

can be seen. In experiment I, the rod

temperatures in the region from

bottom to 2,500 mm are lower compared

to the rod temperatures of

experiment II. Conversely, the temperatures

in the upper part are higher

in experiment I compared to experiment

II. This is evidence for a chimney

effect inside the bundle. Air heated

inside the bundle flows upward, and

colder air of the outer channel is

forced to flow through the bottom

connection inside the rod bundle. In

experiment II, a stagnation region is

formed in the lower part. Ambient air

could only interact in the upper and

middle region of the facility.

4 Conclusion and outlook

In the special case of a complete or a

partial drain-down scenario, decay

heat has to be removed from the FA by

air-cooling. The flow characteristics

and temperature progression in the

assemblies are not investigated in

detail especially not experimentally.

Therefore, experimental tests are

under way within the joint project

SINABEL. Experiments with a rod

power half a year after shutdown are

presented. The investigation shows

that there are huge radial temperature

differences inside the bundle itself.

This indicates, that a sufficient heat

transfer inside a BWR FA is not possible

although the cooling outside is high

enough. Heat can only be transferred

in axial direction. In this event, a

blockade at the bottom of a FA has an

influence on the convection flows inside

the rod bundle. However, these

convection flows only lead to opposite

temperature changes in the lower and

upper part of the rod bundle and has

no significant effect on the maximum

temperature in the investigated range

up to cladding temperatures of 450 °C.

The analysis of the heat transfer

mechanism of BWR FA shortly after

reactor shutdown up to half a year

are under way and will give further

information of these effects and their

dependencies. A grid sensor for combined

temperature and flow velocity

measurement will be inserted in a

quarter of the rod bundle to detect the

flow velocities and their changes over

the test time and at different rod

powers. By modelling the entire FA

structure in original scale with

additional heaters to adjust nearly

adiabatic boundary conditions it will

be possible to make improved predictions

about the cladding temperature

development of a FA in case of

air-cooling conditions. In combination

with numerical investigations with

CFD methods and integral codes, a

significant improvement in this field

of research can be expected. For this

purpose, further publications with

additional insights will follow.

| | Fig. 7.

Axial temperature profiles of the “inner channel” and “rod” after 13.5 h and axial temperature differences (exp. I/II).

Acknowledgement

This work is part of the research

project “SINABEL” and is sponsored

by the German Federal Ministry of

Education and Research (BMBF) under

the contract number 02NUK027A.

Responsibility for the content of this

paper lies with the authors.

Environment and Safety

Thermal Hydraulic Analysis of the Convective Heat Transfer of an Air-cooled BWR Spent Fuel Assembly ı Christine Partmann, Christoph Schuster and Antonio Hurtado


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References

[1] Smith, C.W.: Calculated fuel perforation

temperatures: commercial power reactor

fuels, NEDO-10093, September 1969.

[2] Nourbakhsh, H.P. et al.: Analysis of

spent fuel heatup following loss of

water in a spent fuel pool, NUREG/

CR-6441, Upton: Brookhaven National

Laboratory, March 2002.

[3] Benjamin, A. S. et al.: Spent fuel heatup

following loss of water during storage,

NUREG/CR–0649, US Nuclear

Regulatory Commission, Washington

DC, 1979.

[4] Schuster, C. et al.: Experimental investigation

of the rod load in an evaporating

spent fuel pool, Proceedings on Annual

Meeting on Nuclear Technology, Berlin,

Germany, 2007.

[5] Schulz, S.; Schuster, C.; Hurtado, A.:

Convective heat transfer in a semiclosed

BWR-fuel assembly in absence of

water, Nuclear Engineering and Design,

Volume 272, 2014, Pages 36-44, ISSIV

0029-5493

Authors

Further Development of a Thermal-

Hydraulics Two-Phase Flow Tool

Verónica Jáuregui Chávez, Uwe Imke, Javier Jiménez and V.H. Sánchez-Espinoza

Dipl.-Ing. Christine Partmann

Dr.-Ing. Christoph Schuster

Prof. Dr.-Ing. habil. Antonio Hurtado

Technische Universität Dresden

Institute of Power Engineering

Chair of Hydrogen and Nuclear

Energy

01062 Dresden, Germany

401

OPERATION AND NEW BUILD

The numerical simulation tool TWOPORFLOW is under development at the Institute for Neutron Physics and Reactor

Technology (INR) of the Karlsruhe Institute of Technology (KIT). TWOPORFLOW is a thermal-hydraulics code that is

able to simulate single- and two-phase flow in a structured or unstructured porous medium using a flexible 3-D Cartesian

geometry. It has the capability to simulate simple 1-D geometries (like heated pipes), fuel assemblies resolving the

sub-channel flow between rods or a whole nuclear core using a coarse mesh. The code uses six conservation equations

in order to describe the coupled flow of steam and liquid. Several closure correlations are implemented to model the

heat transfer between solid and coolant, phase change, wall friction as well as the liquid-vapor momentum coupling.

Originally, TWOPORFLOW was used to calculate the flow and heat transfer in micro-channel heat exchangers. The

main purpose of this work is the extension, improvement and validation of TWOPORFLOW in order to simulate the

thermal-hydraulic behavior of Boiling Water Reactor (BWR) cores. For that aim, the code needs some additional

empirical models. In particular, a turbulent lateral mixing model, and a void drift model have been implemented, tested

and validated, adopting relevant tests found in the literature. Regarding reactor conditions, the BFBT critical power

bundle experiments were selected for the validation.

1 Introduction

TWOPORFLOW is a thermal- hydraulic

code based on a porous media

approach to simulate single and twophase

flow in 3D Cartesian coordinates.

The time dependent mass,

momentum and energy conservation

equations for each fluid are solved

with a semi-implicit continuous

Eulerian type method. TWOPOR-

FLOW was originally developed for

the simulation of thermal-hydraulic

phenomena inside micro-channels [1]

[2]. However, the code has been

recently modernized and adapted to

be able to describe thermal-hydraulic

phenomena occurring in Light Water

Reactors (LWRs), specifically BWRs

[3].

2 TWOPORFLOW

capabilities and

main features

TWOPORFLOW is capable to solve

transient or steady state problems in

reactor cores or RPV with a flexible

3D Cartesian geometry which can be

used to represent sub-channels, fuel

assemblies, or even the whole core.

The rod centered and the coolant centered

approaches are available for

sub- channel simulations. TWOPOR-

FLOW uses a system of six conservation

equations. The mass conservation

equations of the two phases are given

by the following equations:



(1)

(2)

(3)

The source term Γ I describes the rate

of evaporation or condensation at the

liquid-vapor interface.

The momentum equations are used

in non-conservative form as follows:


(4)


(5)

For energy conservation equations,

the internal energy (e) is used as the

main variable:

(6)

(7)

TWOPORFLOW has additional models

to close the system of conservation

equations, like solid-coolant heat

transfer, interphase heat exchange,

empirical correlations for wall

friction, empirical correlations for

Operation and New Build

Further Development of a Thermal- Hydraulics Two-Phase Flow Tool ı Verónica Jáuregui Chávez, Uwe Imke, Javier Jiménez and V.H. Sánchez-Espinoza


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OPERATION AND NEW BUILD 402

interphase friction and liquid-vapor

momentum coupling. However, some

models need to be added or improved.

For example, turbulent lateral mixing,

void drift and critical heat flux (CHF)

as well as the post-CHF models. In the

next sections the addition of turbulent

viscosity, void dispersion and turbulent

conductivity, as well as the results

of the validation of these models are

presented.

3 Improvement of physical

models

3.1 Turbulent viscosity

To describe the effect of the turbulent

flow between sub-channels in the

momentum equations a simple algebraic

equation approach is chosen.

According to this approximation, the

turbulent flow can be simulated as a

pseudo fluid having an effective

viscosity (μ), which is the result from

the addition of the molecular and the

turbulent viscosities. This extension is

based on a mixing coefficient (β) which

was determined experi mentally [4].

Such simple model does not account

for the details of turbulence, but it

describes the general mixing behavior

between sub-channels leading to the

following equation for total viscosity.

μ = μ mol + μ tur (8)

μ tur = βρVl (9)

eddy diffusivity. In this work the value

of 0.9 is used [7].


(12)

The turbulent conductivity is added to

the thermal conductivity of the fluid

and affects directly the conductivity

terms in equations (6) and (7).

4 Validation

(13)

4.1 NUPEC PSBT stationary

temperature tests (thermal

mixing)

To validate the implementation of the

turbulent-viscosity and conductivity,

nine tests of the Exercise 1 Phase II

“Steady State Fluid Temperature”

from the NUPEC PSBT benchmark [8]

have been used. The boundary conditions

of the tests are:

• Outlet pressure: 4.92 to 16.58 MPa

• Inlet mass flow: 1.3 to 11.52 kg/s

• Inlet temperature : 86 to 289.2 °C

• Bundle power: 0.4 to 3.44 MW

The quoted measurement error for the

outlet temperatures is 1°C [8].

The tests consist of a 5x5 rod

assembly with constant axial power

distribution. The PSBT benchmark

uses the rod power map shown in

Figure 1.

1.00 1.00 0.25 0.25 0.25

1.00 1.00 1.00 0.25 0.25

1.00 1.00 0.25 0.25 0.25

1.00 1.00 1.00 0.25 0.25

1.00 1.00 0.25 0.25 0.25

| | Fig. 1.

Lateral power distribution PSBT tests.

The meshing in TWOPORFLOW is

constructed by a coolant centered

sub-channel approach, resulting in an

arrangement of 6x6 sub-channels in

directions X and Y respectively; and

27 axial cells in Z direction. The

number of rods per channel is ¼, ½,

or 1 depending on the location

of the sub-channel as can be seen in

Figure 2.

Six different mixing coefficients

are tested for the validation, 0.03,

0.04, 0.05, 0.06, 0.07 and 0.08.

Figure 3 shows the difference

between the average-calculated and –

measured temperatures at the top of

the sub-channels dependent on the

mixing coefficients. With no mixing,

most of the tem peratures are outside

the 10 % scattering band, but an

increasing mixing coefficient leads to

temperatures closer to the measured

values. However, starting at a mixing

coefficient of 0.05 the temperatures

disperse again. The minor deviation

is found using coefficients of 0.05

and 0.06.

3.2 Void dispersion

A void dispersion term (pi) is added to

the vapor momentum equation for

bubbly flow and is calculated from an

assessment of the turbulent kinetic

energy using the next equation [5]:


(10)

This implementation affects directly

the equation (5) adding a term, and

thus gives the following equation:

| | Fig. 2.

View from the top of the TWOPORFLOW’s model of the NUPEC PSBT.

(11)

3.3 Turbulent conductivity

To describe the effect of the turbulent

flow between channels in the energy

equation, the turbulent conductivity

between adjacent sub-channels is

calculated using the turbulent Prandtl

number [6]. This number is defined as

the ratio between the momentum

eddy diffusivity and the heat transfer

| | Fig. 3.

Difference between measured and calculated temperatures dependent on mixing coefficient.

Operation and New Build

Further Development of a Thermal- Hydraulics Two-Phase Flow Tool ı Verónica Jáuregui Chávez, Uwe Imke, Javier Jiménez and V.H. Sánchez-Espinoza


atw Vol. 63 (2018) | Issue 6/7 ı June/July

| | Fig. 4.

Geometry of test Assemblies 1, 01, 02, 03 and 04.

1.15 1.30 1.15 1.30 1.30 1.15 1.30 1.15

1.30 0.45 0.89 0.89 0.89 0.45 1.15 1.30

1.15 0.89 0.89 0.89 0.89 0.89 0.45 1.15

1.30 0.89 0.89 0.89 0.89 0.89 1.15

1.30 0.89 0.89 0.89 0.89 0.89 1.15

1.15 0.45 0.89 0.89 0.89 0.89 0.45 1.15

1.30 1.15 0.45 0.89 0.89 0.45 1.15 1.30

1.15 1.30 1.15 1.15 1.15 1.15 1.30 1.15

A)

1.15 1.30 1.15 1.30 1.30 1.15 1.30 1.15

1.30 0.45 0.89 0.89 0.89 0.45 1.15 1.30

1.15 0.89 0.89 0.89 0.89 0.89 0.45 1.15

1.30 0.89 0.89 0.89 1.15 1.15

1.30 0.89 0.89 0.89 1.15 1.15

1.15 0.45 0.89 0.89 0.89 0.89 0.45 1.15

1.30 1.15 0.45 0.89 0.89 0.45 1.15 1.30

1.15 1.30 1.15 1.15 1.15 1.15 1.30 1.15

B)

OPERATION AND NEW BUILD 403

| | Fig. 5.

Lateral power distribution BFBT.

4.2 NUPEC BFBT stationary

void fraction tests (void

drift)

Fifteen tests of the Exercise 1 Phase I

“steady-state sub-channel grade

benchmark” from the BWR Full- size

Fine-mesh Bundle Test (BFBT)

Benchmark [9] were used to validate

the implementation of the void

dispersion. The tests have a geometry

of 8x8 pin assembly , different lateral

power distributions, (uniform for

assembly 1; Figure 5-A for assemblies

01, 02, 03; and Figure 5-B for

assembly 4), and different axial power

distributions (constant for assemblies

01, 02, 03, and 4; and cosine for

assembly 1).

The boundary conditions of the

tests are:

• Outlet pressure: ~7.15 MPa

• Inlet mass flow: ~15.20 kg/s

• Inlet temperature: ~278 °C

• Bundle power: 1.9 – 6.48 MW

The error in the void measurement

is given as 3% [9].The tests have

been modeled in TWOPORFLOW

using a coolant centered sub-channel

approach, making an arrangement

of 9x9 sub-channels and 24 axial

cells. A small mixing coefficient of

0.007 is set, because the assemblies

do not have mixing vane spacers,

as used in PSBT. The number of rods

per channel is ¼, ½, or 1 depending

on the location of the sub-channel

(Figure 2).

The calculations were run with an

old version of TWOPORFLOW without

void drift, and with the new model.

The average percentage error in void

fraction per assembly of both simulations

with respect to the experimental

data shows a better approximation

using the version of TWOPORFLOW

with void drift (Figure 6).

5 Conclusions and

outlook

The validation results obtained for the

improved TWOPORFLOW code have

shown that the code is capable to

simulate in an appropriate way the

most important thermos-hydraulic

phenomena occurring in a BWR or in

similar conditions.

The next step is to improve and

validate critical heat flux (CHF),

| | Fig. 6.

Average % error of simulations with- and without void dispersion term (pi).

transition boiling, and subcooled boiling

correlations of TWOPORFLOW.

In addition, post-CHF models like

minimum film boiling temperature,

annular film dry out, rewetting, and

cool down of a superheated surface

is needed in order to simulate the

physical phenomena that may happen

during accidental conditions in a BWR

core.

Acknowledgments

This work has been performed at

the Institute for Neutron Physics and

Reactor Technology (INR) of the Karlsruhe

Institute of Technology (KIT). The

authors would like to thank the

Program Nuclear Safety Research of

KIT for the financial support of the

research topic “multi-physics methods

for LWR”.

Operation and New Build

Further Development of a Thermal- Hydraulics Two-Phase Flow Tool ı Verónica Jáuregui Chávez, Uwe Imke, Javier Jiménez and V.H. Sánchez-Espinoza


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404

DECOMMISSIONING AND WASTE MANAGEMENT

Nomenclature

C p

D H

e


FI

Specific Heat (J/kg-K)

Hydraulic diameter (m)

Internal energy (J/kg)

Friction at vapor-liquid interface

(N/m 3 )


Fwk Wall friction for phase k (N/m 3 )

→ g Gravity (kg/m-s 2 )

l

P

pi

Characteristic mixing length (m)

Pressure (Pa)

Void dispersion term (Pa/m)

Pr tur Turbulent Prandtl number (0.9)

Q H

Q I

Q w

Internal heat source in porous structure

(W/m 3 )

Heat exchange between phases

(W/m 3 )

Heat exchange between structure

and fluid (W/m 3 )

t Time (s)

→ V, V Velocity of fluid (m/s)

Greek letters

α

β

Γ I

λ

λ k

λ tur

Volume fraction of vapor

Mixing coefficient

Rate of evaporation/condensation

(kg/m 3 -s)

Total thermal conductivity (W/m-K)

Turbulent conductivity of the fluid

(W/m-K)

Turbulent conductivity (W/m-K)

μ

Effective viscosity (Pa-s)

μ mol Molecular viscosity (Pa-s)

μ tur

Turbulent viscosity (Pa-s)

ρ Fluid or clad density (kg/m 3 )

φ

Subscripts

L

v

Porosity

Bibliography

Liquid Phase

Vapor phase

[1] U. Imke, “Porous media simplidied

simulation of single- and two-phase

flow heat transfer in micro-channel

heat exchangers,” Chemical Engineering

Journal, pp. 295-302, 2004.

[2] B. Alma, U. Imke, R. Knitter, U. Schygulla

and S. Zimmermann, “Testing and

simulation of ceramic micro heat

exchangers,” Chemical Engineering

Journal, no. 135, pp. 179-184, 2008.

[3] J. Jimenez, N. Trost, U. Imke and V.

Sanchez, “Recent developments in

TWOPORFLOW, a two-phase floew

porous media code for transient

thermo- hydraulic simulations,” in

Annual Meeting on Nuclear

Technology, Frankfurt, Germany, 2014.

[4] F. S. Castellana, W. T. Adams and J. E.

Casterline, “Single-Phase Sub-channel

Mixing in a Simulated Nuclear Fuel

Assembly,” Nuclear Engineering and

Design, no. 26, pp. 242-249, 1974.

[5] M. Valette, “Analysis of Subchannel and

Rod Bundle PSBT Experiments with

CATHARE 3,” Science and Technology of

Nuclear Installations, 2012.

[6] S. J. Kim and S. P. Jang, “Effects of the

Darcy number, the Prandtl number,

and the Reynolds number on local

thermal non-equilibrium,” International

Journal of Heat and Mass Transfer,

no. 45, pp. 3885-3896, 2002.

[7] A. Malhotra and S. Kang, “Turbulent

Prandtl number in circular pipes,”

International Journal of Heat and Mass

Transfer, no. 27, pp. 2158-2161, 1984.

[8] NUPEC, “OECD/NRC Benchmark based

on NUPEC PWR Sub-channel and

Bundle Tests (PSBT) Volume I: Experimental

Database and Final Problem

Specifications,” NUPEC, Japan, 2010.

[9] NUPEC, “NUPEC BWR Full-size

Fine-mesh Bundle Test (BFBT)

Benchmark. Volume I: Specifications,”

Japan, 2006.

Authors

Verónica Jáuregui Chávez

Uwe Imke

Javier Jiménez

V.H. Sánchez-Espinoza

Karlsruhe Institute of Technology

Institute for Neutron Physics and

Reactor Technology

Hermann von Helmholtz Platz 1

76344 Eggenstein-Leopoldshafen,

Germany

Besonderheiten bei Messungen

zur radiologischen Charakterisierung

hochradioaktiver Abfälle

Marina Sokcic-Kostic und Roland Schultheis

1 Einführung Beim Betrieb von Kernkraftwerken fallen gelegentlich hochradioaktive Abfälle an, wie zum

Beispiel Bruchstücke defekter Brennelemente oder Filter von heißen Zellen. In manchen Kraftwerken wurden diese

Abfälle über lange Zeit in unterirdischen Depots gelagert. Diese Depots entsprechen jedoch nicht den Zielsetzungen für

eine sichere Langzeitlagerung, bis die im Abfall enthaltenen radioaktiven Isotope ausreichend zerfallen sind und der

Abfall dann als nicht-radioaktiver Abfall weiterverarbeitet oder entsorgt werden kann.

Für solche Abfälle hat die NUKEM

Technologies Engineering Services Abfallbehandlungsmöglichkeiten

konzipiert

und in Projekten umgesetzt,

die hochradioaktive Abfälle charakteri

sieren und entsprechend den

An forderungen für die Langzeitlagerung

konditionieren. Dies schließt

auch eine Volumenverminderung ein,

um so die künftigen Lagerkosten zu

minimieren.

Der Schwerpunkt dieses Artikels

liegt in der Messung von hochaktivem

Abfall und seinen Impli kationen.

2 Charakterisierung

der Abfälle

Nach der Verfüllung der konditio nierten

Abfälle in geeignete Behälter werden

die Abfälle wie folgt charak terisiert:

a) Dosisleistung an der Behälteroberfläche

b) Kontamination der Behälteroberfläche

c) Spezifische Aktivität des Inhaltes

der Behälter

d) Zeit bis zum Abklingen der Aktivität

(Halbwertszeit der enthaltenen

Isotope)

a) und b) sind wichtig für die

Handhabung und den Transport der

Behälter, c) und d) hingegen für die

Lagerung des Abfalls.

Decommissioning and Waste Management

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Die Grenzwerte für eine Klassifizierung

gemäß a)-d) sind in den

einzelnen Ländern teilweise unterschiedlich.

Typische Werte sind:

a) > 2 mSv/h

b) Wischtestwerte:

Beta-Aktivität >

10 7 Teilchen/cm 2 /min

Alpha-Aktivität >

10 6 Teilchen/cm 2 /min

c) Beta-Aktivität > 3.7*10 9 Bq/kg

Alpha-Aktivität > 3.7*10 8 Bq/kg

d) Der Abfall wird als langlebig

deklariert, wenn er einen signifikanten

Anteil von Isotopen mit

einer Halbwertszeit größer 30

Jahre hat (Ausnahme: Cs-137).

Wie gezeigt, gibt es für hochaktiven

Abfall nur einen unteren Grenzwert.

Bei der Planung von Anlagen und

Messeinrichtungen muss man zunächst

auch obere Grenzwerte

definieren. So müssen die Räume für

die Handhabung eine ausreichende

Abschirmung aufweisen. Das fängt

damit an, dass die Räume für die

Handhabung ausreichende Abschirmung

aufweisen. Aber auch die

Instrumentierung muss so ausgelegt

sein, dass sie bei den zu erwartenden

Aktivitäten noch funktionsfähig ist.

3 Allgemeine Eigenschaften

des hochaktiven

Abfalls

Bevor man die geeignete Messmethode

für den hochaktiven Abfall

auswählt, sollte man zunächst einmal

Herkunft und Eigenschaften des

Abfalls betrachten.

Hochaktive Abfälle können sowohl

fest wie auch flüssig sein. Wenn man

einmal die hochaktiven Abfälle aus

der Wiederaufarbeitung von Brennstäben

ausklammert, so eignen sich

für die Langzeitlagerung nur feste

Stoffe, weswegen flüssige Abfallstoffe

zuvor in feste Formen umgewandelt

werden müssen (z.B. durch Eindampfung

etc.). Flüssige Stoffe sind

für die Langzeitlagerung nicht

geeignet, da hier chemische Reaktionen

nicht ausgeschlossen werden

können, die zum Beispiel Korrosion

der Behälter zur Folge haben können.

Kernbrennstoff ist bei Abfällen aus

Kernkraftwerken im Allgemeinen nur

in sehr kleinen Mengen im Abfall

präsent, da generell Kernbrennstoff

kein Abfall ist. Kleinere Mengen

können zum Beispiel dadurch entstehen,

dass Material von defekten

Brennelementen im Reaktor nach

unten fällt (sogenannter „fuel debris“).

Die Hauptaktivität der hoch aktiven

Abfälle resultiert aus Aktivierungsprozessen

(z.B. Aktivierung von

| | Abb. 1.

Geiger-Müller-Zählrohr mit Energiekompensation.

Stahl) oder aus der Freisetzung von

Cs-137, welches durch mangelnde

Rückhaltung der Kernbrennstäbe in

den Kühlkreislauf gelangt. Letzteres

wird durch Filtrierung zurückgehalten.

Diese Filter können dann

hohe Aktivitäten aufweisen.

Etwas anders verhält es sich, wenn

der Abfall aus Unfällen, wie zum

Beispiel Tschernobyl, stammt. Dann

können sehr wohl auch Isotope wie

Uran oder Plutonium in relevanten

Mengen vorliegen.

Um die im Folgenden beschriebenen

Messmethoden so einfach wie

möglich zu halten, ist folgende Vorgehensweise

empfehlenswert:

• Zunächst ist die Herkunft des

Abfalls zu klären

• Dann sollte auf der Basis von

Probenahmen die zu erwartenden

Isotope in einem Labor bestimmt

werden

• Anschließend sollte der Isotopenvektor

bezogen auf Co-60 und

Cs-137 bestimmt werden, welche

dann als Leitisotope bei der Berechnung

der Aktivitäten heran gezogen

werden können.

4 Messung der Oberflächendosisleistung

bei

hochaktiven Abfällen

Für die Messung der Oberflächendosisleistung

werden in der Regel

Geiger-Müller-Zählrohre eingesetzt,

die aufgrund ihrer geringen Baugröße

auch Hot Spot Erkennung erlauben

(Abbildung 1). Der Abstand zwischen

Zählrohr und Oberfläche ist gewöhnlich

mit 10 cm spezifiziert. Die

Zählrohre sind mittels Filter energiekompensiert

und in Sv/h geeicht.

Ein wichtiger Punkt bei Messungen

an hochaktiven Abfällen ist die

Vermeidung einer Paralyse des Zählrohres.

Diese könnte zu einer Unterschätzung

der Dosisleistung und

ernsthaften Konsequenzen für die an

der Hantierung beteiligten Arbeiter

führen.

Aus diesem Grund werden oft

Geräte mit zwei Zählrohren unterschiedlicher

Empfindlichkeit eingesetzt,

wodurch die Messdynamik des

Gerätes erweitert wird. Weiterhin

arbeiten die Geräte in einem totzeitlosen

Modus um den Messbereich zu

erweitern und Messfehler durch die

ansonsten benötigte Totzeitkorrektur

zu minimieren (Abbildung 2).

Hierbei wird die Hochspannung

für das Messgerät zunächst unterhalb

der Einsatzgrenze eingestellt. Anschließend

wird die Hochspannung

über die Einsatzgrenze erhöht und

die Zeit bis zum Auftreten des ersten

Pulses gemessen. Dann wird die Hochspannung

wieder abgesenkt. Nach

einem fest vorgegebenen Zeitintervall

wird der Vorgang wiederholt. Die so

gemessenen Zeitintervalle sind umso

kleiner, je höher die Zählrate ist. Diese

Messmethode ist bei hohen Zählraten

recht genau, bei kleinen Zählraten

sind die statistischen Fehler etwas

größer, was bei hochaktiven Abfällen

ohne Bedeutung ist.

5 Messung der Oberflächenkontamination

bei hochaktiven Abfällen

Wenn die Behälter mit Abfall außerhalb

von kerntechnischen Anlagen

(z.B. in Zwischen- oder Endlager)

transportiert werden müssen, so muss

zur Verhinderung der Verschleppung

von Radioaktivität die Behälteroberfläche

auf Kontaminationsfreiheit

geprüft werden. Wegen der Eigenstrahlung

derartiger Behälter kann

die Kontamination bezüglich Alphaoder

Betastrahlung nicht direkt

gemessen werden. Daher wird ein

zweistufiges Verfahren angewandt.

| | Abb. 2.

Dosisleistungsmessgerät mit zwei totzeitfrei

arbeitenden Geiger-Müller-Zählrohren

DECOMMISSIONING AND WASTE MANAGEMENT 405

Decommissioning and Waste Management

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DECOMMISSIONING AND WASTE MANAGEMENT 406

Zunächst wird die Behälteroberfläche

mit einem speziellen Papier

abgewischt. Dies kann manuell oder

auch automatisiert mit Kassetten

erfolgen. Eine Variante ist auch die

Nutzung von Rollen, die mit einem

doppelseitig klebenden Band bestückt

sind. Beim Abrollen werden durch

den Kleber locker sitzende Kontaminationen

fixiert.

Anschließend wird das Papier oder

die Rollen in einen Raum mit niedriger

Umgebungsaktivität verbracht, in

welchem ein Kontaminationsmessgerät

aufgebaut ist. Dieses Gerät misst

dann die von der Wischprobe ausgehende

Alpha- und Betastrahlung mit

Hilfe von Szintilationszählern. Um die

Empfindlichkeit für Alphastrahlung zu

erhöhen, sind diese teilweise mit ZnS

Pulver oder Ähnlichem an der Außenseite

bestäubt, welche bei Auftreffen

von Alphastrahlung eine hohe Lichtausbeute

ergeben.

6 Messung der spezifischen

Aktivität bei

hochaktiven Abfällen

Die Messung der Aktivität eines

Gebindes (Abbildung 3) beruht gewöhnlich

auf dem Nachweis der von

den Gebinden emittierten Photonen.

Um die Gesamtzahl von Betazerfällen

zu berechnen, benötigt man den

zugehörigen Isotopenvektor. Dieser

beschreibt nicht nur die pro Betazerfall

emittierten Photonen sondern

auch die verdeckten Zerfälle, wie zum

| | Abb. 3.

Fassmonitor bestückt mit Dosisleistungs messgeräten.

| | Abb. 4.

Spektren eines 3x3inch NaI Spektrometers

(oben) und eines 2x2inch Spektrometers

(unten) eines Co-60 Präparates.

Beispiel Betazerfälle ohne Photonenemission

(z.B. Sr-90).

Für Abfälle aus Kernkraftwerken

ist gewöhnlich bei sehr hohen Aktivitäten

eine Dominanz der Isotope

Cs-137 und Co-60 zu finden. Diese

Isotope können gut gemessen werden

und dienen dann als Leitnuklide.

Aufgrund der unterschiedlichen

Halbwertszeiten ist eine Unterscheidung

zwischen Co-60 und Cs-137

nötig. Daher werden zur Messung

gewöhnlich HPGe Detektoren eingesetzt,

die eine exzellente Trennung

ermöglichen. Durch die getrennte

Messung von Co-60 und Cs-137

ist zugleich auch eine der Energie

der Isotope angepasste Absorptionskorrektur

möglich.

Probleme bei hochaktiven Abfällen

entstehen, wenn der Dynamikbereich

des Detektors nicht mehr mit den

Zählraten mitkommt. Bei HPGe

Detektoren ist die Zählrate auf etwa

60.000 Ereignisse im Spektrum

begrenzt. Dies gilt bei rückgekoppeltem

Vorverstärken. Für Transistor

Reset Schaltungen, bei welchen der

Vorverstärker nicht rückgekoppelt

sondern mit einem Transistorschalter

kurzgeschlossen wird, ist dieser Wert

etwas höher (ca. 80 bis 100.000), aber

auch dieser Wert ist bei hoch aktivem

Abfall schnell erreicht.

Man könnte nun auf die Idee

kommen, den Kristall zu verkleinern

um so die Zählrate abzusenken. Doch

dies ist keine gute Idee, wie die

Abbildung 4 der Spektren eines NaI

Spektrometers zeigen:

Wie die Abbildung 4 zeigt, verringert

sich bei Verkleinerung des

Kristalls zwar die Fläche unter den

beiden Co-60 Linien, gleichzeitig

steigt aber der niederenergetische

Untergrund an, da die Wahrscheinlichkeit,

dass die Photonen vollständig

im Kristall absorbiert werden, ebenfalls

niedriger wird. Den gleichen

Effekt finden wir auch bei HPGe

Detektoren.

Im vorliegenden Fall nimmt die

Peakfläche bei einer Verkleinerung

von 3 auf 2 inch (Durchmesser wie

Länge des Kristalls) zwar auf 37 % ab,

die Gesamtzählrate verringert sich

jedoch nur auf 60 %.

Die beste Methode bei sehr hohen

Zählraten ist die Vergrößerung des

Abstandes zwischen Detektor und

Gebinde. Leider sind bei sehr hohen

Abfallaktivitäten die räumlichen Gegebenheiten

oft sehr eingeschränkt,

sodass diese Methode nicht viel

weiterhilft.

Eine andere Methode ist der Einsatz

von Schlitzkollimatoren. Allerdings

muss die Öffnung groß

genug sein um Vielfachstreueffekte

am Kollimator zu unterbinden.

Wenn die bisher beschriebenen

Methoden nicht weiterhelfen, so muss

man Absorptionsplatten vor den

Detektor stellen. Allerdings muss man

dabei für jede Energie (oder Linie)

eine entsprechende Absorptionskorrektur

durchführen. Diese Methode

wurde bislang mit Bleiplatten bis zu

15 cm Dicke erfolgreich angewandt.

Aber auch diese Methoden hat ihre

Grenzen.

Bei spezifischen Aktivitäten von

10 10 Bq/kg oder höher für Co-60 oder

Cs-137 endet der Einsatz von HPGe

Spektrometern. Hier muss man auf

die Messung der Oberflächendosisleistung

zurückgreifen und aus diesen

Werten die Aktivität berechnen. Dies

gelingt nur, wenn man für die Rechnungen

ein Modell entwickelt hat.

Als Beispiel sei hier ein Container

erwähnt, der von oben mit Abfällen

gefüllt wird, und an dessen Unterseite

sich Dosisleistungsmessgeräte befinden.

Zusätzlich wird das Gewicht des

Containers gemessen. Als radio aktives

Isotop wird Co-60 angenommen.

Die MCNP Rechnungen, die hierfür

durchgeführt wurden, zeigten,

dass man grob die Dosisleistung

als Funktion des Container Netto-

Gewichtes beschreiben kann (siehe

Abbildung 5).

Die eingezeichnete Linie präsentiert

alle Werte in einem Fehlerbereich

von +/-30 %. Bei den Rechnungen

schwankten die Dichten im Bereich

von 0,2 bis 1,1 g/cm 3 . Beim Befüllen

wurde ein Füllkegel mit einem

Decommissioning and Waste Management

Special Features of Measurement for the Radiological Characterization of High-level Radioactive Waste ı Marina Sokcic-Kostic and Roland Schultheis


atw Vol. 63 (2018) | Issue 6/7 ı June/July

| | Abb. 5.

Dargestellt ist der Empfindlichkeitsfaktor S als Funktion des Abfallgewichtes. Die Bedeutung von S ist weiter unten im Text näher erklärt.

Öffnungswinkel von 30 Grad simuliert.

Entsprechend dem Gewicht und

der angenommenen Dichte war der

Füllungsgrad des Containers unterschiedlich

und lag im Bereich von

8 bis 72,4 %.

Die Aktivität ergibt sich aus der

Formel (1):

Aktivität [Bq] =

Dosisleistung [Sv/h] *100 /

Verzweigungsverhältnis / S(1)

S ist der in Abbildung 5 dargestellte

Empfindlichkeitsfaktor

Der Maximalbereich eines Standard-Dosisleistungsmessgerätes

liegt

bei etwa 100 Sv/h. Geht man

von S = 1E-11 aus und setzt das

Ver zweigungsverhältnis auf 2, so

erhält man eine maximal messbare

Aktivität von 5E+14 Bq. Das ergibt

bei einem Gewicht von 2.290 kg

eine spezifische Aktivität von 2E+

11 Bq/kg. Bei einem Gewicht von

100 kg erhält man entsprechend

eine spezifische Aktivität von 7,4E

+11 Bq/kg.

Abschließend muss erwähnt werden,

dass der Raum, in welchem die

Messungen durchgeführt werden, als

Heiße Zelle ausgelegt und natürlich

für jeden Zutritt gesperrt sein muss

(Abbildung 6).

Diese Darstellung gilt nur für

Detektoren die unterhalb des Containers

montiert sind. Die Messwerte

seitlicher Detektoren oder solche auf

dem Deckel des Containers sind hingegen

sehr stark vom Füllungsgrad

abhängig.

7 Klassifikation der

hochaktiven Abfälle als

lang lebiger (long lived)

oder als kurzlebiger

(short lived) Abfall

Die Klassifikation von radioaktivem

Abfall als langlebig (LLW long lived

waste) basiert auf der Existenz von

Radionukliden mit einer Halbwertszeit

größer als 30 Jahre, wobei Cs-137

nicht mitbewertet wird. Als Nebenbedingung

wird gefordert, dass die

Aktivität derartiger Nuklide per

Gebinde im Mittel nicht größer ist

als 400 Bq/g. Die hier aufgeführte

Beschreibung ist beispielhaft und

kann entsprechend lokaler Gesetze

variieren.

Zusammenfassung

Der Umgang und die Messung

von hochaktivem Abfall stellt an

Messmethodik und Instrumentierung

sehr spezielle Anforderungen. Anstelle

der sonst üblichen Parameter

wie Nachweisgrenze und Selektivität

stehen der Dynamikbereich der

Messung und obere Messgrenzen im

Vordergrund.

Auch ist die Durchführbarkeit der

Messungen oft an das Vorhandensein

von ausreichenden Abschirmungen

und der Fernhantierung der Operationen

gebunden.

Das Fernziel der Messungen ist

die Abschätzung von Lagerdauer und

Lagerart bis hin zum Zeitpunkt, an

welchem der Abfall durch Zerfall

seiner radioaktiven Isotope gefahrlos

gehandhabt werden kann.

Authors

Dr. Marina Sokcic-Kostic

Roland Schultheis

NUKEM Technologies Engineering

Services GmbH

Industriestraße 13

63755 Alzenau, Germany

DECOMMISSIONING AND WASTE MANAGEMENT 407

| | Abb. 6.

Zelle für den fernhantierten Umgang mit radioaktiven Stoffen.

Decommissioning and Waste Management

Special Features of Measurement for the Radiological Characterization of High-level Radioactive Waste ı Marina Sokcic-Kostic and Roland Schultheis


atw Vol. 63 (2018) | Issue 6/7 ı June/July

408

Inside

International Youth Nuclear Congress (IYNC) – Women in Nuclear (WiN) Global Conference

KTG INSIDE

11. bis 17. März 2018 Bariloche, Argentinien

… inklusive Ausflug in die Geschichte der Kerntechnik Argentiniens

| | IYNC und WiN Congress in Argentinien: Gruppenfoto der Teilnehmenden.

Argentinien ist das Land des Tangos, der Gauchos, von

Rindfleisch und Wein so das Cliché. Kernreaktoren

gehören da kaum dazu. Als Melina Belinco das 1995 von

Maela Viirso gegründete WiN Chapter in Argentinien

wieder belebte, wollte sie die Atomforschung Argentiniens

international zeigen und so traf es sich gut, dass der International

Youth Nuclear Congress 2018 in Argentinien

geplant war. Die Zusammenarbeit der Frauen mit den

Jungen führte zu einer inspirierenden Konferenz in

Bariloche. Dort ist neben Schokolade der Nuklearsektor

äußerst wichtig. Junge Frauen und Männer werden seit

1955 im Instituto Balseiro ausgebildet. Der Kongress

vermittelte nicht nur viel Wissenswertes über Kerntechnik

heute. Er beleuchtete auch die Entwicklung der „Small

Modular Reaktors“ (SMR) und in der Ausstellung neben

den zahlreichen Postern von Teilnehmenden auch den

Entwicklungsstand der vier Reaktoren in Barakah (VAE)

sowie die Vielfalt der chinesischen Kernforschung.

Die Länderreports wurden gemeinsam von WiN und

IYNC erarbeitet und mehrheitlich auch gemeinsam

präsentiert. Das Kongressprogramm war sehr umfangreich

und die Qual der Wahl der zahlreichen technischen

Vorträge, die parallel – teilweise an drei verschiedenen

Orten stattfanden – war oft schwierig.

Unter Gabi Voigt's (WiN Global Präsidentin seit 2016)

speditiver Leitung fand die WiN-Generalversammlung

statt, wo ihre Vorgängerin Se-Moon Park den WiN Honorary

Award entgegennehmen durfte. Der WiN Award ging an

Professor Carla Notari, Argentinien. Mehrere Mentoring

Workshops hatten zum Ziel, junge Frauen zu motivieren,

sich selbstbewusst für eine Karriere in der Nukleartechnik

zu bewerben.

Die begleitenden Besichtigungen zeigten die Geschichte

der argentinischen Nuklearforschung hautnah auf: Schon

1948 überzeugte ein Doktor Richter den damaligen

Präsidenten Perón von seinen Plänen, einen Fusions reaktor

zu bauen. Auf der Isla Huemul, im See von Bariloche,

entstand ein geheimes Fusions-Forschungsinstitut. 1951

verkündete Perón, dass „eine kontrollierte thermonukleare

Reaktionen auf technischer Skala“ erzielt wurde. Doch der

Durchbruch blieb aus und Perón drehte den Geldhahn zu.

Was tun mit den teuren Einrichtungen? Jemand überzeugte

ihn, ein Labor für Kernspaltung (Centro Atomico) und

ein Institut für Nukleartechnik (Instituto Balseiro) zu

gründen.

Zusätzlich gibt es die privatwirtschaftlich organisierte

Firma INVAP, die im Besitz der Provinz Rio Negro ist. Das

Unternehmen hat eine interessante Geschichte. Gegründet

wurde es zur Entwicklung von Forschungsreaktoren und

zur Produktion von medizinischen Isotopen. Dies wurde

zu einem Verkaufsschlager: Südlich von Sydney in

Australien – Veranstaltungsort der WiN Global Conference

2014 – liegt das Kernforschungszentrum ANSTO. Das

Institut hatte damals eben den „Opal“, einen Swimmingpool-Reaktor,

in Betrieb genommen. Er dient als Neutronenquelle

für physikalische Forschungsprojekte, aber

auch zur Herstellung von Mo-99, das in der medizinischen

Diagnostik eine wichtige Rolle spielt. Dies ist einer von

vielen Forschungsreaktoren unterschiedlicher Leistung,

die bei INVAP entwickelt und gebaut wurden. Deren

Entwicklung ist bemerkenswert. Man ist davon weggekommen,

stark angereichertes Uran zu verwenden und

begnügt sich jetzt mit 19,7 %. Argentinien hat Forschungsreaktoren

nach Ägypten, Algerien, Australien, Bolivien,

Kuba, Iran, Peru und Saudi-Arabien geliefert und landete

kürzlich einen weiteren Verkaufserfolg in Petten, Holland!

Wie an vielen Orten der Welt ist das politische Klima für

Kernenergie auch in Argentinien volatil. Deshalb baute

INVAP ein zweites Standbein in der Weltraumtechnik auf.

Diese Sparte ist heute fast wichtiger als der Nuklearbereich.

| | IYNC und WiN Congress in Argentinien: Gruppenfoto der Teilnehmenden.

KTG Inside


atw Vol. 63 (2018) | Issue 6/7 ı June/July

Natürlich baute INVAP auch den RA-6 Reaktor für das

staatliche Centro Atomico in Bariloche. Er wird zu Schulungszwecken

genutzt. Hier werden Reaktoroperateure

ausgebildet. Argentinien hat für die spanischsprachige

Welt auf diesem Gebiet ein „Quasi-Monopol“. Im Zeitalter

der Digitalisierung geht das Centro hier neue Wege: Über

einen separaten Kontrollraum, der gegen Cyber-Attacken

geschützt ist, können Studierende lernen, den Reaktor zu

„fahren“ ohne nach Bariloche reisen zu müssen. Sie sitzen

in ihrer Heimat am Simulator und arbeiten über das

Internet.

Neben Forschungsreaktoren hat Argentinien schon früh

angefangen, Kernkraftwerke zu entwickeln. Der Markt für

angereichertes Uran war damals von den USA beherrscht

und streng reguliert. Um eine Abhängigkeit zu vermeiden,

beschloss man, Natururan als Brennstoff zu verwenden

und lieber das Wasser anzureichern und Schwerwasser

(Deuterium) als Moderator zu nutzen. Die Wasseranreicherungsanlage

hat damals die Schweizer Firma

Sulzer geliefert. Die US-Regierung versuchte dies mit allen

Mitteln zu verhindern – vergeblich. Heute verfügt Argentinien

über drei Kernkraftwerke, alle vom Schwerwassertyp.

In Atucha, unweit von Buenos Aires stehen zwei KKW von

KWU/Siemens, in Embalse ein kanadischer CANDU.

Das Centro Atomico beteiligt sich auch an der Entwicklung

der Reaktoren der Zukunft: Klein und modular.

Die argentinische Variante heißt CAREM und ist wie die

meisten SMR „integriert“, das heißt, alle druckhaltenden

Elemente wie Wärmetauscher und Druckhalter befinden

sich im Reaktordruckbehälter. Der Primärkreislauf kommt

ohne Pumpe aus und wird durch Konvektion angetrieben.

Der Prototyp soll 32 MWe leisten. Zurzeit ist er in Atucha

im Bau. SMR profitieren nicht von der „Economy of Scale“

der Großreaktoren. Sie kompensieren das durch die

geringe Menge an Hochdruck-Installationen.

Man kann nur wünschen, dass die Zusammenarbeit der

Jungen und der Frauen in allen Ländern weitergeführt

wird. Hoffentlich bleibt es nicht bei diesem einen

| | IYNC und WiN Congress in Argentinien: Vor-Ort vor dem Kernkaftwerk Atucha.

gemeinsamen Kongress! Die Kernenergie braucht weltweit

junge Frauen und Männer für zukünftige neue, inhärent

sichere Reaktoren, aber auch für den sicheren Betrieb der

heutigen KKW. Zusätzlich müssen die Ängste vor Radioaktivität

– gerade von Frauen – angesprochen und abgebaut

werden. Ohne den Einbezug der fast CO 2 -freien Kernenergie

ist der Klimaschutz nicht erreichbar!

Irene Aegerter

Irene Aegerter

Irene Aegerter doktorierte nach ihrem Physikstudium am

Eidgenössischen Institut für Reaktorforschung, heute PSI.

Sie war Vizedirektorin des Verbandes Schweizerischer

Elektrizitätsunternehmen (VSE), gründete den Verein

Frauen für Energie in der Schweiz und war Gründungsmitglied

des weltweiten Netzwerkes „Women in Nuclear“

(WiN) sowie Mitglied der Eidgenössischen Kommission

für Sicherheit der Kernanlagen und Vize-präsidentin

der Schweizerischen Akademie der Technischen Wissenschaften.

Links:

http://www.invap.

com.ar/en/invap-2/

about-invap/

invap-headquarters.

html

https://www.cab.

cnea.gov.ar/

http://www.ib.edu.ar/

409

KTG INSIDE

Herzlichen

Glückwunsch!

Juli 2018

93 Jahre wird

8. Dr. Werner Eyrich, Karlsruhe

90 Jahre wird

17. Dipl.-Ing. Karl Josef Sauerwald,

Höchstadt

86 Jahre werden

24. Dipl.-Ing. Joachim May, Burgwedel

27. Dr. Rainer Schwarzwälder, Glattbach

31. Dr. Theodor Dippel,

Eggenstein-Leopoldshafen

83 Jahre wird

20. Dipl.-Ing. Ralf Wünsche, Hannover

82 Jahre werden

1. Dipl.-Ing. Hans-Jürgen Börner,

Weisenheim

25. Dr. Gert Dressler, Lingen

81 Jahre werden

6. Dipl.-Ing. Paul Börner, Steinau

16. Dr. Dieter Hennig, Berlin

24. Dipl.-Ing. (FH) Klaus Haase, Bruchsal

29. Dr. Herbert Reutler, Köln

80 Jahre wird

30. Dr. Philipp Dünner, Odenthal

79 Jahre werden

10. Dr. Bernhard Steinmetz,

Bergisch Gladbach

23. Heinz Stahlschmidt, Erlangen

26. Dipl.-Ing. Ewald Passig, Bochum

78 Jahre werden

2. Prof. Dr. Anton Bayer, Ilmmünster

2. Dr. Manfred Hagen, Berlin

16. Dipl.-Ing. Dietrich Kuschel, Fulda

31. Dr. Peter Schneider-Kühnle

77 Jahre werden

1. Norbert Semann, Bruchsal

24. Dipl.-Ing. Friedrich Wietelmann,

Erkrath

25. Dr. Heinz-Wilhelm Bock, Erlangen

75 Jahre werden

4. Prof. Dr. Ioannis K. Hatzilau,

Athen/GR

10. Dipl.-Ing. Dieter Eder, Alzenau

70 Jahre werden

6. Dr. Hans-Urs Zwicky, Remigen

19. Dr. Wolfgang Boeßert, Pirna

65 Jahre wird

18. Dipl.-Chem. Christel Herzog,

Dresden

60 Jahre werden

8. Dipl.-Ing. Uwe Süss, Stutensee

14. Jochen Rotzsche, Oldenburg

18. Dipl.-Masch.-Ing. Manfred Meyer,

Schwegenheim

27. Prof. Dr. Joachim Axmann,

Braunschweig

28. Dr. Peter Schreiber, Hohenstorf

50 Jahre werden

5. Dr. Peter Engelhard, Essen

9. Lothar Seiffert, Furtwangen

21. Heike Mohrhardt, Hockenheim

KTG Inside


atw Vol. 63 (2018) | Issue 6/7 ı June/July

410

NEWS

Wenn Sie keine

Erwähnung Ihres

Geburtstages in

der atw wünschen,

teilen Sie dies bitte

rechtzeitig der KTG-

Geschäftsstelle mit.

KTG Inside

Verantwortlich

für den Inhalt:

Die Autoren.

Lektorat:

Natalija Cobanov,

Kerntechnische

Gesellschaft e. V.

(KTG)

Robert-Koch-Platz 4

10115 Berlin

T: +49 30 498555-50

F: +49 30 498555-51

E-Mail:

natalija.cobanov@

ktg.org

www.ktg.org

August 2018

94 Jahre wird

1. Prof. Dr. Wolfgang Stoll, Hanau

92 Jahre wird

17. Dr. Rolf Schütte, Marburg

88 Jahre wird

2. Dipl.-Phys. Wolfgang Schwarzer,

Weilerswist

87 Jahre werden

9. Prof. Dr. Hans-Jürgen Laue, Karlsruhe

11. Dipl.-Ing. Siegfried Dreyer, Overath

22. Dr. Dieter Eitner, Mannheim

84 Jahre werden

15. Dipl.-Phys. Heinrich Glantz,

Eggenstein-Leopoldshafen

24. Dipl.-Ing. Heinz Großer, Geesthacht

83 Jahre werden

16. Dr. Dietmar Albert, Salzgitter

29. Dr. Hans-Jürgen Engelmann, Peine

82 Jahre werden

16. Dipl.-Ing. Dietrich Seeliger, Escheburg

26. Dr. Günther Lill, Herzogenaurach

31. Dr. Hartwig Poster, Radeberg

80 Jahre werden

6. Prof. Dr. Rudolf Avenhaus, Baldham

9. Dr. Carl-Otto Fischer, Berlin

21. Dr. Gerhard Schücktanz, Altdorf

25. Dipl.-Phys. Armin Jahns,

Bergisch Gladbach

79 Jahre werden

1. Dipl.-Ing. Gerhard Becker,

Neunkirchen-Seelscheid

8. Dipl.-Ing. Gottfried Merten, Rastatt

24. Dipl.-Ing. Hans Wild, Bruchsal

29. Dr. Adolf Hüttl,

Monte Estoril/Portugal

31. Dr. Dietrich Ekkehard Becker,

Deisenhofen

78 Jahre werden

20. Dr. Herwig Pollanz,

Linkenheim-Hochstetten

24. Dipl.-Ing. Franz Schüler,

Bubenreuth

77 Jahre werden

12. Dr. Klaus Riedle, Uttenreuth

17. Dipl.-Ing. Jörg-Hermann Gutena,

Emmerthal

20. Dr. Willi Theis, Wien/A

21. Dipl.-Phys. Peter Kahlstatt, Hameln

29. Dipl.-Volkswirt Eckhard Strecker,

Bonn

76 Jahre wird

28. Dipl.-Ing. Hans-J. Fröhlich, Berzhahn

75 Jahre wird

26. Dr. Roland Wutschig, Kaiserslautern

70 Jahre werden

2. Dipl.-Ing. Gerd-Rainer Lang,

Sessenbach

11. Dipl.-Ing. Ulrich Gräber, Stuttgart

16. Dr. Gerhard Gräbner, Frankfurt/M.

29. Dipl.-Ing. Peter Jung, Linkenheim

65 Jahre werden

2. Dipl.-Ing. Claus Peter Barthelmes,

Erlangen

3. Michael Eigenbauer, Burgau

17. Dipl.-Ing. Volker Duill, Vechelde

20. Dr. Bernd Schubert, Hamburg

23. Dr. Reinhard Marquart, München

26. Dipl.-Ing. Friedrich Hodde, Jülich

29. Ing. Günter Schwarzl, Braunschweig

60 Jahre werden

2. Dipl.-Ing. Steffen Kniest, Dresden

7. Dipl.-Ing. Eberhard H. Rausch,

Stockstadt

12. Dipl.-Ing. (FH) Uwe Schuster,

Neckarsulm

17. Dr. Wolfgang Faber, Hannover

31. Dipl.-Ing. Michael Brielmayer,

Mainhausen

31. Dipl.-Ing. Hans-Michael Kursawe,

Herzogenaurach

50 Jahre werden

12. Ronny Ziehm, Alzenau

15. Dipl.-Phys. Anja Koschel, Düsseldorf

28. Dipl.-Ing. Frank Staude, Winterbach

Die KTG gratuliert ihren Mitgliedern

sehr herzlich zum Geburtstag und

wünscht ihnen weiterhin alles Gute!

Top

Framatome and the EPR

reactor: a robust history and

the passion it takes to succeed

(framatome) “The Taishan 1 EPR

reactor in China started up on June 6.

This first chain reaction comes as the

culmination of intense work accomplished

by the nuclear industry for

which Framatome is one of the key

actors, as a part of the EDF Group.

Framatome, the designer of this

Generation III+ reactor, is proud, as I

am proud, of this first start up, which

underscores and rewards the years of

engineering devoted to achieving this

success alongside our client TNPJVC.

As one of the largest commercial contracts

ever granted to the French

nuclear sector and, more generally,

in the history of the civil nuclear industry,

the Taishan 1 & 2 project confirms

Framatome’s position as a major

nuclear actor. On this project, we have

handled the design engineering, the

| | Framatome and the EPR reactor: a robust history and the passion it takes to succeed.

View of the Taishan site in China.

supply of the nuclear islands, and the

provision of the fuel assemblies, as

well as related technology transfers.

Today, Framatome is involved in

the Today, Framatome is involved in

the six EPR reactors under construction

worldwide: Taishan 1&2 in China,

Olkiluoto 3 in Finland, Flamanville 3

in France and Hinkley Point C in the

United Kingdom. The EPR reactor,

specified for a 60-year operating lifetime

and with a capacity of 1,650

MWe, is the first Generation III+ pressurized

water reactor and has been

designed to deliver unsurpassed levels

of safety, durability and performance.

All of Framatome’s expertise in

nuclear project management and our

lessons learned from the construction

of numerous nuclear facilities in

France and around the world, have

gone into this genuine “first-of- a-

kind” reactor model. The EPR therefore

benefits from decades of research

and development aimed at ensuring

the safe and secure operation of the

reactor. The EPR has been designed to

minimize its environmental footprint

and optimize waste management and

the exposure of its operators and

maintenance personnel.

As the reactor designer, Framatome

has brought all its experience to bear

in the field of licensing alongside a

great number of regulatory bodies,

particularly in France, Finland, China,

News


atw Vol. 63 (2018) | Issue 6/7 ı June/July

the United Kingdom and the United

States.

Framatome makes a point of

supporting its customers in the startup

of its EPR reactors and places its

depths of competencies at the disposal

of operators to help drive major new

build projects. So that today, so that

tomorrow, nuclear energy is and shall

remain a strategic choice for lowcarbon

and ever more reliable, safe

and competitive electricity.”

| | www.framatome.com

Westinghouse AP1000 plant

to load fuel

(westinghouse) Westinghouse Electric

Company, China State Nuclear Power

Technology Corporation (SNPTC)

and CNNC Sanmen Nuclear Power

Company Limited (SMNPC) announced

in April that the world’s first

unit of AP1000 nuclear power plant

located in Sanmen, Zhejiang Province,

China, has received the fuel load permit

from China’s National Nuclear

Safety Administration (NNSA) and has

commenced initial fuel loading.

“Today we have reached a tremendous

milestone for Westinghouse and

our AP1000 plant technology,” said

José Emeterio Gutiérrez, Westinghouse

president and chief executive

officer. “This is the next major step in

delivering the world’s first AP1000

plant to our customer and demonstrating

the benefits of our advanced

passive safety technology to the

world.”

Sanmen Unit 1 has successfully

completed all the necessary functional

tests as well as technical, safety and

Chinese regulatory reviews. The fuel

load process will be followed by initial

criticality, initial synchronization to

the electrical grid, and conservative,

step by step, power ascension testing,

until all testing is safely and successfully

completed at 100% power.

“This major project milestone

marks the start of the final commissioning

program for Sanmen Unit 1,”

said David Durham, Westinghouse

New Projects Business senior vice

president. “I am confident that our

teams will continue to operate at the

highest levels – at Sanmen, as well as

the Haiyang and Vogtle projects and

in our ongoing support of the worldwide

operating fleet.”

Commenting on Westinghouse’s

partnership with the Chinese government

and suppliers as key contributors

to the successful delivery of clean

energy, Gavin Liu, president – Asia

Region stated, “Westinghouse is

proud to be a partner in China’s

| | Westinghouse Sanmen (China) AP1000 plant to load fuel.

forward-looking nuclear energy

program, an effort that will provide

clean-air electricity to power China’s

economy. Through technology transfer,

localization and infrastructure

development, Westinghouse continues

to collaborate with our Chinese

partners and supports the development

of China’s nuclear power

industry.”

In 2007, Westinghouse successfully

won the bid for China’s generation

III+ nuclear power projects to build

two units of AP1000 reactors in Sanmen,

Zhejiang Province and two units

in Haiyang, Shandong Province. The

company has two additional units

currently under construction at the

Vogtle Electric Generating Plant near

Waynesboro, Georgia.

First criticality of the reactor is

expected to be achieved in JUne/July

2018.

| | www.westinghousenuclear.com

Nuclear pioneers looking

to solve our most pressing

challenges

• New “beyond electricity” capabilities

include process heat, deep

decarbonization

• Several advanced reactor designs

moving toward regulatory approval

• Novel uses for advanced nuclear

technologies will improve their

economics

• “When TerraPower was formed 12

years ago, we were not a nuclear

reactor developer. What we were

looking to do was to solve energy

poverty for one billion people and

to decarbonize the world.”

That was TerraPower’s President

Chris Levesque, speaking to a rapt

crowd at this year’s Nuclear Energy

Assembly (NEA), NEI’s annual conference.

No fewer than four panels

at the event drew packed audiences

excited to hear what the new types

of reactors just over the horizon

will bring.

More than 40 companies and research

institutions are investigating

small modular reactor (SMR) and

advanced nuclear reactor concepts.

And more are on the way.

On NEA’s first day, Dominion

Energy announced it is investing in

GE Hitachi Nuclear Energy’s brandnew

BWRX-300 SMR design. “The

BWRX-300 represents a significant

improvement in the economics of new

nuclear, an imperative for the longterm

viability of the industry,” GE

Hitachi Executive Vice President of

Nuclear Plant Projects Jon Ball said.

But one of the main advantages of

advanced nuclear technologies is the

new and innovative uses they offer

beyond generating electricity. Their

ability to operate at higher temperatures

makes them available for industries

needing process heat for chemicals

production, desalination and

hydrogen production.

Utah Associated Municipal Power

Systems (UAMPS) is teaming with

small modular reactor developer

NuScale Power LLC to build a power

plant at the Idaho National Laboratory

in the 2020s. UAMPS Chief

Executive Officer Douglas Hunter

said NuScale’s small footprint and

enhanced safety will allow its

industrial customers to make the

most of the reactors’ process heat by

moving “right up to our fence line.”

Kathryn McCarthy, vice president

for research and development at

Canadian Nuclear Laboratories

(CNL), said one potential new revenue

stream for SMRs in Canada is producing

hydrogen to decarbonize

the transportation sector, including

long-distance trucks, trains and the

Toronto light rail system. CNL also is

looking at how nuclear plants can

operate in load-following mode to

better balance intermittent wind and

solar generation.

Levesque said TerraPower’s Traveling

Wave Reactor design is now

moving out of the research phase and

entering the test phase, with a view to

obtaining regulatory approvals from

the U.S. Nuclear Regulatory Commission

or China’s National Nuclear

Safety Administration. The company

also is working on a molten chloride

fast reactor concept and has several

domestic and overseas partners on

both projects.

411

NEWS

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412

NEWS

*)

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

“Advanced reactor development is

a heavy lift and we need talent, capital

and intellectual property collaboration

from more than one country.

Technical primacy no longer has to be

confined to one country – these are

global projects that need global resources,”

Levesque said.

The Electric Power Research Institute

(EPRI) recently released a report

on the economic viability of advanced

reactor designs, which feature fewer

and simpler systems, components and

buildings and can be built cheaper

and quicker. “The good news is that a

lot of these technologies are headed to

the lower end of the cost spectrum,”

said Tina Taylor, EPRI’s senior director

of research and development and deputy

chief nuclear officer.

| | www.nei.org

World

IEA: Only 4 out of 38 cleanenergy

technologies are on

track to meet long-term

climate goals

(iea) The International Energy Agency’s

new and most comprehensive

analysis of the clean-energy transition

finds that only 4 out of 38 energy

technologies and sectors were on track

to meet long-term climate, energy

access and air pollution goals in 2017.

According to the IEA report and

website www.iea.org/tcep/ nunclear

power is part of the clean-energy technologies

but more efforts are needed

for expansion.

The findings are part of the IEA’s

latest Tracking Clean Energy Progress

(TCEP), a newly updated website

released today that assesses the latest

progress made by key energy technologies,

and how quickly each technology

is moving towards the goals of

the IEA’s Sustainable Development

Scenario (SDS).

Some technologies made tremendous

progress in 2017, with solar PV

seeing record deployment, LEDs

quickly becoming the dominant source

of lighting in the residential sector,

and electric vehicle sales jumping by

54%. But IEA analysis finds that most

technologies are not on track to meet

long-term sustainability goals. Energy

efficiency improvements, for example,

have slowed and progress on key technologies

like carbon capture and storage

remains stalled. This contributed

to an increase in global energy-related

CO 2 emissions of 1.4% last year.

Operating Results January 2018

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 910 880 744 682 567 682 567 255 336 753 100.00 100.00 99.77 99.77 100.82 100.82

OL2 Olkiluoto BWR FI 910 880 744 688 417 688 417 244 987 599 100.00 100.00 99.92 99.92 100.58 100.58

KCB Borssele PWR NL 512 484 744 381 482 381 482 158 588 401 99.82 99.82 99.82 99.82 100.47 100.47

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 380 365 744 285 615 285 615 131 450 488 100.00 100.00 100.00 100.00 101.05 101.05

KKG Gösgen 7) PWR CH 1060 1010 744 749 519 749 519 305 944 106 100.00 100.00 99.98 99.98 95.04 95.04

KKM Mühleberg BWR CH 390 373 744 287 530 287 530 124 625 675 100.00 100.00 99.89 99.89 99.09 99.09

CNT-I Trillo PWR ES 1066 1003 744 789 686 789 686 239 814 110 100.00 100.00 100.00 100.00 99.19 99.19

Dukovany B1 PWR CZ 500 473 744 373 503 373 503 109 003 986 100.00 100.00 99.88 99.88 100.40 100.40

Dukovany B2 PWR CZ 500 473 617 307 527 307 527 104 930 065 82.93 82.93 82.12 82.12 82.67 82.67

Dukovany B3 PWR CZ 500 473 744 370 166 370 166 102 992 593 100.00 100.00 100.00 100.00 99.51 99.51

Dukovany B4 PWR CZ 500 473 744 371 602 371 602 103 643 343 100.00 100.00 100.00 100.00 99.89 99.89

Temelin B1 1,2) PWR CZ 1080 1030 0 0 0 106 481 294 0 0 0 0 0 0

Temelin B2 PWR CZ 1080 1030 744 809 669 809 669 102 299 615 100.00 100.00 100.00 100.00 100.77 100.77

Doel 1 PWR BE 454 433 744 338 599 338 599 134 553 346 100.00 100.00 99.99 99.99 100.21 100.21

Doel 2 PWR BE 454 433 744 340 286 340 286 132 592 554 100.00 100.00 99.72 99.72 100.49 100.49

Doel 3 3) PWR BE 1056 1006 0 0 0 251 169 221 0 0 0 0 0 0

Doel 4 PWR BE 1084 1033 744 814 535 814 535 255 360 377 100.00 100.00 100.00 100.00 99.95 99.95

Tihange 1 PWR BE 1009 962 744 761 737 761 737 291 600 613 100.00 100.00 100.00 100.00 101.82 101.82

Tihange 2 PWR BE 1055 1008 744 783 542 783 542 249 733 079 100.00 100.00 99.06 99.06 100.36 100.36

Tihange 3 PWR BE 1089 1038 744 813 456 813 456 269 708 286 100.00 100.00 100.00 100.00 100.36 100.36

Operating Results April 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 1,2) DWR 1480 1410 5 3 166 2 859 287 343 051 345 0.75 75.18 0.29 70.54 0.29 66.85

KKE Emsland 4) DWR 1406 1335 720 1 005 788 4 012 463 339 335 746 100.00 100.00 100.00 100.00 99.40 99.17

KWG Grohnde DWR 1430 1360 720 968 379 3 021 184 369 648 763 100.00 78.29 100.00 75.54 93.51 72.95

KRB C Gundremmingen 1,4) SWR 1344 1288 487 543 440 3 387 209 323 967 102 67.69 91.92 67.07 91.63 55.76 87.00

KKI-2 Isar DWR 1485 1410 720 1 040 580 4 223 814 345 822 137 100.00 100.00 99.98 99.99 96.97 98.53

GKN-II Neckarwestheim DWR 1400 1310 720 984 700 3 962 700 324 085 834 100.00 100.00 100.00 99.84 97.87 98.58

KKP-2 Philippsburg 4) DWR 1468 1402 720 932 322 4 049 510 359 217 026 100.00 100.00 99.73 99.91 86.41 94.30

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TCEP provides a comprehensive,

rigorous and up-to-date analysis of

the status of the clean-energy transition

across a full range of technologies

and sectors, their recent progress, deployment

rates, investment levels, and

innovation needs. It is the result of

a bottom-up approach backed by

the IEA’s unique understanding of

markets, modeling and energy statistics

across all fuels and technologies,

and its extensive global technology

network, totaling 6,000 researchers

across nearly 40 technology collaboration

programmes.

The analysis includes a series of

high-level indicators that provide an

overall assessment of clean energy

trends and highlight the most important

actions needed for the complex

energy sector transformation.

For the first time, the analysis also

highlights more than 100 key innovation

gaps that need to be addressed

to speed up the development and

deployment of these clean energy

technologies. It provides an extensive

analysis of public and private clean

energy research and development

investment. It found that total public

spending on low-carbon energy technology

innovation rose 13% in 2017,

to more than USD 20 billion.

“There is a critical need for more

vigorous action by governments,

industry, and other stakeholders to

| | IEA: Only 4 out of 38 clean-energy technologies are on track to meet long-term climate goals. IEA's website.

drive advances in energy technologies

that reduce greenhouse gas emissions,”

said Dr Fatih Birol, the IEA’s

Executive Director. “The world doesn’t

have an energy problem but an

emissions problem, and this is where

we should focus our efforts.”

A total of 11 of 38 technologies

surveyed by the IEA were significantly

not on track. In particular, unabated

coal electricity generation (meaning

generation without Carbon Capture,

Utilisation and Storage, or CCUS),

which is responsible for 72% of power

sector emissions, rebounded in 2017

after falling over the last three years.

Meanwhile, two technologies,

onshore wind and energy storage,

were downgraded this year, as their

progress slowed. This brought the

number of technologies “in need of

improvement” to a total of 23.

413

NEWS

Operating Results February 2018

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 910 880 672 618 566 1 301 132 255 955 319 100.00 100.00 100.00 99.88 101.15 100.98

OL2 Olkiluoto BWR FI 910 880 672 620 840 1 309 258 245 608 439 100.00 100.00 99.85 99.89 100.42 100.50

KCB Borssele PWR NL 512 484 672 344 895 726 377 158 933 296 99.87 99.84 99.87 99.84 100.56 100.51

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 380 365 672 258 171 543 786 131 708 659 100.00 100.00 100.00 100.00 101.09 101.07

KKG Gösgen 7) PWR CH 1060 1010 672 720 590 1 515 109 306 709 696 100.00 100.00 99.98 99.98 101.16 100.94

KKM Mühleberg BWR CH 390 373 672 259 910 547 440 124 885 585 100.00 100.00 99.95 99.92 99.17 99.13

CNT-I Trillo PWR ES 1066 1003 672 713 087 1 502 773 240 527 197 100.00 100.00 100.00 100.00 99.11 99.15

Dukovany B1 PWR CZ 500 473 672 338 390 711 893 109 342 375 100.00 100.00 100.00 99.93 100.71 100.55

Dukovany B2 PWR CZ 500 473 672 338 186 645 713 105 268 250 100.00 91.03 100.00 90.61 100.65 91.20

Dukovany B3 PWR CZ 500 473 672 335 837 706 003 103 328 430 100.00 100.00 100.00 100.00 99.95 99.72

Dukovany B4 PWR CZ 500 473 672 336 572 708 174 103 979 916 100.00 100.00 100.00 100.00 100.17 100.02

Temelin B1 1,2) PWR CZ 1080 1030 0 0 0 106 481 294 0 0 0 0 0 0

Temelin B2 PWR CZ 1080 1030 672 733 731 1 543 400 103 033 346 100.00 100.00 100.00 100.00 101.10 100.92

Doel 1 PWR BE 454 433 672 306 486 645 085 134 859 832 100.00 100.00 99.99 99.99 100.36 100.28

Doel 2 PWR BE 454 433 672 307 293 647 580 132 899 847 100.00 100.00 99.99 99.85 100.58 100.54

Doel 3 3) PWR BE 1056 1006 0 0 0 251 169 221 0 0 0 0 0 0

Doel 4 PWR BE 1084 1033 672 740 003 1 554 538 256 100 379 100.00 100.00 100.00 100.00 100.59 100.25

Tihange 1 PWR BE 1009 962 672 690 038 1 451 775 292 290 651 100.00 100.00 100.00 100.00 102.21 102.01

Tihange 2 PWR BE 1055 1008 672 719 012 1 502 554 250 452 091 100.00 100.00 100.00 99.51 102.06 101.17

Tihange 3 PWR BE 1089 1038 672 734 674 1 548 130 270 442 960 100.00 100.00 99.99 100.00 100.30 100.33

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414

NEWS

This year, the TCEP tracks progress

against the Sustainable Development

Scenario, introduced in the World

Energy Outlook 2017, which depicts a

rapid but achievable transformation

of the energy sector. It outlines a path

to limiting the rise of average global

temperatures to “well below 2°C,” as

specified in the Paris Agreement, as

well as increasing energy access

around the world and reducing air

pollution.

In this scenario, meeting longterm

sustainability goals requires an

ambitious combination of more

energy efficient buildings, industry

and transport, and more renewables

and flexibility in power.

The findings this year are compiled

in an updated website, which provides

easy navigation across technologies

and sectors, and draws links across

the IEA’s resources. The report will

be updated throughout the year as

new data becomes available, and will

be complemented by cutting-edge

analysis and commentary on notable

developments on the global clean

energy transition.

| | www.iea.org/tcep/

Reactors

NEI. New report sees threat

of blackouts if nuclear

retirements continue

• Gas pipeline failure could lead to

significant disruption in Midwest,

Mid-Atlantic

• White House says power plant

retirements are negatively impacting

resilience of grid

• Preservation of nuclear power

plants could lead to added resilience

(nei) A new NEI study conducted by

ICF details how a future gas pipeline

disruption, combined with continued

nuclear power plant retirements

and/or failure to improve natural gas

infrastructure, could lead to prolonged

electricity service disruption

in the areas served by the PJM Interconnection.

The report comes as the Trump

administration, the Federal Energy

Regulatory Commission and PJM

grapple with the issue of the electricity

grid’s diminishing resilience

due to premature nuclear power plant

retirements.

The ICF report finds that a disruption

to natural gas pipelines would

have a major, prolonged effect on electricity

service in the Mid-Atlantic, if

nuclear power plants are not there as a

backup resource and/or natural gas

infrasctructure is not enhanced.

“Such an event could result in the

loss of nearly 27 gigawatts [GW] of

gas-fired generation, with 18 GW

serving the PJM Mid-Atlantic area, depending

on the severity and location

of such event,” the ICF report says.

“When combined with the retirement

of a similar amount of nuclear

capacity, the analysis implies such an

event would put as much as 22 percent

of the area’s load at risk of being

shed in the highest load hours.”

Earlier this month, the White

House released a statement saying

that “impending retirements of fuelsecure

power facilities are leading

to a rapid depletion of a critical part

of our Nation’s energy mix, and impacting

the resilience of our power

grid.”

According to the ICF report, during

a gas pipeline disruption caused by

extreme weather or equipment failure

and lasting 60 days, the PJM service

area would experience “load losses for

more than 200 hours spread across as

many as 34 days.”

“Of the nearly 18 GW of gas-fired

capacity that could be impacted by

such an event, over 45 percent has no

backup fuel capability and would be

immediately unavailable during such

an event,” the report says.

The report examines a scenario,

The Policy Case, in which nuclear

power plants continue to run thanks

to prudent state and federal policies.

Under that scenario, nuclear power

plants would be able to compensate

for the losses in natural gas generation

due to an unexpected interruption.

“The nuclear capacity that remains

online is able to offset the gas generation

impacted by the infrastructure

event, resulting in load being served

in all hours over the 60-day period,”

the report says.

Nuclear power plants have a longterm

supply of fuel onsite. A steady

supply of fuel delivered via pipelines is

necessary to generate electricity from

natural gas. Some natural gas plants

have the ability to run on oil as a

backup fuel, but those supplies would

only last for a handful of days before

needing to be refilled.

The steady supply of onsite fuel is

one reason nuclear power plants are

able to continue supplying electricity

during extreme weather, including

2017’s Hurricane Harvey in Texas.

Last year, Energy Secretary Perry

directed FERC to “take swift action”

to address threats to the resiliency

of the U.S. electric grid and issue a

rule requiring organized markets to

develop rules to compensate “fuelsecure”

electricity generators for the

resiliency they provide. FERC declined

to adopt that proposed rulemaking,

but the agency did open a new proceeding

in which it directed regional

transmission organizations (RTOs)

and independent system operators

(ISOs) to assess grid resilience and

recommend actions. The ICF report

has been submitted to FERC as part of

this proceeding.

Last month, 10,000 megawatts of

nuclear generating capacity failed to

clear PJM Interconnection’s annual

capacity auction. NEI President and

CEO Maria Korsnick said that result

showed the urgent need for electricity

market reform.

NEI Senior Director of Policy

Development Matt Crozat adds that

mounting evidence points to serious

underlying flaws in how electricity

markets are set up.

“This new study underscores

nuclear power’s vital role in ensuring

a reliable and resilient supply of

electricity,” Crozat said.

“Policymakers and administrators

interested in continuing and strengthening

the resilience of America’s grid

should act promptly to ensure nuclear

power plants are fairly compensated

in the marketplace for the reliable,

emission-free electricity they provide.”

| | www.nei.org

Fuel Management

and Disposal

U.S.: Huge bipartisan

majority passes Used Fuel

Management Bill in House

• Republicans and Democrats join

together to pass bill 340-72

• Passage major victory for Rep.

Shimkus; sets up possible Senate

legislation later this year

• Bill expedites Yucca Mountain licensing,

provides for centralized

interim storage

The U.S. House of Representatives

made overdue progress toward

solving the long-standing issue of

used fuel management, with the passage

of the Nuclear Waste Policy

Amendments Act of 2018 (HR 3053)

with a bipartisan vote of 340-72.

Rep. John Shimkus (R-Ill.), chairman

of the House Energy and

Commerce Committee’s environment

News


atw Vol. 63 (2018) | Issue 6/7 ı June/July

subcommittee, authored the bill “to

reform the used fuel management

and disposal program to assure its

sustainability going forward.” A

previous version of the bill passed

the committee last year by a vote

of 49-4.

The Nuclear Energy Institute

lauded the passage of the bill.

“Today’s bipartisan vote represents

overdue progress towards solving a

longstanding issue. House passage of

the Nuclear Waste Policy Amendments

Act begins a much needed step

forward regarding reform to implement,

at long last, the federal government’s

statutory obligation to manage

used nuclear fuel,” NEI President and

Chief Executive Officer Maria Korsnick

said.

“Earlier this year we marked a

troubling milestone: 20 years of

government failure to meet its legal

obligation to take possession of used

fuel. This abdication of responsibility

has harmed electricity consumers and

U.S. taxpayers.”

The original Nuclear Waste Policy

Act became law in 1982, creating a

structured program requiring the

U.S. Department of Energy to begin

removing used fuel from reactor sites

in January 1998. To cover the program’s

costs, DOE and reactor owners

entered into contracts under which

owners paid a fee of one-tenth of a

cent per kilowatt-hour of nuclear electricity

generated into a Nuclear Waste

Fund. To date, electricity consumers

have paid more than $40 billion into

the fund and with interest accruing

more than $1.7 billion annually, a

balance of more than $38 billion

remains.

Since 1987, DOE focused on developing

a repository at Yucca Mountain,

Nevada, spending approximately $10

billion on the program and submitting

a license application to the U.S.

Nuclear Regulatory Commission in

2008. In 2010, DOE declared Yucca

Mountain “unworkable” and unsuccessfully

attempted to withdraw its

application.

Since DOE missed the January

1998 deadlines, the courts have

held the government liable for

DOE’s inaction, awarding reactor

owners damages for DOE’s failure

to meet its January 1998 statutory

deadline.

“U.S. taxpayers have paid more

than $7 billion in damages, and will

continue to pay more than $2 million

a day until the federal government

moves the fuel from plant sites,”

Korsnick noted.

The technical staff of the NRC,

after the court ordered it to complete

its safety and environmental reviews

of DOE’s application, found the

repository in compliance with all

applicable regulations. The court

also ordered DOE to stop collecting

Nuclear Waste Fund fees, which it

did starting mid-2014.

This bill includes provisions that

would move the Yucca Mountain project

forward by helping to resolve key

issues such as land withdrawal and

infrastructure issues.

It would also increase the statutory

limit for used fuel to be placed in the

repository to 110,000 metric tons

from the present 70,000 metric tons

and clarify DOE’s authority to advance

privately owned consolidated interim

storage facilities.

Additionally, it provides a pathway

for bringing Nevada and the local

communities to the table to discuss

benefits associated with these projects.

The legislation also addresses

Nuclear Waste Fund fees, preventing

DOE from collecting any fees until

the NRC issues a final determination

on the Yucca Mountain construction

authorization application. It also

restricts DOE’s fee collections to no

more than 90 percent of the amount

Congress appropriates for the program

in any given year.

Support for the bill came from

many stakeholders, including a joint

letter to Congress from NEI, the

American Public Power Association,

the Edison Electric Institute and the

National Rural Electric Cooperation

Association. Other letters of support

were transmitted from labor unions,

including the AFL-CIO, the International

Brotherhood of Electrical

Workers and North America’s Building

Trades Unions.

The Senate has legislation of its

own that could be introduced later

this year.

“The industry recognizes that the

House and Senate have differing

views on how to reform the used fuel

program. We encourage the two

bodies to continue to advance their

respective proposals and reach a

compromise by the end of the year,”

Korsnick said.

“We look forward to continuing

to work with lawmakers to reach

bipartisan consensus on the best

approach for the long-term management

of the nation’s used fuel. We

urge lawmakers to ensure that resulting

legislation protects both electricity

consumers and taxpayers.”

| | www.congress.gov, www.nei.org,

Research

DOE awards $24 million to

10 advanced nuclear projects

Awards focus on boosting efficiency

and safety, lowering costs of advanced

reactor designs

Reactor control, load-following

and prefabrication techniques win

awards

DOE’s Perry: Awards will help

America retain its technological edge

In yet another example of the

current administration’s continuing

enthusiasm for nuclear energy, the

U.S. Department of Energy this week

announced up to $24 million to fund

10 projects that will boost advanced

nuclear reactor designs.

The awards play a fundamental

role in ensuring America retains its

technological leadership in commercial

nuclear energy, the Nuclear

Energy Institute said.

“It’s gratifying to see DOE taking a

leading role in investing in the longterm

future of this critical American

technology that enhances energy

security and boosts grid resilience

while lowering emissions,” NEI Senior

Technical Advisor for New Reactor

and Advanced Technology Everett

Redmond said.

The awards are part of a new

Advanced Research Projects Agency-

Energy (ARPA-E) program, Modeling-

Enhanced Innovations Trailblazing

Nuclear Energy Reinvigoration

(MEITNER), which will identify and

develop technologies that enable

designs for lower cost, more easily

constructible and safer advanced

nuclear reactors.

“Nuclear energy is an essential

component of the U.S. energy mix,

and by teaming up with the private

sector to reduce costs and improve

safety, we are keeping America ahead

of the curve in advanced reactor

design and technology,” U.S. Secretary

of Energy Rick Perry said in the

DOE statement.

“These next-generation ARPA‐E

technologies help us maintain

our competitive, technological edge

globally, while improving the resilience

of the grid and helping provide

reliable, baseload electricity to each

and every American.”

Many of the awards focus on

improving the efficiency and safety of

several advanced nuclear technology

designs now under development. The

designs come in a range of sizes, from

a couple of megawatts to more than

a 1,000 megawatts of generating

capacity. They feature a range of

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| | Mounting of a battery cell in the instrument ANTARES at FRM II.

Photo: Wenzel Schürmann / TUMwebsite.

coolants including high-temperature

gas, molten salt and liquid metal.

Some of these reactors may even be

capable of recovering and reusing elements

in used fuel to produce even

more energy.

The winning project teams will

have access to DOE’s modeling and

simulation resources as they develop

their concepts and will coordinate regularly

with experts from across DOE

and the national laboratory system.

The winning projects and their

award amounts include:

• A novel circulation pump for

molten salt reactors to improve

plant performance, increase pump

lifetime and reduce cost will be

developed by Terrestrial Energy

USA ($3,150,000).

• A new reactor control technology

to enhance passive safety and

reduce costs for the molten salt

reactor and other designs being

developed by Yellowstone Energy

($2,599,185).

• A cost-saving construction method

for concrete reactor components

being developed by General

Atomics that will use factory

pre-cast modules of ultra highstrength

concrete ($1,532,752).

• General Atomics also is developing

a detailed, dynamic model of a

nuclear power system with rapid

load-following capability, enabling

grid synergies with renewable wind

and solar sources ($1,455,762).

• Westinghouse Electric Co. will

develop a “solid core block” of

materials that will self-regulate

the reaction rate in a nuclear

reactor, allowing it to achieve safe

shutdown without external power

or operator intervention

($5,000,000).

An overarching theme of the

awards is the drive to increase the

efficiency and economic performance

of advanced reactor technologies

under development, while lowering

their construction costs, Redmond

noted.

“New energy technologies can improve

our quality of life, benefit the

environment and create jobs, but they

must also be economically viable if

they are to see wide commercialization,”

Redmond said.

“These awards acknowledge the

fact that advanced nuclear technologies

must not only be groundbreaking,

they must also be affordable.”

| | www.nei.org, www.energy.gov

Neutrons pave the way to

accelerated production of

lithium-ion cells

(frmii) Developers from Bosch and

scientists at the Technical University

of Munich (TUM) are using neutrons

to analyze the filling of lithium ion

batteries for hybrid cars with electrolytes.

Their experiments show that

electrodes are wetted twice as fast in a

vacuum as under normal pressure.

One of the most critical and

time-consuming processes in battery

production is the filling of lithium

cells with electrolyte fluid following

the placement the of electrodes in a

battery cell. While the actual filling

process takes only a few seconds,

battery manufacturers often wait

several hours to ensure the liquid is

fully absorbed into the pores of the

electrode stack.

The fact that neutrons are hardly

absorbed by the metal battery housing

makes them ideal for analyzing batteries.

That is why Bosch employees,

in collaboration with scientists from

the TU Munich and the University of

Erlangen-Nuremberg, investigated

the filling process at the neutron

i maging and tomography facility

ANTARES of the research neutron

source FRM II.

Faster in a vacuum

Manufacturers of lithium cells often

fill the empty cells in a vacuum. The

process is monitored indirectly using

resistance measurements. “To make

sure that all the pores of the electrodes

are filled with the electrolyte,

manufacturers build in large safety

margins,” says Bosch developer

Dr. Wolfgang Weydanz. “That costs

time and money.”

In the light of the neutrons, the

scientists recognized that in a vacuum

the electrodes were wetted completely

in just over 50 minutes. Under normal

pressure, this takes around 100

minutes. The liquid spreads evenly in

the battery cell from all four sides,

from the outside in.

In addition, the electrodes absorb

ten percent less electrolyte under

normal pressure. The culprit is gases

that hinder the wetting process, as the

scientists were able to demonstrate for

the first time using the neutrons.

| | mlz-garching.de/antares

Company News

Bilfinger awarded contract

for superconducting magnetic

modules

(bilfinger) Bilfinger has received a

major contract from GSI Helmholtzzentrum

für Schwerionenforschung

GmbH in Darmstadt, Germany, for

the construction and delivery of 83

superconducting magnetic modules

along with twelve additional modules.

The contract is valued at over €20

million. The modules will be used

in the SIS100 accelerator ring at

the Facility for Antiproton and Ion

Research (FAIR). FAIR is one of the

world’s largest research projects with

an investment volume of more than

€1 billion.

“Research facilities are crucial

customers for us. I am therefore

particularly pleased that we have been

able to expand our partnership with

GSI,” says Ronald Hepper, Managing

Director of Bilfinger Noell. The

Bilfinger subsidiary is also working

with the Karlsruhe Institute of

Technology and CERN, the European

Organization for Nuclear Research,

among other research facilities.

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| | Bilfinger awarded contract for superconducting magnetic modules. 3D view of the superconducting

magnetic module (Picture: GSI)

For more than ten years now,

Bilfinger has been involved in the

FAIR project. The FAIR accelerator

will allow numerous experiments to

be conducted to find out more about

the structure of matter and the origin

of the universe as well as to improve

cancer therapy. The SIS100 accelerator

ring generates the particle beams

required for all these experiments.

Bilfinger is currently producing a

series of 110 superconducting dipole

magnet modules for the SIS100.

The basis for this is a prototype that

Bilfinger also helped develop.

The 83 superconducting magnetic

modules which are the subject of the

new order each consist of two magnet

units, a vacuum vessel with a thermal

shield, radiant tubes and other

complex components. Each of these

devices has a length of around 5.2

meters and weighs over five tons.

With their different structures, the

different module types pose a logistic

challenge. The twelve additional

modules serve to connect superconducting

magnet modules in the

SIS100 particle accelerator.

Delivery of the first superconducting

magnetic modules is planned

to start in early 2019.

| | www.bilfinger.com

Framatome partners with

McAfee to support energy

industry cybersecurity

(framatome) Framatome signed an

agreement with McAfee, the deviceto-cloud

cybersecurity company, to

distribute cybersecurity solutions to

energy transmission, distribution

and generation facilities worldwide.

Together, Framatome and McAfee

will work with these facilities to

help protect their digital assets and

support the reliable production of

electricity.

“In a rapidly evolving digital

landscape, holistic and robust cybersecurity

programs are critical to

protecting nuclear energy facilities

and electrical power and distribution

infrastructure,” said Catherine Cornand,

senior executive vice president

of Framatome’s Installed Base Business

Unit. “This partnership with

McAfee will enhance our ability to

provide customers with the right

combination of cutting-edge technologies

and expertise.”

This partnership builds on the

cybersecurity support Framatome

provides to international power generation

fleets and bulk electric system,

allowing the company to deliver an

even stronger suite of services to

customers, backed by McAfee’s highly

respected brand and cybersecurity

products.

“McAfee’s diverse client portfolio

and global threat intelligence network

give us a comprehensive and real-time

view of cybersecurity threats,” said

Tom Moore, vice president, World

Wide Embedded Sales of McAfee.

“We look forward to combining our

significant knowledge base and

time-tested, evolving solutions with

Framatome’s experience and proficiency

in the electric power industry

to serve this unique customer base.”

When combined with Framatome’s

cyber products and services, the

McAfee cybersecurity hardware,

software, support and incident response

services are a comprehensive

solution to protect international

digital systems from cyberattacks.

This solution also helps meet international

regulations.

| | www.framatome.com

NUKEM Technologies:

Successful finalization of

the ITER Deactivation study

(nukem) ITER is a tokamak under construction

in Cadarache, France. Its

goal is to demonstrate the feasibility of

producing energy by fusion reaction of

Deuterium and Tritium. The machine

should generate first plasma in December

2025. Before construction is even

finished, the deactivation and dismantling

was analyzed already.

For the ITER Deactivation study a

consortium of the companies AMEC

Foster Wheeler (now known as WOOD

PLC and supported by AJR consulting)

and NUKEM Technologies Engineering

Services GmbH (NTES) was

selected. Within the consortium

French, UK, and German knowledge

on nuclear issues was present.

In close cooperation with ITER

Organization about 100 documents

needed to be mined for relevant

information to get to a good understanding

of the equipment which will

be deactivated and dismantled after

the end of plasma operations of ITER.

The documents had sometimes several

hundred pages (e.g. the Safety Report

of ITER) or otherwise huge dimensions

(e.g. the “Bill of materials” with

information on materials used for IT-

ER construction contains several

hundred thousand cells …). The

3D-Models of all facilities of ITER

which will be under French nuclear

legislation were essential for the

analysis to get a good understanding

not only on the equipment’s location

and dimensions, but also on accessibility.

Additional information on

safety parameters of the rooms (radiation

zones, ventilation zones etc.)

needed to be brought together to

better understand possible restrictions.

Throughout the project the expertise

of France, UK and Germany

present in the consortium was used to

estimate the working procedures and

time requirements for all deactivation

operations required. In some cases

knowledge based assumptions where

developed.

For some parts of ITER the analysis

was performed depending on the

systems, while for others the analysis

was performed on a Level by Level and

Room by Room base. Due to the

changes partly resulting from the

first analysis results the basis of the

analysis needed to be changed within

the process. When needed, assumptions

on the parameters of the equipment

required for the deactivation

analysis could be made based on the

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experience of the consortium and

approved by ITER leading to a common

basis for the analysis.

The project’s success was based

on the wide knowledge base provided

by the participating companies. The

frequent meetings within the consortium

and with ITER staff were very

helpful to get a clear understanding of

the processes within all facilities. The

cooperation was on a very good level

of understanding of each other even

with some surprises concerning

different regulations on how to process

or treat radiological issues. The

sequencing of all necessary steps to

deactivate the whole plant took into

account not only tritiated waste

treatment issues but always put safety

first e.g. by establishing specific

methods of operations. For example

before any room is entered by

personnel the first action is to

decontaminate the rooms, if necessary

using robots. The study does

not take into account future possible

dismantling methods but is based on

existing methods.

It was proven that the safe deactivation

of ITER is possible. After the

presentation of the results to ITER at

the final meeting mid of March, ITER

expressed its full satisfaction to the

consortium team.

| | www.nukemtechnologies.com

Orano TN awarded a contract

for the supply of dry storage

casks and services in the USA

(orano) Orano TN, the nuclear

logistics subsidiary of Orano, has been

selected by the American electric

utility Omaha Public Power District

(OPPD) for the final offload of more

than 900 used fuel assemblies contained

in the pool of its Fort Calhoun

nuclear power station in Nebraska. To

carry out this operation, Orano TN

will supply 30 NUHOMS® dry storage

systems and the associated “pool to

pad” transfer services for several tens

of millions of dollars.

NUHOMS® canisters, already in

use at the Fort Calhoun facility, are

approved for both onsite storage

and offsite transportation, ensuring

long-term used fuel management.

“We are proud to continue our

nearly 15-year trusted relationship

with our client OPPD and the Omaha

community,” said Greg Vesey, Senior

Vice President of Orano TN. “As

before, our commitment is to transfer

the used fuel to the Independent

Spent Fuel Storage Installation in a

safe, accelerated and cost-effective

manner.”

| | The beginning of towing of FPU ‘Academik Lomonosov’ to Pevek.

Orano TN’s NUHOMS® systems

have securely stored used nuclear fuel

in the United States for more than two

decades, with installations at 33 sites

around the country.

| | www.orano.group

Westinghouse accident tolerant

fuel development moves

forward with cooperation

agreement with ENUSA

(westinghouse) Westinghouse Electric

Company announced that it will collaborate

in the development of its En-

Core® Fuel, the revolutionary

accident-tolerant fuel (ATF) design,

with ENUSA Industrias Avanzadas

(ENUSA) through a Frame Cooperation

Agreement (FCA).

“This agreement serves to

strengthen the technical and commercial

relations between ENUSA and

Westinghouse as we work to develop

leading nuclear fuel technology,” said

Torbjörn Norén, European Fuel Group

and EMEA Fuel Delivery Director at

Westinghouse. “Westinghouse’s work

with ENUSA in the Spanish and

European Fuel Group markets will

help to facilitate agreements with

customers to launch EnCore Fuel

demonstration programs in their

plants.”

Under the terms of the agreement,

the newly signed FCA establishes

the framework that will regulate the

different Joint Development Programs

(JDPs) to be launched between

both companies. The first JDP will

evaluate the application of the

segmented rod concept and develop

models of ATF / EnCore fuel behavior.

The first JDP will be followed by

other JDPs in the area of codes and

methods, spent fuel management,

and fuel fabrication and inspection

technology.

| | www.westinghousenuclear.com

Russia’s floating nuclear

plant arrives in Murmansk f

or fuelling

(nucnet) Russia’s first commercial

floating nuclear power station, the

Akademik Lomonosov, has arrived in

the port city of Murmansk in the far

northwest of Russia where it will be

loaded with nuclear fuel, state nuclear

corporation Rosatom said.

The plant was moved from its construction

site in St Petersburg with no

nuclear fuel on board. Fuel will be

loaded in Murmansk and the plant

will then be moved for deployment at

Pevek, an Arctic port town in the

country’s far north-eastern region of

Chukotka, in the summer of 2019.

Commissioning is scheduled for the

autumn of 2019.

The Akademik Lomonosov – set to

be the only operating floating nuclear

plant in the world – will be the first

vessel of a proposed fleet of floating

plants with small pressurised water

reactor units that can provide energy,

heat and desalinated water to remote

and arid areas of the country.

It will be the first floating nuclear

station to be built and deployed

since the MH-1A, also known as

the Sturgis, in the US in 1967. The

Sturgis was towed to the Panama

Canal Zone that it supplied with

10 MW of electricity from October

1968 to 1975.

The 21,000-tonne vessel has two

Russian-designed KLT-40S reactor

units with an electrical power

generating capacity of 35 MW each,

sufficient for a city with a population

of around 200,000 people.

Rosatom said the Akademik

Lomonosov will replace capacity lost

when the Bilibino nuclear station in

Chukotka is shut down. According to

Rosatom, Bilibino generates 80% of

electricity in the isolated region.

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Unit 1 at Bilibino is scheduled to be

permanently shut down in December

2018. The remaining three units will

be shut down in December 2021,

Rosatom has said.

| | www.rosatom.ru

GNF Wins $ 250 Million

Contract To Provide BWR Fuel

For Entergy

(nucnet) US-based Global Nuclear

Fuel (GNF) has been awarded a longterm

contract by Entergy Nuclear to

continue to fuel its boiling water

reactors.

The new fuel supply contract,

valued at more than $250m, runs

from 2019 to 2031 and includes 10 reloads

of GNF3 nuclear fuel. Entergy

will be the first customer to take delivery

of GNF3 in reload quantities.

GNF, a General Electric-led joint

venture with Hitachi, said the GNF3

fuel assembly, manufactured in

Wilmington, North Carolina, is

designed to offer improved fuel cycle

economics, increased performance

and flexibility.

GNF3 lead use assemblies have

been operating in several US nuclear

power plants, including four lead use

assemblies in Entergy’s River Bend

nuclear station in Louisiana since

2015.

In 2019, River Bend will become

the first plant in which GNF3 will be

installed in reload quantities. GNF3

will be installed in reload quantities at

Grand Gulf in Mississippi in 2020.

| | nuclear.gepower.com

Forum

10 th Anniversary International

Forum ATOMEXPO 2018

(atomexpo) The ATOMEXPO International

Forum is the largest congress

and exhibition event of the nuclear

industry. Inaugurated in 2009 on

the initiative of Rosatom State Corporation,

the forum is held annually.

Traditionally, the event brings

together leaders of major companies

of the global nuclear industry

and related industries, government

agencies, representatives of international

and public organizations,

and leading experts. Over the years,

ATOMEXPO has become a global

event for the exchange of views and

best practices in the field of effective

nuclear power use, and a popular

venue for business meetings, partnership

agreements and launch of new

projects.

The 10 th International Forum

ATOMEXPO 2018 was held in May in

Sochi. This year the key theme is:

Global Partnership – Joint Success.

The plenary discussion on this theme

will be attended by: Aleksei Likhachev,

Director General of Rosatom State

Atomic Energy Corporation; William

D. Magwood IV, Director General of

the Nuclear Agency of the Organization

for Economic Cooperation

and Development; Liubov Glebova,

member of Russian Federation

Council; Mohamed Shaker, Ministry

of Electricity and Renewable Energy

of Egypt; Mikhail Chudakov, IAEA

Deputy Director general; Hortensia

Jimenez Rivera, General Executive

Director of the Bolivian Atomic

Energy Agency; Necati Yamac,

Deputy Undersecretary of Ministry of

Energy and Natural Resources of

Turkey; Vladimir Semashko, Deputy

Prime Minister of the Republic of

Belarus.

The program includes over 20

round tables and panel sessions,

which will be attended by more than

200 experts. The participants will

address topical issues on multilateral

cooperation in all areas of nuclear

technology application, focusing on

international partnerships in the construction

of nuclear power plants, digital

technologies, research, staff training,

infrastructure development and

security. Signing of important documents

is also planned: agreements on

strategic cooperation and partnership,

commercial contracts, and project

development documents.

The third Russia-IAEA Nuclear

Management School for Managers in

Nuclear Organizations will be held for

the first time at ATOMEXPO. A special

event will premiere the ATOMEXPO

AWARDS, where an international

professional award for outstanding

services will be awarded to global

companies that have made a significant

contribution to the development

of the nuclear industry and the

use of nuclear power for the benefit of

mankind.

It is expected that over 3,000 representatives

of over 600 companies

from over 60 countries will take part

in ATOMEXPO 2018. The exhibition

will present the latest technologies,

products and solutions developed

by more than 115 Russian and international

nuclear power companies

and related industries.

Between 2009 and 2017, about

38,000 attendees from 83 countries

took part in ATOMEXPO events, 987

companies made presentations, and

1,116 experts spoke at 141 round

tables. More than 300 agreements

and international cooperation documents

were signed during the nine

forums. Platinum sponsors of the

anniversary ATOMEXPO 2018 are

Sovcombank and VTB bank; Gold

sponsor – Gazprombank; partners –

TVEL, RosRAO, National Operator

for Radioactive Waste Management;

energy sponsor – REA; operator –

Atomexpo.

| | www.rosatom.ru

People

Holger Bröskamp retires after

15 years with GNS

(gns) Holger Bröskamp (60) retired at

the end of April 2018 after 15 years in

the management of GNS. Bröskamp

was spokesman of the GNS Board of

Managing Directors from March 2003

to September 2011, since then he has

been its deputy chairman.

“During his one and a half decades

at the head of GNS, Holger Bröskamp

has always given equal priority to safe

disposal and the well-being of our

employees,” says Dr. Hannes Wimmer,

Chairman of the Management Board

of GNS, in praise of his colleague

who is leaving the company. “With

these principles, he has had a

lasting influence on GNS and laid

the foundations for our success

today.”

GNS Supervisory Board Chairman

Dr. Guido Knott adds: “Holger

Bröskamp enjoys an excellent reputation

in all decisive waste disposal

issues in politics, among experts and

among domestic and foreign GNS

customers.In a tense period between

phase-out of nuclear energy, the

extension of its operating life and the

second phase-out decision, he made a

decisive contribution to objectifying

the debates. On behalf of all GNS

shareholders and the Supervisory

Board, I would like to thank Mr.

Bröskamp for his outstanding commitment”.

Of the activities Holger Bröskamp

was most recently responsible for,

Dr. Hannes Wimmer has taken over

waste management and the retrieval

of residues from reprocessing, while

Managing Director Dr. Jens Schröder

is now responsible for spent fuel

management.

The operation of the interim

storage facilities in Ahaus and

Gorleben, which were also under

Bröskamp’s responsibility, was

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already taken over by the federal

government last year as part of the

reorganization of nuclear waste

management in Germany.

The Board of Management Directors

of GNS now consists of Dr. Hannes

Wimmer (CEO), Dr. Jens Schröder

(CTO) and Georg Büth (CFO).

| | www.gns.de

Publications

Reaktorsicherheit für

Leistungskernkraftwerke –

Die Entwicklung im

politischen und technischen

Umfeld der Bundesrepublik

Deutschland

Das in der atw Juli 2013, S. 461/462

besprochene Werk von Peter Laufs ist

jetzt in der 2. Auflage erschienen. Neu

aufgenommen und detailliert behandelt

wurden die Themen Alterungsmanagement,

der Rückbau von

Kernkraftwerken und die Entsorgung

radioaktiver Abfälle. Neben diesen

umfangreichen Ergänzungen wurde

der Stoff der 1. Auflage überarbeitet,

erweitert und an die neuere Entwicklung

seit Erscheinen der 1. Auflage

angepasst. Wegen des gewachsenen

Umfangs wurde das Werk in zwei Bände

aufgeteilt, die auch einzeln

bezogen werden können. Der 2. Band,

der auch die genannten neuen Kapitel

enthält, war umgehend vergriffen

und musste nachgedruckt werden.

Ursächlich war offenbar das starke

Interesse an der umfassenden Darstellung

der Endlagerpolitik in

Deutschland und vergleichend in

einigen anderen Ländern. In dieser

geschlossenen Form findet man das

sonst nirgendwo.

Das Buch ist in einer auch für den

interessierten Nichtfachmann verständlichen

Sprache geschrieben und

bietet mit seinem umfangreichen

Literaturverzeichnis viele Materialien

für weiterführende Studien.

Laufs, P.: Reaktorsicherheit für

Leistungskernkraftwerke – Die Entwicklung

im politischen und technischen

Umfeld der Bundesrepublik

Deutschland.

• 2. Auflage. Band 1: 866 Seiten,

389 Abb., 29 Tab., Springer, Berlin

2018, ISBN 978-3-662-53452-6,

124,99 €, eBook 89,99 €;

Band 2: 624 Seiten, 434 Abb.,

24 Tab., Springer, Berlin 2018

ISBN 978-3-662-54163-0,

124,99 €, eBook 89,99 €

| | www.springer.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

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

April 2018

• Uranium: 20.00–21.75

• Conversion: 7.50–8.50

• Separative work: 36.00–39.00

May 2018

• Uranium: 21.75–22.80

• Conversion: 8.00–8.75

• Separative work: 36.00–39.00

| | Source: Energy Intelligence

www.energyintel.com

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

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

2018

I. quarter: 89.88; 5.838.003

| | Source: BAFA, some data provisional

www.bafa.de

EEX Trading Results

April 2018

(eex) In April 2018, the European

Energy Exchange (EEX) increased

volumes on its power derivatives

markets by 12 % to 246.6 TWh (April

2017: 220.7 TWh). In particular, the

markets for Italy (38.8 TWh, +100 %)

and France (19.8 TWh, +48 %)

contributed to this development. Also

the smaller markets for Spain

(7.2 TWh, +54 %) and the Netherlands

(4.3 TWh, +358 %) recorded

significant growth. In Options on

Phelix-DE Futures, at 36.7 TWh,

EEX achieved the highest volume

since the launch of this product.

The April volume comprised

158.3 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

39.08 €/MWh. The Settlement

Price for peak load contract (Phelix

Futures) with delivery in 2019

amounted to 44.85 €/MWh.

On the EEX markets for emission allowances,

the total trading volume increased

by 56 % to 232.9 million

tonnes of CO 2 in April (April 2017:

149.6 million tonnes of CO 2 ). In

particular, the development was driven

by a significant increase of volumes on

the EUA derivatives market where EEX

recorded an increase of 32 % to 88.0

million tonnes of CO 2 (April 2017: 66.5

million tonnes of CO 2 ). Primary

News


atw Vol. 63 (2018) | Issue 6/7 ı June/July

market auctions contributed 79.3 million

tonnes of CO 2 to the total volume.

The EUA price with delivery in

December 2018 amounted to

12.65/14.01 €/ EUA (min./max.).

May 2018

(eex) In May 2018, the European

Energy Exchange (EEX) increased

volumes on its power derivatives

markets by 23% to 306.9 TWh (May

2017: 248.7 TWh) which is the highest

volume since March 2017. In the

benchmark product for European

power trading, the Phelix-DE Future,

EEX achieved new monthly record of

170.4 TWh. Furthermore, EEX was

able to significantly increase volumes

in its markets for Italy (58.4 TWh,

+60%), Spain (9.7 TWh, +52%) and

the Netherlands (3.2 TWh, +283%).

Volumes on the options markets have

tripled as against the previous year to

31.7 TWh (May 2017: 10.3 TWh).

The May volume comprised

191.7 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

44.42 €/MWh. The Settlement Price

for peak load contract (Phelix

Futures) with delivery in 2019

amounted to 51.88 €/MWh.

On the EEX markets for emission

allowances, the total trading volume

more than quadrupled to 477.8 million

tonnes of CO 2 in May (May 2017:

105.1 million tonnes of CO 2 ). On the

EUA secondary market (Spot and

Derivatives), EEX achieved a new

record volume of 204.8 million tonnes

of CO 2 . Also trading in EUA options

reached a new peak of 214.0 million

tonnes of CO 2 . Primary market auctions

contributed 59.0 million tonnes

of CO 2 to the total volume.

The EUA price with delivery

in December 2018 amounted to

12.97/16.31 €/ EUA (min./max.).

| | www.eex.com

MWV Crude Oil/Product Prices

March 2017

(mwv) According to information and

calculations by the Association of the

German Petroleum Industry MWV e.V.

in March 2018 the prices for super

fuel, fuel oil and heating oil noted

inconsistent compared with the

pre vious month February 2018. The

average gas station prices for Euro

| | 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.

| | 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.

super consisted of 134.15 €Cent

( February 2018: 137.27 €Cent, approx.

-2.27 % in brackets: each information

for pre vious month or rather previous

month comparison), for diesel fuel of

118.03 €Cent (119.50; -1.23 %) and

for heating oil (HEL) of 59.87 €Cent

(59.15 €Cent, +1.22 %).

The tax share for super with a

consumer price of 134.15 €Cent

(137.27 €Cent) consisted of

65.45 €Cent (48.79 %, 65.45 €Cent)

for the current constant mineral oil

tax share and 21.42 €Cent (current

rate: 19.0 % = const., 21.85 €Cent)

for the value added tax. The product

price (notation Rotterdam) consisted

of 38.46 €Cent (28.67 %, 38.46 €Cent)

and the gross margin consisted of

8.82 €Cent (6.57 %; 11.82 €Cent).

Thus the overall tax share for super

results of 67.79 % (66.68 %).

Worldwide crude oil prices

(monthly average price OPEC/Brent/

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

higher, approx. +0.77 % (-4.26 %) in

March 2018 compared to February

2018.

The market showed a stable

development with slightly higher

prices; each in US-$/bbl: OPEC

basket: 63.76 (63.48); UK-Brent:

66.02 (65.32); West Texas Intermediate

(WTI): 62.72 (62.23).

April 2017

In April 2018 the prices for super fuel,

fuel oil and heating oil noted higher

compared with the pre vious month

March 2018. The average gas station

prices for Euro super consisted

of 138.96 €Cent (March 2018:

134.15 €Cent, approx. +3.59 % in

brackets: each information for previous

month or rather previous month

comparison), for diesel fuel of

121.09 €Cent (118.03; +2.59 %) and

for heating oil (HEL) of 63.12 €Cent

(59.87 €Cent, +0.27 %).

The tax share for super fuel with

a consumer price of 138.96 €Cent

(134.15 €Cent) consisted of

65.45 €Cent (47.10 %, 65.45 €Cent)

for the current constant mineral oil

tax share and 22.19 €Cent (current

rate: 19.0 % = const., 21.42 €Cent)

for the value added tax. The product

price (notation Rotterdam) consisted

of 41.93 €Cent (30.17 %, 41.93 €Cent)

and the gross margin consisted of

9.39 €Cent (6.76 %; 8.82 €Cent).

Thus the overall tax share for super

results of 66.10 % (67.79 %).

Worldwide crude oil prices

(monthly average price OPEC/Brent/

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

higher, approx. +7.39 % (+0.77 %)

in April 2018 compared to March

2018.

The market showed a stable

development with higher prices; each

in US-$/bbl: OPEC basket: 68.47

(63.76); UK-Brent: 72.11 (66.02);

West Texas Inter mediate (WTI):

66.25 (62.25).

| | www.mwv.de

421

NEWS

News


atw Vol. 63 (2018) | Issue 6/7 ı June/July

422

Confidence in Nuclear Safeguards at Risk

as Trump Quits One Deal to Pursue Another

John Shepherd

NUCLEAR TODAY

References:

Statement by

Yukiya Amano,

https://bit.ly/2HlS7O

Islamic Republic

News Agency report,

https://bit.ly/2sLiIjg

By the time you sit down to read this article, Donald Trump and Kim Jong Un may have had an historic sit-down of their

own – in fact the first meeting between a sitting US president and a leader of North Korea.

The schedule for the meeting has been as unpredictable as

the two leaders themselves and, at the time of writing,

their proposed meeting has been variously described as a

“summit” and “meet and greet”. The description appears

designed to keep expectations in check.

What is certain is that the “denuclearisation” of the

Korean peninsula is on the agenda – although both sides

appear to have different ideas as to what that means.

Under normal circumstances, an outbreak of peace,

harmony and moves to develop an atmosphere of mutual

trust between the US and North Korea would be very

welcome. But whatever relationship the leaders may or

may not be seeking to nurture, I fear the circumstances

that have led up to this appointment with destiny puts at

risk the confidence the international community must

have in the regulated use of civil nuclear power worldwide.

I say this because in the run up to the talks, President

Trump pulled the US out of the 2015 Joint Comprehensive

Plan of Action (JCPOA) with Iran.

Through the JCPOA, Iran effectively had decided to

make a ‘fresh start’ with the international community. The

agreement was with the ‘P5+1’ group of world powers

comprising the US, UK, France, China, Russia and Germany.

Iran agreed to limit its sensitive nuclear activities and

allow in International Atomic Energy Agency (IAEA)

inspectors in return for the lifting of economic sanctions.

The JCPOA was designed to end years of tension and

fears about military aspects of Iran's nuclear activities. Just

last March, IAEA director-general Yukiya Amano said “Iran

is implementing its nuclear-related commitments” under

the JCPOA.

In May, Amano stressed that Iran was “subject to the

world’s most robust nuclear verification regime under the

JCPOA, which is a significant verification gain”. As of

5 June, Amano said the IAEA “can confirm that the nuclearrelated

commitments are being implemented by Iran”.

The IAEA has said repeatedly that, before the end of

2003, “an organisational structure was in place in Iran

suitable for the coordination of a range of activities

relevant to the development of a nuclear explosive device”.

However, the Agency “also assessed that these activities

did not advance beyond feasibility and scientific studies,

and the acquisition of certain relevant technical

competences and capabilities”. The IAEA said it had “no

credible indications of activities in Iran relevant to the

development of a nuclear explosive device after 2009”.

But all this was not enough for the US president. He

effectively questioned the credibility of the IAEA.

Instead, the US president decided it was time to embrace

North Korea, which has a record of frustrating, dodging

and rebuffing all safeguards attempts by the world’s nuclear

community, largely under the auspices of the IAEA.

Readers may recall the 1994 framework agreement,

under which North Korea agreed to freeze work at its then

gas-graphite moderated reactors and related facilities and

to allow the IAEA to monitor that freeze. Pyongyang was

also required to “consistently take steps to implement the

North-South Joint Declaration on the Denuclearization of

the Korean Peninsula” and to remain a party to the Non-

Proliferation Treaty (NPT). In exchange, the US agreed to

lead an international consortium to construct two light

water power reactors, and to provide 500,000 tons of

heavy fuel oil per year until the first reactor came online

with a target date of 2003.

But North Korea delayed safeguards inspections

designed to verify its past nuclear activities – which in turn

meant the proposed power reactors did not materialise. The

international community also feared North Korea had an

illicit highly-enriched uranium programme. Subsequent US

intelligence reports appeared to support those concerns and

it was later revealed that North Korea, together with Libya

and Iran, had illegally acquired gas-centrifuge technology

from a Pakistani nuclear scientist.

North Korea’s behaviour since then has not improved

and it remains to be seen whether Chairman Kim sees a

meeting with President Trump as an opportunity to start

making amends.

However, it cannot make sense for the US to have

worked so hard with the IAEA on bringing Iran into line

with strictly-supervised civil nuclear operations – only to

then say “the JCPOA is not good enough”. It adds insult to

injury to then imply the US is ready to do business with the

rogue nuclear nation of North Korea!

What trust can the general public have in the safeguards

and monitoring that the IAEA is tasked to carry out

anywhere in the face of the US administration’s actions?

Worse still, what example does the actions of the US set

for other nations in terms of their relationships and

agreements with the international nuclear community?

President Trump has in the past alluded to his admiration

for one predecessor in particular – President Dwight

Eisenhower. If so, then all may not be lost, because the

genesis of the IAEA was Eisenhower’s ‘Atoms for Peace’

address to the United Nations in 1953.

Whatever the outcome of the talks with Chairman Kim,

I hope someone in the US administration will be bold

enough to get President Trump to reflect on the Atoms for

Peace ideals.

Mr Amano has warned in recent weeks that if the JCPOA

were to fail, “it would be a great loss for nuclear verification

and for multilateralism”.

And for its part, Iran’s ambassador to the European

Union, Peiman Seadat, has been quoted as saying “years of

intense multilateral diplomacy are now on critical life

support and without effective and timely intervention, the

JCPOA will simply die”.

Perhaps we could all remind the US president of the

IAEA’s work for the greater nuclear good via his communications

channel of choice? Feel free to Tweet a courteous

but steadfast defence of the Agency to @realDonaldTrump.

Author

John Shepherd

Shepherd Communications

3 Brooklands

West Sussex

BN43 5FE

Nuclear Today

Confidence in Nuclear Safeguards at Risk as Trump Quits One Deal to Pursue Another ı John Shepherd


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