atw 2018-02

nucmag.com 

2018 

2 

81 

Gas Cooled 

Reactor Development 

in China 

85 ı Environment and Safety 

Severe Accident Safety Research for Reactor Buildings 

95 ı Operation and New Build 

Knowledge Management and TRIZ for Safe Shutdown Capability 

ISSN · 1431-5254 

24.– € 

104 ı Decommissioning and Waste Management 

Corrosion Processes of Alloyed Steels in Salt Solutions 

134 ı Nuclear Today 

Playing Politics with Nuclear is all Part of the Game

atw Vol. 63 (2018) | Issue 2 ı February 

Some Questions and Answers 

About Energy 

Dear Reader, The question is always on the agenda whether people are really aware about facts on energy. 

The following energy quiz with 12 questions should point out some interesting facts. The answers are given on page 132 

of this issue of atw. 

1. True or false: 

The global energy demand will 

decrease in the next decades! 

a. True 

b. False 


The global electricity demand will 

decrease in the next decades! 

a. True 

b. False 


The global coal production 

is always decreasing! 

a. True 

b. False 

4. What percentage of world’s electricity 

production was produced from nuclear 

in 2017? 

a. 1 % 

b. 6 % 

c. 11 % 

d. 20 % 

8. Which technology has the lowest 

CO 2 footprint? 

a. Photovoltaics 

b. Wind 

c. Nuclear 

d. Hydropower 

9. What energy source has Bill Gates 

invested in, and championed, over the 

last few years? 

a. Nuclear power 

b. Photovoltaics 

c. Wind energy 

d. Tidal energy 

10. What energy source has the smallest 

number of lost lifetime-days 

(due to health hazards and accidents) 

per kilowatt-hour produced? 

a. Coal 

b. Natural gas 

c. Wind 

d. Nuclear power 

11. What subjects someone 

to the most radiation? 

71 

EDITORIAL 

5. What percentage of world’s electricity 

production was produced from wind plus 

solar in 2017? 

a. 1 % 

b. 5 % 

c. 10 % 

d. 20 % 

6. Which country has the most 

fossil fuel resources? 

a. Saudi Arabia 

b. Russia 

c. United States of America 

d. China 

e. EU 

7. What country/region will emit the most 

carbon dioxide in 2018? 

a. Living next to a nuclear power plant. 

b. Flying from Europe to other continents 

c. Eating a 250 g bag of potato chips 

every day 

d. Living in Guarapari, Brazil 


The number of nuclear power plants 

worldwide will decrease in the future. 

a. True 

b. False 

Christopher Weßelmann 

– Editor in Chief – 

a. United States of America 

b. Nigeria 

c. EU 

d. China 

Editorial 

Some Questions and Answers About Energy


72 

EDITORIAL 

Einige Fragen und Antworten 

zum Thema Energie 

Liebe Leserin, lieber Leser, Diskussion über das Thema Energie wird häufig die Frage aufgeworfen, inwieweit 

diese von Fakten bestimmt wird bzw. die Fakten überhaupt bekannt sind. Das folgende Energiequiz soll mit seinen 

12 Fragen einige interessante Fakten aufzeigen. Die Antworten finden Sie auf Seite 132 dieser Ausgabe der atw. 

1. Richtig oder falsch: 

Der globale Energiebedarf wird 

in den nächsten Jahrzehnten sinken! 

a. Wahr 

b. Falsch 


Der weltweite Strombedarf wird 

in den nächsten Jahrzehnten sinken! 

a. Wahr 

b. Falsch 


Die weltweite Kohleförderung nimmt ab! 

8. Welche Technologie hat den niedrigsten 

CO 2 -Fußabdruck? 

a. Photovoltaik 

b. Wind 

c. Kernenergie 

d. Wasserkraft 

9. In welche Energiequelle hat Bill Gates 

in den letzten Jahren investiert und sich 

dafür öffentlich eingesetzt? 

a. Kernkraft 

b. Photovoltaik 

c. Windenergie 

d. Gezeitenenergie 

a. Wahr 

b. Falsch 

4. Welchen Anteil hatte die Kernenergie 

an der weltweiten Stromproduktion 

im Jahr 2017? 

a. 1 % 

b. 6 % 

c. 11 % 

d. 20 % 

5. Welchen Anteil hatten Wind und Sonne 

an der weltweiten Stromproduktion 

im Jahr 2017? 

a. 1 % 

b. 5 % 

c. 10 % 

d. 20 % 

6. Welches Land verfügt über die größten 

fossilen Energieressourcen? 

a. Saudi-Arabien 

b. Russland 

c. Vereinigte Staaten von Amerika 

d. China 

e. EU 

7. Welches Land bzw. welche Region 

wird 2018 die höchsten Kohlendioxidemissionen 

verzeichnen? 

a. Vereinigte Staaten von Amerika 

b. Nigeria 

c. EU 

d. China 

10. Welche Energiequelle verzeichnet die 

geringste Anzahl an Ausfalltagen 

(aufgrund von Gesundheitsgefahren 

und Unfällen) pro produzierter Kilowattstunde? 

a. Kohle 

b. Erdgas 

c. Wind 

d. Kernkraft 

11. Was verursacht die höchste 

Strahlenbelastung? 

a. Wohnen neben einem Kernkraftwerk. 

b. Fliegen von Europa 

zu anderen Kontinenten 

c. Täglich 250 g Chips essen 

d. Leben in Guarapari, Brasilien 


Die Zahl der Kernkraftwerke weltweit 

wird in Zukunft abnehmen. 

a. Wahr 

b. Falsch 

Christopher Weßelmann 

– Chefredakteur – 

Editorial 

Einige Fragen und Antworten zum Thema Energie

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74 

Issue 2 

February 

CONTENTS 

81 

Gas Cooled 

Reactor Development 

in China 

| | Outside view of the two boiling water reactors at the Olkiluoto site in Finland. The reactors with a gross electric output of 910 MWe each 

are successfully operated by Teollisuuden Voima Oyj – TVO. Ever since the early 1990s, the OL1 and OL2 capacity factors have remained 

between 93 and 97 percent. (Courtesy: TVO) 

Editorial 

Some Questions and Answers 

About Energy 71 

Einige Fragen und Antworten 

zum Thema Energie 72 

Abstracts | English 76 

Abstracts | German 77 

Calendar . . . . . . . . . . . . . . . . . . . . . . . .80 

Energy Policy, Economy and Law 

Development of High Temperature 

Gas Cooled Reactor in China 81 

Wentao Guo and Michael Schorer 

Spotlight on Nuclear Law 

The Liability According to § 26 of the 

German Atomic Energy Act – A Wallflower? 84 

Die Haftung nach § 26 AtG – 

ein Mauerblümchen? 84 

Christian Raetzke 

81 

| | The construction of Shidao Bay HTGR. 

Inside Nuclear with NucNet 

WANO to Increase Focus on New Nuclear as 

Industry’s Centre of Gravity Shifts Towards Asia 78 

85 

NucNet 

| | COCOSYS nodalisation scheme. 

DAtF Notes. . . . . . . . . . . . . . . . . . . . . . 79 

Contents


Environment and Safety 

Investigation of Conditions Inside the Reactor 

Building Annulus of a PWR Plant of KONVOI 

Type in Case of Severe Accidents with Increased 

Containment Leakages 85 

Ivan Bakalov and Martin Sonnenkalb 

Sensitivity Analysis of MIDAS Tests 

Using SPACE Code: Effect of Nodalization 90 

Shin Eom, Seung-Jong Oh and Aya Diab 

75 

CONTENTS 

90 

Operation and New Build 

The Application of Knowledge Management 

and TRIZ for solving the Safe Shutdown Capability 

in Case of Fire Alarms in Nuclear Power Plants 95 

Chia-Nan Wang, Hsin-Po Chen, Ming-Hsien Hsueh and Fong-Li Chin 

95 

| | Isometric View of the MIDAS Facility. 

| | Application of knowledge management and TRIZ. 

Decommissioning and Waste Management 

Corrosion Processes of Alloyed Steels 

in Salt Solutions 104 

Bernhard Kienzler 

Research and Innovation 

Design and Development of a Radio eco logical 

Domestic User Friendly Code for Calculation 

of Radiation Doses and Concentration 

due to Airborn Radio nuclides Release During 

the Accidental and Normal Operation 

in Nuclear Installations 111 

|104 

111 

| | Localized corrosion phenomena of steel 1.4306. 

Events 

Event Report: 

Nuklearforum Schweiz – Future Management 

– Key Solutions for Nuclear Facilities 121 

Event Report: 

Nuklearforum Schweiz – Zukunftsmanagement 

– zentrale Lösungsansätze für Kernanlagen 121 

Matthias Rey 

KTG Inside . . . . . . . . . . . . . . . . . . . . . . 123 

News . . . . . . . . . . . . . . . . . . . . . . . . . 129 

Nuclear Today 

Playing Politics with Nuclear 

is All Part of the Game 134 

John Shepherd 

Imprint 110 

| Summary of Code Algorithms. 

A. Haghighi Shad, D. Masti, M. Athari Allaf, K. Sepanloo, S.A.H. Feghhi 

and R. Khodadadi 

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

Contents


76 

ABSTRACTS | ENGLISH 

WANO to Increase Focus on New Nuclear 

as Industry’s Centre of Gravity Shifts 

Towards Asia 

NucNet | Page 78 

The World Association of Nuclear Operators 

(WANO) intends to focus more on new nuclear units 

coming into operation around the world as the 

“ centre of gravity” in the industry shifts from the US 

and Europe to the Middle East and Asia. The 

organisation’s chief executive officer, Peter Prozesky, 

told NucNet that new-build projects in China, India, 

Turkey and the United Arab Emirates are giving 

WANO the opportunity to make sure those countries 

start the operational life of their new units “in a very 

positive way”. In supporting countries with new 

units beginning operation, WANO is working more 

closely with the International Atomic Energy Agency 

(IAEA). One of the IAEA’s tasks is to help emerging 

nuclear countries develop the infrastructure and 

capability they need to have nuclear power as part of 

their energy mix. 

Development of High Temperature Gas 

Cooled Reactor in China 

Wentao Guo and Michael Schorer | Page 81 

High temperature gas cooled reactor (HTGR) is one 

of the six Generation IV reactor types put forward 

by Generation IV International Forum (GIF) in 

2002. This type of reactor has high outlet temperature. 

It uses Helium as coolant and graphite as 

moderator. Pebble fuel and ceramic reactor core are 

adopted. Inherit safety, good economy, high generating 

efficiency are the advantages of HTGR. 

According to the comprehensive evaluation from 

the international nuclear community, HTGR has 

already been given the priority to the research and 

development for commercial use. A demonstration 

project of the High Temperature Reactor-Pebblebed 

Modules (HTR-PM) in Shidao Bay nuclear 

power plant in China is under construction. In this 

paper, the development history of HTGR in China 

and the current situation of HTR-PM will be introduced. 

The experiences from China may be taken as 

a reference by the international nuclear community. 

The Liability According to § 26 of the 

German Atomic Energy Act – A Wallflower? 

Christian Raetzke | Page 84 

According to German law, liability for damage 

caused by radioactivity can arise from several 

regulation. In most cases, liability under the Paris 

Convention on Third Party Liability in the Field of 

Nuclear Energy, which applies in the field of nuclear 

power, is at the forefront of discussion. According to 

§ 26 of the German Atomic Energy Act, liability is 

somewhat in the shadow of the Paris Convention. It 

applies to the handling of radioactivity in medicine, 

research and industry (e. g. for test emitters) as well 

as activities involving natural and depleted uranium 

and nuclear fusion. The article outlines the basic 

elements of liability under Section 26 of the German 

Atomic Energy Act, which may become increasingly 

important in future due to recent developments 

such as the phasing out of nuclear power in 

Germany. 

Investigation of Conditions Inside the 

Reactor Building Annulus of a PWR Plant of 

KONVOI Type in Case of Severe Accidents 

with Increased Containment Leakages 

Ivan Bakalov and Martin Sonnenkalb | Page 85 

Improvements of the implemented severe accident 

management (SAM) concepts have been done in all 

operating German NPPs after the Fukushima Daiichi 

accidents following recommendations of the 

German Reactor Safety Commission (RSK) and as a 

result of the stress test being performed. The 

efficiency of newly developed severe accident 

management guidelines (SAMG) for a PWR KONVOI 

reference plant related to the mitigation of challenging 

conditions inside the reactor building (RB) 

annulus due to increased containment leakages 

during severe accidents have been assessed. Based 

on two representative severe accident scenarios the 

releases of both hydrogen and radionuclides into the 

RB annulus have been predicted with different 

boundary conditions. The accident scenarios have 

been analysed without and with the impact of 

several SAM measures (already planned or proposed 

in addition), which turned out to be efficient to 

mitigate the consequences. The work was done 

within the frame of a research project financially 

supported by the Federal Ministry BMUB. 

Sensitivity Analysis of MIDAS Tests Using 

SPACE Code: Effect of Nodalization 

Shin Eom, Seung-Jong Oh and Aya Diab | Page 90 

The nodalization sensitivity analysis for the ECCS 

(Emergency Core Cooling System) bypass phenomena 

was performed using the SPACE (Safety 

and Performance Analysis CodE) thermal hydraulic 

analysis computer code. The results of MIDAS 

(Multi- dimensional Investigation in Downcomer 

Annulus Simulation) test were used. The MIDAS 

test was conducted by the KAERI (Korea Atomic 

Energy Research Institute) for the performance 

evaluation of the ECC (Emergency Core Cooling) 

bypass phenomenon in the DVI (Direct Vessel 

Injection) system. The main aim of this study is to 

examine the sensitivity of the SPACE code results 

to the number of thermal hydraulic channels 

used to model the annulus region in the MIDAS 

experiment. The numerical model involves three 

nodalization cases (4, 6, and 12 channels) and 

the result show that the effect of nodalization 

on the bypass fraction for the high steam flow rate 

MIDAS tests is minimal. For computational 

efficiency, a 4 channel representation is recommended 

for the SPACE code nodalization. For the 

low steam flow rate tests, the SPACE code overpredicts 

the bypass fraction irrespective of the 

nodalization finesse. The over- prediction at low 

steam flow may be attributed to the difficulty 

to accurately represent the flow regime in the 

vicinity of the broken cold leg. 

The Application of Knowledge 

Management and TRIZ for solving 

the Safe Shutdown Capability in Case of 

Fire Alarms in Nuclear Power Plants 

Chia-Nan Wang, Hsin-Po Chen, 

Ming-Hsien Hsueh and Fong-Li Chin | Page 95 

The Fukushima nuclear disaster in 2011 has raised 

widespread concern over the safety of nuclear 

power plants. This study employed knowledge 

management in conjunction with the Teoriya 

Resheniya Izobreatatelskih Zadatch (TRIZ) method 

in the formulation of a database to facilitate the 

evaluation of post-fire safe shutdown capability 

with the aim of safeguarding nuclear facilities in the 

event of fire. The proposed approach is meant to 

bring facilities in line with US Nuclear Regulatory 

Commission (NRC) standards. When implemented 

in a case study of an Asian nuclear power plant, our 

method proved highly effective in the detection of 

22 cables that fell short of regulatory requirements, 

thereby reducing 850,000 paths to 0. This study 

could serve as reference for industry and academia 

in the development of systematic approaches to the 

upgrading of nuclear power plants. 


in Salt Solutions 

Bernhard Kienzler | Page 104 

A summary is given of the corrosion experiments 

with alloyed Cr-Ni steels in salt solutions performed 

at Research Centre Karlsruhe (today KIT), Institute 

for Nuclear Waste Disposal (INE) in the period 

between 1980 and 2004. Alloyed steels show 

significantly lower general corrosion in comparison 

to carbon steels. However, especially in salt brines 

the protective Cr oxide layers on the surfaces of 

these steels are disturbed and localized corrosion 

takes place. Data on general corrosion rates, and 

findings of pitting, crevice and stress corrosion 

cracking are presented. 

Design and Development of a Radioecological 

Domestic User Friendly Code for 

Calculation of Radiation Doses and Concentration 

due to Airborn Radionuclides 

Release During the Accidental and Normal 

Operation in Nuclear Installations 

A. Haghighi Shad, D. Masti, 

M. Athari Allaf, K. Sepanloo, 

S.A.H. Feghhi and R. Khodadadi | Page 111 

A domestic user friendly dynamic radiological dose 

and model has been developed to estimate radiation 

doses and stochastic risks due to atmospheric and 

liquid discharges of radionuclides in the case of a 

nuclear reactor accident and normal operation. In 

addition to individual doses from different pathways 

for different age groups, collective doses and 

stochastic risks can be calculated by the developed 

domestic user friendly KIANA Advance Computational 

Computer Code and model. The current Code 

can be coupled to any long-range atmospheric 

dispersion/short term model which can calculate 

radionuclide concentrations in air and on the 

ground and in the water surfaces predetermined 

time intervals or measurement data. 

Event Report: Future Management – 

Key Solutions for Nuclear Facilities 

Matthias Rey | Page 121 

Future management requires careful planning and 

knowledge of what options are available, how far 

optimizations make sense and which measures and 

process changes have already proven themselves 

elsewhere. The 2017 advanced course of the Swiss 

Nuclear Forum took up this topic. On the first day 

of the course, the focus was on solutions for 

optimizing system operation and maintenance. The 

second day focused on the employees in their 

changing environment. As a novelty this year, the 

topics of the morning input presentations were 

discussed in depth in workshops on both afternoons. 

Playing Politics with Nuclear is all Part 

of the Game 

John Shepherd | Page 134 

If a week is a long time in politics – a statement 

attributed to former British prime minister Harold 

Wilson – then what about a month, or several 

months – a period relevant for the use of nuclear 

power? The nuclear industry has long accepted that 

it can be used as a political football, to be kicked into 

goal or off the pitch completely depending on the 

situation at hand. Our industry therefore has power 

in the political sense too, but with power comes 

responsibility – nuclear leaders know that only too 

well and now is as good as time as ever to lead by 

example. 

Abstracts | English


WANO wird sich mit der Verlagerung der 

Aktivitäten nach Asien verstärkt auf den 

Kernkraftwerksneubau konzentrieren 

NucNet | Seite 78 

Die World Association of Nuclear Operators 

( WANO) will sich verstärkt auf Kernkraftwerksneubauten 

konzentrieren, da sich der „Schwerpunkt“ 

der Branche von den USA und Europa in den Nahen 

Osten und nach Asien verlagert. Peter Prozesky, 

Chief Executive Officer von WANO, erläuterte, dass 

Neubauprojekte in China, Indien, der Türkei und 

den Vereinigten Arabischen Emiraten WANO die 

Möglichkeit geben, dass diese Länder mit den 

Erfahrungen von WANO in die Kernenergie einsteigen. 

Bei der Unterstützung von Ländern, in 

denen neue Anlagen in Betrieb genommen werden, 

arbeitet WANO eng mit der Internationalen Atomenergie-Organisation 

(IAEO) zusammen. Eine der 

Aufgaben der IAEO besteht darin, die zuküftigen 

Nuklearstaaten darin zu unterstützen, die Infrastruktur 

und das Know-how zu entwickeln, das 

sie benötigen, um die Kernenergie als Teil ihres 

Energiemixes zu nutzen. 

Entwicklung des gasgekühlten 

Hochtemperaturreaktors in China 

Wentao Guo und Michael Schorer | Seite 81 

Der gasgekühlte Hochtemperaturreaktor (HTGR) ist 

einer von sechs Reaktortypen der Generation IV, die 

2002 vom Generation IV International Forum (GIF) 

vorgestellt wurde. Charakteristisch für diesen Reaktortyp 

sind die hohe Kühlmittelaustrittstem peratur 

aus dem Reaktor, Helium als Kühlmittel, Graphit 

als Moderator, kugelförmige Brenn elemente sowie 

keramischer Reaktorkernein bauten. Vorteile von 

HTGR sind inhärente Sicherheit, Wirtschaftlichkeit 

sowie hohe Effizienz der Brennstoffnutzung. Nach 

einer umfassenden Eva luierung durch hat die Entwicklung 

von HTGR bis hin zur kommerziellen 

Nutzung Priorität. Ein Demonstrationsprojekt für 

einen HTR-Modul reaktor befindet sich am Standort 

Shidao Bay in China in Bau. In diesem Beitrag 

werden die Entwicklungsgeschichte von HTGR in 

China und die aktuelle Situation der HTR-PM- 

Projekte vor gestellt. Die Erfahrungen aus China sind 

eine international nutzbare Referenz. 

Die Haftung nach § 26 AtG – 

ein Mauerblümchen? 

Christian Raetzke | Seite 84 

Die Haftung für Schäden aus Radioaktivität kann 

sich nach deutschem Recht aus mehreren Quellen 

ergeben. In der Diskussion steht meist die Haftung 

nach dem Pariser Übereinkommen (PÜ) im Vordergrund, 

die im Bereich der Kernenergie gilt. Etwas 

im Schatten des PÜ steht die Haftung nach § 26 AtG. 

Sie gilt für den Umgang mit Radioaktivität im 

Bereich der Medizin, Forschung und Industrie 

( etwa bei Prüfstrahlern) sowie für Aktivitäten rund 

um natürliches und abgereichertes Uran und für die 

Kernfusion. Der Artikel skizziert die Grund elemente 

der Haftung nach § 26 AtG, die aufgrund jüngerer 

Entwicklungen wie dem Kernenergieausstieg in 

Deutschland möglicherweise künftig an Bedeutung 

gewinnen wird. 

Untersuchungen zu den Zuständen im 

Ringraum des Reaktorgebäudes eine DWR 

vom Typ KONVOI im Falle von schweren 

Störfällen mit erhöhten Leckagen aus dem 

Containment 

Ivan Bakalov and Martin Sonnenkalb | Seite 85 

Die anlageninternen Notfallschutzkonzepte der in 

Betrieb befindlichen KKW in Deutschland wurden 

nach den Unfällen in Fukushima Daiichi verbessert 

und damit Empfehlungen der Reaktorsicherheitskommission 

(RSK) und neue Erkenntnisse aus den 

Stress Tests umgesetzt. Die Wirksamkeit von neu 

entwickelten Maßnahmen des mitigativen Notfallschutzes 

für eine DWR-Referenzanlage vom Typ 

KONVOI hinsichtlich der Zustände im Ringraum 

des Reaktorgebäudes bei erhöhten Leckagen aus 

dem Containment während schwerer Störfälle 

wurde analysiert. Die Freisetzung von Wasserstoff 

und Radionukliden in den Ringraum des Reaktorgebäudes 

wurde an Hand von zwei repräsentativen 

schweren Störfallszenarien unter der Annahme 

unterschiedlicher Randbedingungen untersucht. 

Die Analysen wurden ohne und mit mitigativen 

Notfallmaßnahmen (bereits umgesetzte oder 

zusätzliche Maßnahmen) durchgeführt, und die 

Ergebnisse bestätigten die Wirksamkeit aller Maßnahmen. 

Die Arbeiten wurden im Rahmen eines 

Forschungsprojektes der GRS finanziell unterstützt 

vom BMUB durchgeführt. 

Sensitivitätsanalyse von MIDAS-Tests mit 

SPACE-Code: Auswirkung der Nodalisierung 

Shin Eom, Seung-Jong Oh und Aya Diab | Seite 90 

Die Sensitivitätsanalyse zur Nodalisierung für die 

Bypass-Phänomene des ECCS (Emergency Core 

Cooling System) wurde mit Hilfe des thermo hydraulischen 

Analyse-Computercodes SPACE ( Safety and 

Performance Analysis CodE) durchgeführt. Dazu 

wurden die Ergebnisse des MIDAS-Tests (Multidimensional 

Investigation in Downcomer Annulus 

Simulation) verwendet. Der MIDAS-Test wurde vom 

KAERI (Korea Atomic Energy Research Institute) zur 

Leistungsbewertung des ECC ( Emergency Core 

Cooling) Bypass-Phänomens im DVI (Direct Vessel 

Injection) System durchgeführt. Das Hauptziel dieser 

Studie ist es, die Sensitivität der SPACE-Code-Ergebnisse 

für die thermo hydrau lischen Unterkanäle zu 

untersuchen, die zur Modellierung des Ringraums im 

MIDAS- Experiment verwendet werden. Aus Gründen 

der Rechen effizienz wird für die SPACE-Code- 

Nodalisierung eine 4-Kanal-Darstellung empfohlen. 

Knowledge Management und TRIZ 

für die Sicherstellung der Abschaltfähigkeit 

bei Feueralarmen in Kernkraftwerken 

Chia-Nan Wang, Hsin-Po Chen, 

Ming-Hsien Hsueh und Fong-Li Chin | Seite 95 

Die Katastrophe von Fukushima im Jahr 2011 hat 

die Frage nach der Sicherheit von Kernkraftwerken 

erneut gestellt. In dieser Studie wurde Wissensmanagement 

in Verbindung mit der Teoriya Resheniya 

Izobreatatelskih Zadatch (TRIZ) Methode bei der 

Formulierung einer Datenbank eingesetzt, um die 

Bewertung der Fähigkeit zur sicheren Abschaltung 

nach einem Brand in einem Kernkraftwerk zu 

ermöglichen. Der vorgeschlagene Ansatz zielt 

darauf ab, die Anlagen mit den Standards der 

US Nuclear Regulatory Commission (NRC) in 

Einklang zu bringen. Bei der Implementierung in 

einer Fallstudie eines asiatischen Kernkraftwerks 

erwies sich die Methode als sehr effektiv bei der 

Feststellung von 22 Kabeln, die nicht den vorgegebenen 

Anforderungen entsprachen, wodurch 

850.000 mögliche Ereignispfade auf 0 reduziert 

wurden. Diese Studie kann auch als Referenz 

dienen für die Entwicklung systematischer Ansätze 

zur weiteren Modernisierung von Kernkraftwerken. 

Korrosionprozesse legierter Stähle 

in Salzlösungen 

Bernhard Kienzler | Seite 104 

Es wird eine Zusammenfassung der Experimente 

zur Korrosion von legierten Cr-Ni Stählen in 

Salzlösungen vorgestellt. Die Experimente wurden 

Im Forschungszentrum Karlsruhe (heute KIT), 

Institut für Nukleare Entsorgung (INE) im Zeitraum 

zwischen 1980 und 2004 durchgeführt. Legierte 

Stähle zeigten eine deutlich geringere Flächenkorrosion 

im Vergleich zu den ebenfalls untersuchten 

Kohlenstoffstählen. Jedoch findet in den 

Salzlösungen eine Störung der Korrosionsschutzschichten 

aus Cr-Oxiden auf den Stahloberflächen 

statt, die zu lokalen Korrosionsprozessen führt. 

Flächenkorrosionsraten und die Beobachtungen 

hinsichtlich Lochfrass-, Spalt- und Spannungsrißkorrosion 

werden aufgezeigt. 

Entwicklung eines Codes zur Berechnung 

der Strahlendosis und -konzentration bei 

Freisetzung von luftgetragenen Radionukliden 

während des unfallbedingten 

und normalen Betriebes kerntechnischer 

Anlagen 

A. Haghighi Shad, D. Masti, 

M. Athari Allaf, K. Sepanloo, 

S.A.H. Feghhi und R. Khodadadi | Seite 111 

Zur Abschätzung von Strahlendosen und stochastischen 

Risiken durch atmosphärische und flüssige 

Radionuklidemissionen bei einem Reaktorunfall 

und im Normalbetrieb wurde ein benutzerfreundliches 

dynamisches radiologisches Freisetzungs- und 

Dosismodell entwickelt. Zusätzlich zu den Einzeldosen 

aus verschiedenen Pfaden für verschiedene 

Nuklide können Kollektivdosen und stochastische 

Risiken mit Hilfe des entwickelten benutzerfreundlichen 

KIANA Advance Computational Computer 

Codes und Modells berechnet werden. Der aktuelle 

Code kann mit jedem weiträumigen atmosphärischen 

Ausbreitungs-/Kurzzeitmodell gekoppelt 

werden, mit dem Radionuklidkonzentrationen 

in der Luft und am Boden und in Gewässern 

berechnet werden können. 

Tagungsbericht: Zukunftsmanagement – 

zentrale Lösungsansätze für Kernanlagen 

Matthias Rey | Seite 121 

Zukunftsmanagement erfordert sorgfältige Planung 

und Wissen darüber, welche Optionen zur Verfügung 

stehen, wieweit Optimierungen sinnvoll 

sind und welche Maßnahmen und Prozessänderungen 

sich allenfalls bereits anderswo 

bewährt haben. Der Vertiefungskurs 2017 des 

Nuklearforums Schweiz nahm diese Thematik auf. 

Im Zentrum standen Lösungsansätze zum Optimieren 

von Systembetrieb und Instandhaltung 

sowie die Mitarbeitenden in ihrer sich verändernden 

Umwelt. Als Novum wurden die Themen 

der Inputreferate des Vormittags in Workshops 

vertieft diskutiert. 

Mit der Kernenergie zu spielen 

ist Teil der Politik 

John Shepherd | Seite 134 

Eine Woche ist in der Politik eine lange Zeit! Dieser 

Satz wird dem ehemaligen britischen Premierminister 

Harold Wilson zugeschrieben. Was ist 

dann mit einem Monat oder mehreren Monaten, 

wie sie für eine langfristige Technologie wie der 

Kernenergie bestimmend sind? Die kerntechnische 

Industrie hat längst akzeptiert, dass sie als politischer 

Spielball genutzt werden kann, um je nach 

Situation ins Tor oder vom Spielfeld geschossen zu 

werden. „Nuklearpolitiker“ wissen, dass Entscheidungen 

zur Kernenergie nicht nur „Macht“ 

bedeuten, sondern auch Verantwortung. Heute 

geht es deshalb darum hier mit gutem Beispiel 

voranzugehen. 

77 

ABSTRACTS | GERMAN 

Abstracts | German


78 

INSIDE NUCLEAR WITH NUCNET 

WANO to Increase Focus on New 

Nuclear as Industry’s Centre of Gravity 

Shifts Towards Asia 

NucNet 

The World Association of Nuclear Operators (WANO) intends to focus more on new nuclear units coming 

into operation around the world as the “centre of gravity” in the industry shifts from the US and Europe to 

the Middle East and Asia. 

The organisation’s chief executive officer, Peter Prozesky, 

told NucNet that new-build projects in China, India, Turkey 

and the United Arab Emirates are giving WANO the 

opportunity to make sure those countries start the 

operational life of their new units “in a very positive way”. 

He said the rate of new-build in these new nuclear 

markets means there could be challenges, even for existing 

companies, related to rapid expansion. There could be 

challenges to the ability of some expanding companies 

to provide experienced and qualified people to staff their 

new units, he said. 

In supporting countries with new units beginning 

operation, WANO is working more closely with the 

International Atomic Energy Agency (IAEA). One of the 

IAEA’s tasks is to help emerging nuclear countries develop 

the infrastructure and capability they need to have nuclear 

power as part of their energy mix. 

Mr Prozesky said WANO, whose members operate some 

440 nuclear reactor units in more than 30 countries, has 

developed a strong relationship between its London office 

and IAEA headquarters in Vienna to ensure that experience 

is regularly shared. He said: “The IAEA gets involved with 

new entrants a lot earlier than we do. They are focusing on 

member countries and setting up infrastructure, while 

WANO needs to engage when new-build contracts get 

signed. The aim is now to have WANO involved as early as 

possible.” 

WANO is developing training modules and support 

missions for new nuclear countries. Modules cover the 

period from the start of contractual work to commercial 

operation, and aim to help utilities and companies during 

the construction and commissioning phases. Early engagement 

with the IAEA is part of WANO’s Compass plan, which 

was conceived in 2015 and updated at this year’s biennial 

general meeting, in Gyeongju, South Korea. 

The revised schedule for Compass, which also includes 

plans to make WANO more effective in areas such as 

life-extensions and decommissioning of plants, is 2022. 

The original Compass ran until 2019, but that target has 

now been revised, Mr Prozesky said. 

Earlier this year the IAEA and WANO agreed to increase 

their cooperation to strengthen operational safety and to 

support countries that are planning or considering 

launching nuclear power programmes. They said they 

can maximise safety benefits, increase efficiency and 

avoid conflicting advice by increasing cooperation on 

safety peer review services. 

Increasing the efficiency of the reviews will be particularly 

important in anticipation of the increasing number of 

nuclear facilities worldwide in coming decades, WANO 

chairman Jacques Regaldo said at the time. “By 2030, half 

of the nuclear power reactors will be based in Asia, and we 

will have many newcomers to nuclear power,” he said. 

“There is real value for WANO to work together with the 

IAEA and others to help maximise the safety and reliability 

of nuclear power plants.” 

In an August 2017 report the IAEA said it foresees a 

significant decline in nuclear expansion in North America 

and in northern, western and southern Europe, with only 

slight increases in Africa and western Asia. 

But significant growth is projected in central and 

eastern Asia, where nuclear power capacity is expected to 

undergo an increase of 43 % by 2050. 

WANO has been discussing plans for a new regional 

centre in Asia to meet demand for expertise and missions 

from companies operating new units. The organisation 

already has regional centres in Atlanta, Moscow, Paris and 

Tokyo, with a head office in London. 

WANO has decided to look into the possibility of setting 

up a new regional centre, starting with a proposal to open 

a branch of the London office in Shanghai. The main aim of 

this office will be to develop local expertise. 

The second phase of opening a new regional centre 

would then include converting the branch office into a 

support centre which would provide support services to 

other regions. These initial preparations depend on a vote 

by WANO members, probably in 2018. When the support 

centre is operating as it should, it would become a fully 

operational regional centre. 

Mr Prozesky said WANO is holding discussions with 

its Chinese members about “the sharing of financial 

responsibility” for funding the Shanghai office through the 

first two phases. 

At its biennial general meeting, WANO discussed the 

implications of financial and market pressures. Corporate 

organisations “have huge responsibilities” to ensure that 

operating nuclear plants are carefully managed and 

adequately resourced in these difficult times, Mr Prozesky 

said. 

The organisation also started a discussion on how it 

should be supporting units when they approach the end of 

their designed lifetime. 

Members spoke about the need to increase cooperation 

amongst like-minded organisations such as the IAEA and 

the Paris-based Nuclear Energy Agency. 

WANO recently announced the signing of a cooperation 

agreement with the International Youth Nuclear Congress 

(IYNC), recognition of the fact that WANO needs to find 

ways to transfer knowledge from people who have been in 

the industry for the past 40 years to those who are entering 

it today. 

Mr Prozesky said it was “quite sobering” to talk to young 

operators in control rooms today and find that some of 

them weren’t born when the Chernobyl accident happened 

in 1986. He said: “It is essential that transfer all the 

accumulated knowledge and the industry’s experience to 

Inside Nuclear with NucNet 

WANO to Increase Focus on New Nuclear as Industry’s Centre of Gravity Shifts Towards Asia ı NucNet


the new generation. We must find out how to make the 

industry attractive to the younger generation.” 

Mr Prozesky said members have asked WANO “to do a 

little bit more” on providing support as opposed to just 

carrying out assessments of their businesses. He said 

another point in the updated Compass document is 

associated with putting more energy into leadership 

development. “We find in our assessment process across 

the world, when looking at corporate organisations and 

power plants, that there is a need for WANO to develop 

products and services aimed at creating leaders for the 

nuclear industry. 

“So, we will be putting some energy into that over the 

next four years. Particularly again, the focus and emphasis 

will be on new entrants and new units, but there is an 

overall need for developing leadership in the rest of the 

world as well.” 

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 

DATF EDITORIAL NOTES 

79 

Notes 

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DAtF Notes


80 

CALENDAR 

Calendar 

2018 

05.02.-07.02.2018 

Components and Structures under Severe 

Accident Loading Cossal (COSSAL). 

Cologne, Germany. OECD/NEA, GRS, 

www.grs.de, www.oecd-nea-org 

07.02.-08.02.2018 

8. Symposium Stilllegung und Abbau 

kerntechnischer Anlagen. Hanover, Germany. 

TÜV Nord, www.tuev.nord.de 

26.02.-01.03.2018 

Nuclear and Emerging Technologies for Space 

2018. Las Vegas, NV, USA. American Nuclear Society 

(ANS), www.ans.org 


7. Fachgespräch Endlagerbergbau. Essen, 

Germany, DMT, GNS, www.dmt-goup.com 

04.03.-09.03.2018 

82. Jahrestagung der DPG. Erlangen, Germany, 

Deutsche Physikalische Gesellschaft (DPG), 

www.dpg-physik.de 

11.03.-17.03.2018 

International Youth Nuclear Congress (IYNC). 

Bariloche, Argentina, IYNC and WiN Global, 

www.iync.org/category/iync2018/ 

26.03.-27.03.2018 

Fusion energy using tokamaks: can development 

be accelerated? London, United Kingdom, 

The Royal Society, royalsociety.org 

08.04.-11.04.2018 

International Congress on Advances in Nuclear 

Power Plants – ICAPP 18. Charlotte, NC, USA, 

American Nuclear Society (ANS), www.ans.org 

08.04.-13.04.2018 

11 th International Conference on Methods and 

Applications of Radioanalytical Chemistry – 

MARC XI. Kailua-Kona, HI, USA, American Nuclear 

Society (ANS), www.ans.org 

17.04.-19.04.2018 

World Nuclear Fuel Cycle 2018. Madrid, Spain, 

World Nuclear Association (WNA), 

www.world-nuclear.org 

18.04.-19.04.2018 

9. Symposium zur Endlagerung radioaktiver Abfälle. 

Vorbereitung auf KONRAD – Wege zum G2- 

Gebinde. Hanover, Germany, TÜV NORD Akademie, 

www.tuev-nord.de/tk-era 

22.04.-26.04.2018 

Reactor Physics Paving the Way Towards More 

Efficient Systems – PHYSOR 2018. Cancun, Mexico, 

www.physor2018.mx 

08.05.-10.05.2018 

29 th Conference of the Nuclear Societies in Israel. 

Herzliya, Israel. Israel Nuclear Society and Israel 

Society for Radiation Protection, ins-conference.com 

13.05.-19.05.2018 

BEPU-2018 – ANS International Conference on 

Best-Estimate Plus Uncertainties Methods. Lucca, 

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

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

13.05.-18.05.2018 

RadChem 2018 – 18 th Radiochemical Conference. 

Marianske Lazne, Czech Republic, 

www.radchem.cz 

14.05.-16.05.2018 

ATOMEXPO 2018. Sochi, Russia, 

atomexpo.ru 

15.05.-17.05.2018 

11 th International Conference on the Transport, 

Storage, and Disposal of Radioactive Materials. 

London, United Kingdom, Nuclear Institute, 

www.nuclearinst.com 

20.05.-23.05.2018 

5 th Asian and Oceanic IRPA Regional Congress on 

Radiation Protection – AOCRP5. Melbourne, 

Australia, Australian Radiation Protection Society 

(ARPS) and International Radiation Protection 

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

29.05.-30.05.2018 

49 th Annual Meeting on Nuclear Technology 

AMNT 2018 | 49. Jahrestagung Kerntechnik. 

Berlin, Germany, DAtF and KTG, 

www.nucleartech-meeting.com 

03.06.-07.06.2018 

38 th CNS Annual Conference and 42 nd CNS-CNA 

Student Conference. Saskotoon, SK, Canada, 

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

03.06.-06.06.2018 

HND2018 12 th International Conference of the 

Croatian Nuclear Society. Zadar, Croatia, Croatian 

Nuclear Society, www.nuklearno-drustvo.hr 

04.06.-07.06.2018 

10 th Symposium on CBRNE Threats. Rovaniemi, 

Finland, Finnish Nuclear Society, ats-fns.fi 

04.06.-08.06.2018 

5 th European IRPA Congress – Encouraging 

Sustainability in Radiation Protection. 

The Hague, The Netherlands, Dutch Society for 

Radiation Protection (NVS), local organiser, 

irpa2018europe.com 

06.06.-08.06.2018 

2 nd Workshop on Safety of Extended Dry Storage 

of Spent Nuclear Fuel. Garching near Munich, 

German, GRS, www.grs.de 

17.06.-21.06.2018 

ANS Annual Meeting “Future of Nuclear in the 

Shifting Energy Landscape: Safety, Sustainability, 

and Flexibility”. Philadelphia, PA, USA, American 

Nuclear Society (ANS), www.ans.org 

25.06.-26.06.2018 

index2018 – International Nuclear Digital 

Experience. Paris, France, Société Française 

d’Energie Nucléaire, 

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

27.06.-29.06.2018 

EEM – 2018 15 th International Conference 

on the European Energy Market. Lodz, Poland, 

Lodz University of Technology, Institute of Electrical 

Power Engineering, Association of Polish Electrical 

Engineers (SEP), www.eem18.eu 

29.07.-02.08.2018 

International Nuclear Physics Conference 2019. 

Glasgow, United Kingdom, www.iop.org 

05.08.-08.08.2018 

Utility Working Conference and Vendor 

Technology Expo. Amelia Island, FL, USA, 


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 

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. 

EPRI – Electric Power Research Institute, 

San Francisco, CA, USA, www.epri.com 

09.09.-14.09.2018 

Plutonium Futures – The Science 2018. San Diego, 

United States, American Nuclear Society (ANS), 

www.ans.org 

10.09.-13.09.2018 

Nuclear Energy in New Europe – NENE 2018. 

Portoroz, Slovenia, Nuclear Society of Slovenia, 

www.nss.si/nene2018/ 

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 

30.09.-04.10.2018 

TopFuel 2018. Prague, Czwech Republic, 

European Nuclear Society (ENS), American Nuclear 

Society (ANS). Atomic Energy Society of Japan, 

Chinese Nuclear Society and Korean Nuclear Society, 

www.euronuclear.org 

30.09.-05.10.2018 

Pacific Nuclear Basin Conferences – PBNC 2018. 

San Francisco, CA, USA, American Nuclear Society 

(ANS), www.ans.org 

02.10.-04.10.2018 

7 th EU Nuclear Power Plant Simulation ENPPS 

Forum. Birmingham, United Kingdom, Nuclear 

Training & Simulation Group, www.enpps.tech 

14.10.-18.10.2018 

12 th International Topical Meeting on Nuclear 

Reactor Thermal-Hydraulics, Operation and 

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

www.nuthos-12.org 

14.10.-18.10.2018 

NuMat 2018. Seattle, United States, 

www.elsevier.com 

16.10.-17.10.2018 

4 th GIF Symposium at the 8 th 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 

22.10.-26.10.2018 

NUWCEM 2018 Cement-based Materials for 

Nuclear Waste. Avignon, France, French 

Commission for Atomic and Alternative Energies 

and Société Française d’Energie Nucléaire, 

www.sfen-nuwcem2018.org 

24.10.-25.10.2018 

Chemistry in Power Plant. Magdeburg, Germany, 

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

11.11.-15.11.2018 

ANS Winter Meeting. Orlando, FL, USA, 


Calendar


Development of High Temperature 

Gas Cooled Reactor in China 

Wentao Guo and Michael Schorer 

1 Introduction of HTGR Recent developments in High Temperature Gas Cooled Reactor (HTGR) attracted 

widespread attention. China, Japan, South Africa, USA, Russia and France are all actively initiating the development 

work of HTGR. Some developing countries expressed great interest in this type of reactor [1]. 

| | Fig. 1. 

The 10 MWt High Temperature 

Gas-cooled Reactor (HTGR) 

| | Fig. 2. 

The Pebble fuel element 

of the HTGR 

HTGR is one of the six Generation IV reactors put forward 

by Generation IV International Forum (GIF) in 2002. 

This type of reactor has high outlet temperature. It uses 

Helium as coolant and graphite as moderator. The helium 

temperature at the reactor core inlet/outlet is 250/750 °C. 

Pebble fuel and ceramic reactor core are adopted. At the 

center of each poppy seed-size fuel particle is a uranium 

kernel. Layers of carbon and silicon carbide contain the 

radioactive material [2]. Figure 1 shows the overall 

structure of the HTR-10 MW Test Module constructed by 

Institute of Nuclear and New Energy Technology, Tsinghua 

University (INET). Figure 2 shows the pebble fuel element 

structure of HTGR. 

The most important feature of modular high temperature 

gas cooled reactor is that under any accident conditions, 

including large loss of coolant accident (LLOCA), 

the reactor can keep in safe state without any human or 

machine intervention. 

Modular HTGR also has other advantages such as: 

1. High generating efficiency: Its efficiency is 25 % higher 

than pressurized water reactor (PWR) nuclear power 

plants because of the high outlet temperature. 

2. 2. Short construction period: 100 MWe HTGR adopts 

modular construction approach. Construction period 

can be reduced to two years. Compared to PWR power 

plants which have 5 to 6 years of construction, the 

interest payment during construction is reduced and 

the construction investment can be reduced by 20 %. 

3. 3. Simple system: The HTGR has passive safety features 

which greatly simplify the system. Engineering safety 

facilities like emergency core cooling system and full 

grade containment don’t need to be installed, which 

can reduce the construction investment. 

2 The development history of China’s HTR 

and its current situation 

The HTGR research and development work in China started 

in 1970s. By implementing the National High-Technology 

Project (863), Tsinghua University designed and 

built HTR-10 MW Test Module under the support 

of China National Nuclear Corporation (CNNC). It 

realized the first power generation on January 7, 

2003 [3]. 

In 2006, Tsinghua University in Beijing, China 

Nuclear Engineering Group Corporation (CNEC) 

and China Huaneng Group co-financed the 

construction of the HTR demonstration project, 

after which a complete industrial chain is formed. 

In this system, Institute of Nuclear and New Energy 

Technology, Tsinghua University is the liability 

subject of R&D in charge of technology R&D, 

providing design and technical support; CNEC 

is the major special project implementation 

body, responsible for designing, purchasing and 

constructing the demonstration project of 

nuclear island and its auxiliary system; Huaneng Shandong 

Shidao Bay Nuclear Power CO., LTD. takes charge of the 

investment operations of the demonstration project [4]. 

The High Temperature Reactor-Pebble-bed Modules 

(HTR-PM) under construction has two reactors and 

one turbine. On December 9, 2012, the construction of 

Shandong Rongcheng Shidao Bay HTR demonstration 

project started. On April 20, 2015, civil construction of the 

basements came to an end and turned to the intensive 

equipment installation stage. The key point for construction 

was shifted from civil construction to installation 

construction. On June the 24 th , after two months of 

arduous struggle, the Shidao Bay Nuclear Power Project 

completed the pouring task of the reactor building 

walls for the first modular High Temperature Gas-cooled 

Demonstration Reactor in the world [5]. The reactor 

building walls were poured to 41.30 meters, marking 

the HTGR project meeting the requirement of heavy 

equipment lifting. On June the 27 th , capping of the Shidao 

Bay HTGR conventional island is finished [6]. This is 

another major project after the pouring task on June 24 th . 

On March 3, 2016, the construction of the reactor 

pressure vessel (RPV) and metal components inside the 

reactor was finished and they were transported to the site. 

On September 14, 2016, they finished installing the RPV 

for the first and second reactor as well as the internal metal 

components of RPV for the first reactor. The cylindrical 

vessel, 25 meters high and weighing 610 tons, is the 

biggest, heaviest and most complicated pressure vessel for 

a nuclear reactor, according to a statement from Huaneng 

Shandong Shidao Bay Nuclear Power Co. (HSNPC), the 

plant’s builder and operator. On October 14, 2016, the 

demonstration project finished all the tests of inverse 

power transmission successfully. On December 29, 2016, 

the main control room in Shidao Bay nuclear power plant 

is ready to be used. On January 21, 2017, the installation of 

the reactor core vessel was finished. The reactor core vessel 

is the key component of the metal structures inside the 

81 

ENERGY POLICY, ECONOMY AND LAW 


Development of High Temperature Gas Cooled Reactor in China ı Wentao Guo and Michael Schorer


ENERGY POLICY, ECONOMY AND LAW 82 

reactor core. It is used to support the reactor core and 

locate the reactor core components. On June 8, 2017, the 

installation of the ceramic components inside the second 

reactor core was finished, which means half of the 

installation progress of the main facilities in the nuclear 

island has been done. Before August 11, 2017, the fuel 

production line has produced 250,000 pebbles, which met 

the requirement of connecting to the grid for HTR-PM. 

The project is planned to be completed and put into operation 

at the end of 2017/beginning of 2018, but probably 

it will be delayed (Figure 3). The design lifetime of 

HTR-PM is 40 years. 

| | Fig. 3. 

The construction of Shidao Bay HTGR conventional island was finished 

on June 27, 2015 (photo credits: Shidao Bay NPP). 

3 Safety features of HTGR 

One of the most important safety issues for nuclear power 

plant is decay heat removal. In the Three Mile Island and 

Fukushima Daiichi nuclear accidents, the reactor cores are 

overheated and melt down due to the failure of decay heat 

removal. In Chernobyl accident, the failure of decay heat 

removal system caused the resulting sequences after the 

initial exploration due to the fission power increment. 

So developing a highly reliable emergency core cooling 

system with reliable water and electricity supply is very 

important for a light water reactor (LWR). 

But for HTGR, inherent safety can be achieved based 

on three physical ideas: 1. using silicon carbide (SiC), 

which has very good heat-resistance, as the fuel cladding; 

2. lowering the volumetric power density of the reactor 

core significantly; 3. using identical small reactor modules 

to replace a large reactor in order to make sure that the 

reactor core won’t be heated to the temperature limit [7]. 

Besides physical ideas, the safety of HTGR can be 

protected from three engineering designs: 

1. Multiple barriers to prevent the release of 

radioactivity 

The HTGR has three safety barriers to prevent the release 

of radioactivity. The first barrier is the fuel particles coated 

with SiC. The maximum temperature of the fuel particles 

is designed to be limited to 1,600 °C under any operation 

or accident conditions. Less than 1,600 °C, the coat of the 

particles can maintain integrated [8]. The second barrier 

is the pressure boundary of the primary circuit, which 

contains the reactor pressure vessel, the steam generator 

pressure vessel and the hot gas duct pressure vessel which 

connects the previous two vessels. The likelihood for 

these three vessels to have ruptures can be neglected. The 

third barrier is the bounding volume, which contains the 

primary circuit cabin, Helium purification cabin as well as 

fuel loading and unloading cabin. They can prevent the 

radioactive gas to be released into the atmosphere. 

2 Passive decay heat removal system 

The thermal design of HTGR has already considered that 

in case of any accidents, the cooling of the reactor core 

doesn’t need any active decay heat removal system. The 

decay heat in the reactor core can be removed from the 

core to the surface cooler outside of the reactor pressure 

vessel passively through heat conduction and radiation. 

Then the heat can be passed to the atmosphere from the 

surface cooler by nature convection. If the primary circuit 

lost pressure and the main and the auxiliary decay heat 

removal system are out of work, the decay heat can still be 

removed from the core to the outside. The reactor core 

meltdown can be avoided. Under accident conditions, 

because the decay heat cannot be removed by the main 

decay heat removal system, the temperature of the pebbles 

will be increased. In order to make sure the maximum 

temperature of the pebbles will not exceed 1,600 °C, some 

restrictions to the power density and geometry of the 

reactor core are necessary. That’s the reason why the 

capacity of the HTGR is usually small. 

3 Negative temperature coefficient has good reactivity 

compensation 

The reactor has a relatively high negative temperature 

coefficient for the fuel and moderator and if it is under 

normal condition, the margin between the maximum 

temperature of the pebbles and its limit is large. The 

negative temperature coefficient can give a good reactivity 

compensation. When a positive reactivity is introduced 

into the reactor, it can be automatically shut down thanks 

to the reactivity compensation from the negative temperature 

coefficient [9]. 

The long term operation of HTR-10 and different 

safety experiments have proved the inherent safety of 

HTGR, which improved the public acceptance of nuclear 

reactors. 

4 Fuel technology 

In 2005, INET built a prototyping fuel-production facility 

with a capacity of 100,000 fuel elements per year. In order 

to solidify the fabrication level, INET started to construct 

HTGR fuel-production factory in Baotou, Northern China 

in 2013. The fuel-production equipment was installed in 

2014. In 2015, they started the commissioning and trial 

production. Some experiments have been done in Petten, 

the Netherlands. The irradiation test of five fuel spheres of 

the HTR-PM started in October 2012 in the high flux 

reactor (HFR) and finished on December 30, 2014. The 

fuel sphere quality, which is one of the key technologies in 

HTR-PM project, has been proved to meet the requirements 

[7]. 

On August 15, 2016, the construction of the fuel 

production line in Baotou was finished and the fuel pebble 

production started. By July 17, 2017, the fuel production 

line has already produced 200,000 pebbles. It means 

that the fuel production of HTGR has shifted from trial 

production to industrial production. It also means that the 

fuel production technology of HTGR in China is leading 

the world, which has great significance for achieving 

commercialization and export of HTGR [10]. 

When a fuel element is discharged from the bottom 

of the RPV to the fuel handling system, its burn-up is 

measured immediately. If its burn-up does not reach the 

design burn-up limit, it will be recharged into the reactor 




core from the top of the RPV. Otherwise it will be identified 

as a spent fuel and sent to the spent fuel storage system. In 

the spent fuel storage system, spent fuels are put into a 

storage canister. Each storage canister contains 40,000 

spent fuels. After a storage canister is full with spent 

fuels, it is sealed and moved to the ventilated storage well. 

Each storage well contains five vertically placed storage 

canisters. Spent fuels after ten years of storage will be 

moved from the nuclear island to a large intermediate 

storage building on the site and stored there during the 

rest service time of the plant. As for reprocessing, it is 

technically feasible and similar to the technology used in 

PWR. At present, China is still developing this reprocessing 

technology and tends to apply it in the future. 

5 Future expectations of HTGR in China 

The HTGR industrialization has shifted from research 

toward commercial applications. CNEC announced that 

the feasibility study report of the 600 MWe commercial 

high temperature reactor project in Ruijin, Jiangxi province 

has passed the experts auditing and promises to be the 

first commercial Generation IV nuclear power plant in the 

world. At present, China has mastered all the technology of 

HTGR systematically and takes the lead in the world. 

The home manufacture can be realized for 95 % of the 

equipment. 

Next step, CNEC and Jiangxi Province will combine 

together and submit the project proposals to the National 

Development and Reform Commission (NDRC), applying to 

list the project into National Nuclear Long-and-medium 

Term Development Planning. After having the permit, the 

feasibility study of the project will be carried out. Land 

requisition, “Five-outlet-one Dish” 1 

and construction of 

auxiliary facilities will be carried on at the same time. After 

getting the approval from NDRC and obtaining building 

permits from National Nuclear Safety Administration 

( NNSA), the commencement of work for the two units in 

the first-stage project was planned in 2017 and they would 

be combined to the grid around 2021. But due to some 

reasons this project is delayed and hasn’t been started yet. 

6 HTGR cooperation between China and 

other countries 

By the way of multi-module combination, the installed 

capacity of HTGR nuclear power units can be 200 MWe, 

400 MWe, 600 MWe, 800 MWe and 1000 MWe, which can 

be operated with flexibility to suit the market and meet 

the need of different power grid. It is suitable for being 

constructed close to load centers as well as in countries 

and regions with small or middle power grids. 

Many countries in Southeast Asia, Middle East and 

Europe, including some potential users in China, express a 

keen interest in the application of HTGR in nuclear electric 

power generation, sea water desalination, petrochemical 

industry and coal chemical industry. The related business 

cooperation is under way. 

At present, CNEC starts working on HTGR preliminary 

work in Jiangxi, Hunan, Guangdong, Fujian, Shandong, 

Hubei and Zhejiang province successively. Meanwhile, 

CNEC signs the memorandum of understanding (MOU) on 

cooperation with Dubai Nuclear Energy Committee and 

provides King Abdulaziz City for Science and Technology 

(KACST) with the design scheme of HTGR sea water desalination. 

They have also reached a consensus on signing the 

memorandum of understanding on cooperation with Saudi 

Energy City. On April 21, 2015, they signed the MOU 

with South African Nuclear Energy Corporation (NECSA). 

CNEC is jointly with other organization concerned to provide 

nuclear fuels, spent fuel reclamation, nuclear power 

plant operation, technical support, personnel training and 

other integration services to the international market. 

7 Conclusions 

The Generation IV nuclear power system is an advanced 

system which has a major revolution in economy, safety, 

waste treatment and nuclear nonproliferation. HTGR is 

considered to be the most possibly actualized and the most 

promising advanced reactor type in the near future by the 

international nuclear community [9]. 

Under the support of the National High-Technology 

Project, Institute of Nuclear and New Energy Technology, 

Tsinghua University constructed the HTR-10 MW Test 

Module successfully, and achieved joining the national 

power grid with full power. Long-term operation and 

safety tests verified the intrinsic safety of HTGR and 

proved the technical feasibility of HTGR. The success of 

HTR-10 MW Test Module construction and operation 

marks that China has made a breakthrough in the R&D of 

HTGR. China has been included among those advanced 

countries in the development of HTGR technology. The 

construction of the Shidao Bay HTR-PM demonstration 

project is close to an end. Hopefully it will start operation 

in the near future. At that time, it will be the world’s first 

modular HTGR commercial demonstration power plant. 

In early 2006, large pressurized water reactor and 

HTGR were included in the 16 major scientific and 

technological projects by “China’s national policy for 

medium and long-term scientific development” in which 

they are striving to make breakthroughs in 15 years. 

Actualizing the major scientific and technological project 

of HTGR marks that the HTGR technology in which China 

has self-owned intellectual property takes a crucial step 

towards industrialization. 

References 

[1] Zongxin, Wu: The development of high temperature gas-cooled 

reactor in China. Nuclear Power Engineering 21.1 (2000): 39-43. 

[2] http://baike.baidu.com/ 

[3] http://military.china.com/news/568/20150421/19562626.html 

[4] http://digitalpaper.stdaily.com/http_www.kjrb.com/kjrb/ 

html/2014-11/01/content_282325.htm?div=-1 

[5] http://www.cet.com.cn/nypd/hn/1576726.shtml 

[6] http://paper.people.com.cn/zgnyb/html/2015-07/06/ 

content_1585012.htm 

[7] Zhang, Zuoyi, et al.: The Shandong Shidao Bay 200 MW e High- 

Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM) 

Demonstration Power Plant: An Engineering and Technological 

Innovation. Engineering 2.1 (2016): 112-118. 

[8] Tang, Chunhe, et al.: Research and development of fuel element 

for Chinese 10 MW high temperature gas-cooled reactor. Journal 

of Nuclear Science and Technology 37.9 (2000): 802-806. 

[9] Fu Xiaoming, Wangjie, October 2006. Summary of HTGR 

Development in China. Modern Electric Power. 

[10] http://energy.people.com.cn/n1/2017/0718/ 

c71661-29412747.html 

Authors 

Wentao Guo 

Paul Scherrer Institute 

Department of Nuclear Energy and Safety 

5232 Villigen PSI, Switzerland 

Michael Schorer 

Swiss Nuclear Forum 

4600 Olten, Switzerland 

1) Five-outlet-one Dish: 

In order to construct 

rationally and 

orderly, some firstphase 

preparations 

need to be made, 

such as electrifying, 

communication, 

road access, water 

access, gas access 

and land smoothing. 

ENERGY POLICY, ECONOMY AND LAW 83 




Die Haftung nach § 26 AtG – ein Mauerblümchen? 

84 

SPOTLIGHT ON NUCLEAR LAW 

Christian Raetzke 

Die Haftung für Schäden aus Radioaktivität kann sich nach deutschem Recht aus drei Quellen ergeben. In der 

öffentlichen und juristischen Diskussion ist fast immer nur von der Haftung nach dem Pariser Übereinkommen (PÜ) die 

Rede. Das PÜ gilt in Deutschland unmittelbar (siehe auch § 25 AtG – Atomgesetz). Es regelt aber nicht den gesamten 

Bereich der Atomhaftung, sondern – grob gesagt – nur die Haftung im Rahmen der Kernenergie; für diesen Bereich 

mit „besonderem Gefährdungspotential“ wurde ein internationaler Regelungsbedarf gesehen. Das PÜ gilt für 

Kernkraftwerke, im „Front end“ für Anreicherungsanlagen und Brennelementfabriken und im „Back end“ für Aktivitäten 

rund um die Abfälle aus Kernkraftwerken, jeweils einschließlich der entsprechenden Beförderungsvorgänge. 

Als zweite Rechtsgrundlage regelt § 25a AtG die Haftung 

für Reaktorschiffe. Mit der Ausmusterung der Otto Hahn 

ist diese Norm aber vor langer Zeit in der Versenkung 

verschwunden. 

Und dann gibt es schließlich den § 26 AtG. Juristisch ist 

die Norm als sog. Auffangtatbestand gestaltet. Sie erfasst 

alle Schäden „durch die Wirkung eines Kernspaltungsvorgangs 

oder der Strahlen eines radioaktiven Stoffes oder 

durch die von einer Anlage zur Erzeugung ionisierender 

Strahlen ausgehende Wirkung ionisierender Strahlen“, 

die nicht in den Anwendungsbereich des PÜ oder des 

§ 25a AtG fallen. Aus dieser Negativdefinition und 

gleichsam Subtraktion ergibt sich, dass § 26 vor allem auf 

Anlagen und Tätigkeiten außerhalb der Kernenergie (und 

außer Reaktorschiffen) Anwendung findet, also hauptsächlich 

auf den Umgang mit Radioaktivität im Bereich 

der Medizin, Industrie (z. B. Prüfstrahler) und Forschung. 

Unter die Haftung nach § 26 fallen aber auch solche 

Bereiche der Kernindustrie, die aufgrund ihres geringen 

Schadenspotentials vom PÜ ausgeschlossen werden, 

insbesondere Aktivitäten rund um Natururan und abgereichertes 

Uran. Schließlich ordnet § 26 Abs. 2 AtG eine 

entsprechende Geltung für die Kernfusion an. 

Die Haftung nach § 26 AtG trifft den Besitzer 

radio aktiver Stoffe oder von Anlagen zur Erzeugung 

ionisierender Strahlen, weswegen man hier von Besitzerhaftung 

spricht (manchmal wird auch der Begriff 

Isotopenhaftung verwendet, was aber ungenau ist, da es 

eben nicht nur um radioaktive Stoffe geht). Im Falle der 

Beförderung radioaktiver Stoffe haftet nach Abs. 6 der 

Absender. 

Dass § 26 AtG für Aktivitäten gilt, die das PÜ gleichsam 

„übrig lässt“ und die mit einem geringeren Gefahrenpotential 

assoziiert werden, schmälert keinesfalls die 

Bedeutung der Norm. Denn zum einen dürften diese Fälle 

des Umgangs mit Radioaktivität zahlenmäßig diejenigen, 

die sich aus der Nutzung der Kernenergie ergeben, weit 

übersteigen; man denke nur an die vielen Transporte von 

Strahlenquellen für Medizin und Industrie, die jeden Tag 

stattfinden. Zum anderen können sich auch aus diesen 

Anlagen und Tätigkeiten im ungünstigsten Fall zwar kaum 

nationale Katastrophen, aber doch erhebliche Schäden 

bis hin zum Tod von Personen oder zu komplizierten 

Kontaminationen ergeben. 

In der Frage, ob die Haftung nach § 26 AtG eine 

Verschuldenshaftung wie die allgemeine Haftung des 

Bürgerlichen Gesetzbuches (setzt Vorsatz oder Fahrlässigkeit 

voraus) oder eine verschuldensunabhängige 

Gefährdungshaftung (wie im PÜ) sein sollte, hat der 

Gesetzgeber eine mittlere Lösung gewählt, die sog. modifizierte 

Gefährdungshaftung. Im Grundsatz ist es eine 

Gefährdungshaftung: der Geschädigte muss im Prozess 

nicht behaupten und beweisen, dass den Besitzer/ Absender 

ein Verschulden trifft. Vielmehr ist es am Besitzer/ 

Absender, einen Entlastungsbeweis zu führen, wenn er 

kann; immerhin hat er – im Gegensatz zur reinen Ge fährdungs 

haftung – diese Option. § 26 Abs. 1 Satz 2 AtG gibt 

hierfür allerdings qualifizierte (erschwerte) Bedingungen 

vor; fehlendes Verschulden reicht nicht, es müssen weitere 

Umstände wie etwa die nachweisbare „Anwendung jeder 

nach den Umständen gebotenen Sorgfalt“ hinzukommen. 

Das ist eine hohe Hürde. 

Ein zweiter interessanter Aspekt betrifft die Frage 

einer möglichen Kanalisierung. Im PÜ ist die Haftung 

bekanntlich ausschließlich auf den Inhaber (Betreiber) 

einer Kernanlage konzentriert. Zulieferer, Dienstleister 

etc. sind freigestellt; Anspruchsgrundlagen außerhalb des 

PÜ werden ausgeschlossen. Für den Bereich des § 26 AtG 

hat der Gesetzgeber diese Lösung nicht übernommen. 

Dem Geschädigten stehen also neben § 26 AtG auch alle 

anderen Anspruchsgrundlagen des Haftungsrechts zur 

Verfügung und er kann, wenn die Voraussetzungen 

vorliegen, auch andere Beteiligte als den Besitzer/ 

Absender in Anspruch nehmen. Als „Ausgleich“ für diese 

anderen Beteiligten ist in § 4 der Atomrechtlichen 

Deckungsvorsorge-Verordnung (AtDeckV) geregelt, dass 

der Besitzer/Absender sie in bestimmtem Umfang in seine 

eigene Haftpflichtversicherung einbeziehen muss (sog. 

wirtschaftliche Kanalisierung). 

Damit ist auch schon ein dritter Aspekt angesprochen: 

für Tätigkeiten im Bereich des § 26 AtG, die einer 

Genehmigung bedürfen, muss im Genehmigungs verfahren 

eine Deckungsvorsorge (§ 13 AtG) nachgewiesen werden, 

also in der Regel eine Haftpflichtversicherung. Der Betrag 

wird auf der Grundlage der AtDeckV im Genehmigungsverfahren 

festgesetzt. Die Haftung selber ist unbegrenzt; 

übersteigt ein Schaden also den Betrag der Deckungsvorsorge, 

muss der Haftende sein Vermögen einsetzen. 

§ 26 trifft schließlich einige Sonderregelungen für die 

Anwendung von radioaktiven Stoffen oder ionisierender 

Strahlen am Menschen in der medizinischen Forschung 

(da wird die Haftung verschärft) oder bei der Ausübung 

der Heilkunde (dort gilt unter bestimmten Voraussetzungen 

statt § 26 die normale Arzthaftung). 

Soweit ersichtlich, gab es bisher keine Schadensfälle 

im Bereich des § 26, die Anlass zu einschlägiger Rechtsprechung 

geboten hätten; das soll auch möglichst 

so bleiben. Angesichts des Kernenergieausstiegs, der 

juristischen Aufwertung des Strahlenschutzes durch 

das neue Strahlenschutzgesetz und der zunehmenden 

Bedeutung der Fusionsforschung wird § 26 AtG aber 

möglicherweise dennoch etwas aus dem Schatten 

des PÜ heraustreten und vielleicht sein unverdientes 

„ Mauerblümchendasein“ abstreifen. 

Author 

Rechtsanwalt Dr. Christian Raetzke 

CONLAR Consulting on Nuclear Law and Regulation 

Beethovenstr. 19 

04107 Leipzig, Germany 

Spotlight on Nuclear Law 

The Liability According to § 26 of the German Atomic Energy Act – A Wallflower? ı Christian Raetzke


Investigation of Conditions Inside the 

Reactor Building Annulus of a PWR 

Plant of KONVOI Type in Case of Severe 

Accidents with Increased Containment 

Leakages 

Ivan Bakalov and Martin Sonnenkalb 

1 Introduction and analysis method The severe accident at Fukushima Daiichi NPP resulted in 

severe core damage and significant releases of hydrogen and radioactive materials from primary containment boundary 

into or through the reactor buildings of three out of the six reactors (units 1 to 3). Based on analyses of the accident 

progression it was realized that accidentally increased leaks from the inertized containment contributed to the 

radionuclide and hydrogen release into the reactor building, thus leading to hydrogen explosions, severely damaging 

the reactor building constructions. 

The Fukushima Daiichi accident triggered 

worldwide stress tests and 

re-assessments of the NPP plant 

safety. In Germany the process 

resulted in an improvement and 

extension of the existing severe accident 

management (SAM) concept 

by both additional preventive and 

mitigative measures. The main improvements 

in the mitigative domain 

is a new concept of severe accident 

management guidelines (SAMG) with 

strategies and procedures intended to 

be used by the plant crisis team for 

mitigation of the consequences of 

severe accidents. The SAMG concept 

follows relevant recommendations 

of the German Reactor Safety Commission 

RSK [1]. 

Analyses of the hydrogen as well 

as aerosol and noble gas behaviour 

in case of increased containment 

leakages into the reactor building 

annulus of a German PWR KONVOI 

reference plant under severe accident 

conditions have been performed using 

the GRS lumped parameter code 

COCOSYS. The investigation carriedout 

focusses on the assessment of the 

efficiency of newly developed SAM 

measures as described in the new 

SAMG handbook or some measures 

proposed in addition for a PWR 

reference plant of KONVOI type. The 

assessed strategies are related to the 

mitigation of challenging conditions 

inside the reactor building (RB) 

annulus due to design based and 

increased containment leakages 

during severe accidents. 

The analyses are based on previous 

GRS investigations of the hydrogen 

mitigation concept with passive autocatalytic 

recombiners (PAR) inside 

the PWR KONVOI containment [2] as 

well as the reassessment of the effectiveness 

of the filtered containment 

venting concept of PWR KONVOI [3]. 

The main findings contribute to 

further improvement of the planned 

mitigative SAM measures in case of 

enhanced containment leakages into 

the reactor building annulus under 

severe accident conditions. 

1.1 COCOSYS plant model 

The COCOSYS nodalisation scheme 

of the PWR KONVOI plant with focus 

on the RB annulus is presented in 

Figure 1. The nodalisation of the 

containment and the RB annulus is 

developed in such a way that thermal 

and gas stratification processes 

expected under accident conditions, 

local and global convection flows 

between the compartments, and longterm 

convection processes inside 

the containment could be simulated 

appropriately. Therefore, a refined 

subdivision of the containment compartments 

and RB annulus rooms and 

free space was chosen. The model 

considers all relevant gaseous and 

liquid flows through different compartment 

connections such as free 

openings, fire protection doors, burst 

membranes, drainages, etc. For the 

purpose of heat and mass transfer 

modelling inside the containment 

and the RB annulus heat structures 

representing the walls, floors, ceilings 

and metal internals are introduced 

into the model. With all these features 

the model adequately represents all 

relevant design specific features of the 

PWR KONVOI reference plant – both 

inside the containment as well as the 

RB annulus. 

The containment has a total free 

volume of 70,000 m 3 . It is subdivided 

into four areas which can have 

different convection flow regimes 

depending on the initial event of a 

sequence and the break/discharge 

location. The first area represents the 

containment compartments, in which 

the reactor pressure vessel and the 

steam generators are located. The 

| | Fig. 1. 

COCOSYS nodalisation scheme of the RB annulus and location of containment penetrations through the containment steel shell. 

85 

ENVIRONMENT AND SAFETY 


Investigation of Conditions Inside the Reactor Building Annulus of a PWR Plant of KONVOI Type in Case of Severe Accidents with Increased Containment Leakages ı Ivan Bakalov and Martin Sonnenkalb


ENVIRONMENT AND SAFETY 86 

second and third area comprises the 

operating containment compartments 

and the containment dome. The 

fourth area includes all the compartments 

outside the missile protection 

cylinder, the periphery of the containment. 

The volume of the RB annulus is 

subdivided into four areas with a total 

volume of 50,000 m 3 . The first area is 

the annular gap, located above elevation 

21.5 m, which has a total volume 

of 14,900 m 3 . This area, in turn, is 

divided into six axial levels along the 

height of the gap. It is connected to 

the lower part of the annular gap 

( second area) below elevation 21.5 m 

and has a free volume of 4,300 m 3 . In 

this area vertical fire protection walls 

with metal sheets are located, which 

do not allow atmospheric flow in 

azimuthal direction. The third area 

comprises several separate annulus 

rooms located on building floors at 

elevation 6 m to 21.5 m. The annulus 

rooms at elevation 6 m to 9 m are 

separated from the annular gap by 

ventilation systems. The connections 

between these separate rooms are 

provided with fire protection doors 

and fire protection flaps, which automatically 

close, if the room temperature 

exceed ~70° C. The fourth area 

represents all annulus rooms below 

elevation 6 m with a total volume of 

23,100 m 3 . Those rooms have only a 

negligible atmosphere exchange with 

the annular gap above. 

Moreover, the model consists of 

all relevant plant systems used during 

accidents (e.g. the RB annulus exhaust 

air system) or operational systems 

foreseen as SAM measures in the 

SAMG handbook (e.g. the annulus 

air supply/suction system and the 

annulus air recirculation systems). 

The filtered containment venting 

system and the hydrogen recombination 

system with about 65 PARs 

installed inside the containment are 

introduced in the input deck as well, 

using the modelling capabilities of the 

engineered safety features, integrated 

in the COCOSYS code. 

The COCOSYS model also includes 

the containment design leakage of 

0.25 vol.-%/d into the RB annulus. 

For the base case analyses the design 

leakage is assumed to be at the most 

unfavorable place in the area of the 

cable penetrations at elevation 12 m 

(Figure 1 right side), e.g. the leakage 

is located opposite to the single 

suction point of the RB annulus 

exhaust air system, operated in case 

of an accident. In addition, leakages 

are defined from the environment 

through the auxiliary building main 

gate into the lower annulus rooms 

(leakages represented by red arrows 

in Figure 1). 

1.2 Selected representative 

Severe Accident Scenarios 

Two representative and different 

severe accident scenarios – the base 

cases – have been selected for the 

analyses. Some characteristics of the 

scenarios are summarized here, the 

timing of main events is provided in 

Table 1: 

• MBL – a medium break LOCA with 

a failure of the emergency core 

cooling system after the emergency 

water supply tank inventory is 

empty; core degradation starts 

delayed; sequence results in a 

maximum water inventory in the 

containment sump and a late 

filtered containment venting. 

• ND* – a transient with a failure of 

steam generator feedwater supply; 

failure of injection of active 

emergency core cooling systems; 

primary circuit depressurization 

procedure to avoid reactor pressure 

vessel failure at high-pressure; 

core degradation starts early; 

sequence results in a minimum 

water inventory in the containment 

sump and an earlier containment 

venting. 

The two representative base cases 

were already used in earlier analyses 

[2], [3] with respect to the reassessment 

of other mitigative SAM measures. 

In both cases, no melt relocation 

from the reactor cavity into the containment 

sump after melt penetration 

of the biological shield was assumed, 

just water ingress into the cavity and 

therefore extended steam production. 

As melt relocation into the sump with 

cooling of the relocated melt amount 

seems to be a realistic scenario leading 

to reduced production of combustible 

gases, two additional variant calculations 

were done with melt relocation 

into the containment sump. Furthermore, 

a series of COCOSYS variant 

calculations were carried out in order 

to investigate the influence of the 

following specific aspects: 

• Operation/failure of the RB annulus 

exhaust air system installed for 

accident conditions. 

• Variation of the size of containment 

leakages into the reactor 

building annulus: design leakage 

(base case) and a 10 times larger 

leakage. 

• Variation of the containment 

leakage location in the area of 

containment cable penetrations. 

Moreover, the efficiency of different 

SAM measures for mitigation of the 

consequences in the RB annulus, 

documented in the SAMG handbook 

of the reference plant, was analysed. 

These measures are as follows: 

• Use of RB annulus air supply/ 

suction system – provision of a 

controlled ventilation to reduce 

the hydrogen concentration in the 

annulus. 

• Use of RB annulus air recirculation 

system – mixing of the annulus 

atmosphere and elimination of gas 

stratification. 

• Use of emergency air filtration 

system – extraction of air from the 

RB annulus through a filtration 

system to reduce the release of 

radionuclides into the environment. 

The following SAM measure was 

additionally investigated as a possible 

alternative method for hydrogen 

reduction in the annulus. It is related 

to a optional recommendation of the 

RSK [1]. 

• Implementation of a small number 

of PARs in the RB annulus upper 

part to prevent combustible gas 

mixtures. 

2 Results – Quantification 

of the effectiveness of 

selected AM measures 

Selected results are presented in the 

following only for one base case 

scenario (MBL) with the operation of 

RB annulus exhaust air system used in 

case of accidents and for some variant 

Scenario 

Start of steam/water 

leak flow into 

containment 

Start of 

core melting 

RPV failure and 

melt release 

into cavity 

Water ingression into 

cavity and possible 

melt release into sump 

Start of filtered 

containmentventing 

ND* 1.4 hr 3.5 hr 6.5 hr 17.1 hr 66.5 hr 

MBL 0.0 hr 5.8 hr 8.9 hr 13.5 hr 82.2 hr 

| | Tab. 1. 

Timing of characteristic events of severe accident progression of base case scenarios. 




calculations. Moreover, the results of 

the two severe accident scenarios 

(MBL and ND*) for the base cases 

with increased containment leakages 

are compared regarding their effect 

on the accident consequences. 

2.1 Base case with containment 

design leakage 

The hydrogen concentration in the RB 

annulus is presented in Figure 2 (left 

side). In the base case no formation of 

combustible gas mixtures (> 4 vol.-% 

hydrogen) in the RB annulus is 

observed during the calculated time 

period, and some fire protection 

doors and flaps between the separated 

rooms of the annulus close automatically 

when the atmosphere 

temperature reaches 70 °C limiting 

the hydrogen and radionuclide inflow 

into these areas (Figure 2 right side). 

Due to the operation of the RB annulus 

exhaust air system, the hydrogen 

concentration remains below 1 vol.-% 

and decreases further in the long term 

when the containment filtered venting 

starts reducing the hydrogen leakage 

from the containment. Gas stratification 

with slightly different gas 

concentrations at different elevations 

is formed in the annulus gap due to 

the operation of the annulus exhaust 

air system. 

| | Fig. 2. 

H 2 concentration in the RB annulus for base case scenario (MBL) with operation of RB annulus exhaust air system; 

RB annular gap (left) and RB annulus rooms (right). 

| | Fig. 3. 

H 2 concentration in the RB annulus for base case (left) and variant case (right) with a 10 times larger containment leakage, 

both cases with operation of RB annulus exhaust air system. 


2.2 Variant calculation with a 

10 times larger containment 

leakage 

As already mentioned, one of the 

goals is to investigate the conditions in 

the RB annulus in case of increased 

containment leakages. For this purpose, 

a COCOSYS variant calculation 

was performed assuming a 10 times 

larger containment leakage. The RB 

annulus exhaust air system was 

assumed to be in operation as in the 

base case. It sucks steam-air mixture 

from one selected location of the RB 

annulus at about 12 m level. Figure 3 

compares the hydrogen concentration 

and Figure 4 the aerosol concentration 

in the base case and the variant 

calculation. The overall behaviour in 

the RB annulus is the same, but the 

variant with 10 times larger containment 

leakage leads to the formation of 

combustible gas mixtures (> 4 vol.-% 

hydrogen) in the upper annulus area 

and a higher aerosol concentration 

especially in the early accident phase 

with large releases from the reactor 

circuit during core melting. The 

results show that the RB annulus 

exhaust air system is not efficient 

enough to keep the H 2 concentration 

below the lower combustible limit of 

| | Fig. 4. 

Aerosol concentration in the RB annulus for base case (left) and variant case (right) with a 10 times larger containment leakage, 

both cases with operation of RB annulus exhaust air system. 

| | Fig. 5. 

Comparison of pressure in the containment (left) and MCCI gas generation (right) for the cases with and without melt relocation. 

4 vol.-% H2 in all RB annulus areas. 

The following three gas concentration 

zones are established (Figure 3 right): 

• RB annulus above 16 m with 

hydrogen concentrations up to 

~ 5 vol.-%. 

• RB annulus at ~12 m (leak location) 

with low hydrogen concentrations 

up to ~ 2 vol.-%. 

• RB annulus at ~ 6 m and below 

with very low hydrogen concentrations 

< 0.1 vol.-%. 





2.3 Variant calculation with a 


leakage and consideration 

of a potential melt relocation 

into containment sump 

As already noted, all previous analyses 

conducted by GRS have been performed 

assuming no melt relocation 

from the reactor cavity into the containment 

sump after melt penetration 

of the biological shield. Since the melt 

is very likely to melt-through the biological 

shield, a variant calculation 

with a 10 times larger containment 

leakage and a failure of the RB annulus 

exhaust air system was performed 

assuming melt relocation into the 

containment sump. 

After penetration of the biological 

shield, the corium spreads into the 

containment sump and comes into 

contact with the sump water. This 

results in a higher steam generation, 

which in turn leads to a faster longterm 

containment pressurization 

compared to the case without melt 

relocation (Figure 5). Because of the 

higher steam production, the filtered 

containment venting starts significantly 

earlier than in the case without 

melt relocation. 

Shortly after the melt relocation 

into the sump, the corium solidifies 

within a very short time period and 

the generation of combustible gases 

(H 2 and CO) is terminated. Due to the 

overall lower gas production, the H 2 

concentrations in the containment, 

and thus also in the RB annulus, are 

significantly lower compared to those 

| | Fig. 6. 

Comparison of H 2 concentration in the containment (left) and H 2 concentration in the RB annulus (right) for the cases with and 

without melt relocation. 

| | Fig. 7. 

Comparison of containment pressure (left) and H2 mass generated during MCCI (right) for the MBL and ND* base cases. 

in the calculations without melt 

relocation (Figure 6) and the lower 

combustible limit is no longer reached 

in the RB annulus. 

2.4 Effect of the selected severe 

accident scenarios on the 

accident consequences 

In order to investigate the effect on 

the accident consequences, the results 

of the two analyzed severe accident 

scenarios (MBL and ND*) have been 

compared for the base cases with 

increased containment leakages. 

A comparison of the containment 

pressure response calculated for 

the two base cases with increased 

containment leakages is shown in 

Figure 7 (left). The comparison 

demonstrate that in the ND* base 

case, the filtered containment venting 

starts about 16 hours earlier than in 

the MBL base case. Figure 7 (right) 

depicts a comparison of the hydrogen 

mass generated during the MCCI for 

the two accident scenarios. Because of 

the earlier venting in the ND* base 

case, less hydrogen is generated until 

the start of containment depressurization. 

This is due to the fact that in the 

ND* base case the MCCI duration is 

shorter than that in the MBL base 

case. Hence, for the ND* case, a total 

amount of hydrogen of about 3,700 kg 

is generated, while for the MBL case, 

the total hydrogen mass, generated 

until the start time of filtered containment 

venting, is about 4,000 kg. The 

hydrogen concentrations in the RB 

annulus calculated for the two base 

case scenarios are compared in 

Figure 8. Due to the earlier start of 

containment venting in the ND* base 

case the maximum hydrogen concentration 

in the RB annulus is lower than 

that in the MBL base case. From the 

comparison it is evident that the 

hydrogen lower combustible limit of 

4 vol.% is not exceeded until the 

beginning of the containment depressurization. 

| | Fig. 8. 

Comparison of H 2 concentration in the RB annulus ring (left) and H 2 concentration in the RB annulus rooms (right) for the MBL and 

ND* base cases. 

2.5 Variant calculations with a 


leakage and AM measures 

As part of the assessment of potential 

mitigative AM measures the efficiency 

of the RB annulus air supply/suction 

system to reduce the hydrogen concentration 

in the RB annulus was 

investigated. For this purpose, a 

variant calculation with a 10 times 

larger design leakage and a failure of 

the RB annulus exhaust air system 

was carried out (Figure 9 left) and 

another one assuming that the RB air 

supply/exhaust systems are put into 




operation as AM measure at approx. 

50 h after the accident onset (Figure 9 

right). The results show a significantly 

increased hydrogen concentration in 

the RB annulus in case of a failure of 

the RB annulus exhaust air system 

(Figure 9 left). 

Further, in this case the use of 

the RB annulus air supply/exhaust 

systems is efficient to reduce the 

hydrogen concentration and prevent 

the formation of combustible gas 

mixtures in the annulus rooms. With 

the operation of the system the hydrogen 

is removed from the annulus 

quickly and the hydrogen concentration 

remains below 1 vol.-% for the 

long term. In that case, the use of the 

emergency air filtration system of the 

plant is needed in addition to limit 

the radionuclide releases into the 

environment. 

In addition, a possible alternative 

method for hydrogen reduction in the 

annulus was investigated assuming 

the installation of a small number of 

medium size PARs in the upper RB 

annulus (Figure 10 right). The results 

are compared with a variant calculation 

with a 10 times larger design 

leakage and failure of the RB annulus 

exhaust air system (Figure 10 left). 

The results show that already the 

implementation of PARs of medium 

size can significantly reduce the 

hydrogen concentration in the RB 

annulus and keep it well below 

lower combustible limits. The hydrogen 

depletion starts at approx. 40 h 

(150,000 s) after the accident onset 

if the concentration exceeds about 

1 to 2 vol.-%. Thus, an AM concept 

with the installation of some PARs in 

the annulus is considered a very 

efficient mitigation measure for preventing 

formation of combustible gas 

mixtures in the RB annulus not just in 

the case presented. 

3 Conclusions 

The behaviour of hydrogen as well as 

aerosol and noble gases released into 

the reactor building annulus of a 

German PWR KONVOI reference plant 

resulting from increased containment 

leakages under severe accident conditions 

was investigated using the 

GRS code COCOSYS. Two representative 

and different severe accident 

scenarios – the base cases – have been 

selected for the analyses. 

The calculation results show no 

formation of combustible gas mixtures 

in the RB annulus during the observation 

period for the base case with 

containment design leakage and 

operation of RB annulus exhaust 

| | Fig. 9. 

H 2 concentration in the RB annulus for variant cases with 10 times larger leakages and failure of RB annulus exhaust air system (left) 

and with AM measure “operation of RB annulus air supply/exhaust systems” (right). 

| | Fig. 10. 

H 2 concentration in the RB annulus for variant cases with 10 times larger leakages and failure of RB annulus exhaust air system (left) 

and with AM measure “PARs in the RB annulus” (right). 

air system. It was identified that in 

this case separate annulus rooms are 

isolated at an early stage by the automatic 

closing of fire protection doors, 

thus preventing a further increase in 

the hydrogen concentration in these 

rooms. 

In contrast, the variant calculation 

with a 10 times larger containment 

design leakage leads to formation of 

combustible mixtures in the upper RB 

annulus area. In this case, the RB 

annulus exhaust air system is not 

efficient enough to prevent formation 

of combustible gas mixtures in the 

upper RB annulus area. 

Further, the variant calculation 

assuming melt relocation into the 

containment sump demonstrated that 

the corium spreading into the sump 

results in a higher steam generation, 

which leads to a faster long-term 

containment pressurization. After the 

melt relocation into the sump, the 

corium solidifies within a short time 

and the generation of combustible 

gases (H 2 and CO) coming from 

MCCI is terminated. As a result, the 

H 2 concentrations in the containment 

as well as in the RB annulus are 

significantly lower compared to those 

in the case without melt relocation. In 

this case, the lower combustible limit 

of 4 vol.% in the RB annulus is no 

longer reached. 

Moreover, the results of the two 

analyzed severe accident scenarios 

(MBL and ND*) were compared in 

order to investigate their effect on 

the accident consequences. From the 

comparison it was identified that in 

the ND* base case, the filtered 

containment venting starts about 

16 hours earlier than in the MBL base 

case. As a result, the maximum hydrogen 

concentration in the RB annulus, 

calculated for the ND* base case, is 

lower than that in the MBL base case. 

The comparison showed that in the 

ND* base case the hydrogen concentration 

does not exceed the lower 

combustible limit of 4 vol.% until the 

beginning of the containment depressurization. 

Within the scope of the project, the 

efficiency of different AM measures 

for mitigation of accident consequences 

in the reactor building annulus 

was analyzed. The assessment 

results show that the operation of RB 

annulus air supply/suction system 

significantly reduces the hydrogen 

concentration and prevents formation 

of combustible gas mixtures in RB 

annulus. Therefore, the use of these 

ventilation systems is considered as a 

very promising accident management 

measure for reducing the hydrogen 

concentration in the reactor building 

annulus. However, in that case the 






emergency air filtration system of 

the plant is needed in addition to limit 

the radionuclide releases into the 

environment. 

With respect to mitigation of the 

hydrogen risk in the annulus it is 

demonstrated that the implementation 

of a small number of PARs would 

be a very efficient and fully passive 

mitigation measure without additional 

aerosol release into the environment. 

Acknowledgments 

The authors like to acknowledge the 

German Federal Ministry for the 

Environment, Nature Conservation, 

Building and Nuclear Safety for the 

financial support of the project 

3615R01345. 

References 

[1] Recommendation of German Reactor 

Safety Commission (RSK): Hydrogen 

Release from Containment. Annex of 

the Proceedings of 475 th meeting of 

RSK, 15.04.2015. 

[2] Band, S., Schwarz, S., Sonnenkalb, M.: 

Nachweis der Wirksamkeit von 

H 2 -Rekombinatoren auf der Basis 

ergänzender analytischer Untersuchungen 

mit COCOSYS für die 

Referenzanlage GKN-2. Final Report 

of BMUB project 3609R01375, 

GRS-A-3652, March 2012. 

[3] Schwarz, S., Sonnenkalb, M.: Analyse 

der Belastung von Gleitdruckventuriwäschern 

in SHB-Ventingsystemen 

von DWR KONVOI und 

SWR-72 bei Unfällen. Final Report 

of BMUB project 3613R01320, 

GRS-A-3820, August 2015. 

Authors 

Ivan Bakalov 

Research Fellow 

Gesellschaft für Anlagen- und 

Reaktorsicherheit (GRS) gGmbH, 

Kurfürstendamm 200 

10719 Berlin, Germany 

Dr. Martin Sonnenkalb 

Department Head 

Gesellschaft für Anlagen- und 

Reaktorsicherheit (GRS) gGmbH, 

Schwertnergasse 1 

50667 Cologne, Germany 

Sensitivity Analysis of MIDAS Tests Using 

SPACE Code: Effect of Nodalization 

Shin Eom, Seung-Jong Oh and Aya Diab 

1 Introduction The SPACE thermal hydraulic analysis computer code has been developed by KHNP (Korea 

Hydro and Nuclear Power) [1]. The SPACE code is based on the three-field governing equations (vapor, continuous 

liquid, and droplet). It improves the accuracy by solving the mass, energy, and momentum conservation equations for 

each phase and adopts the proven numerical methods as well as the models for various thermal hydraulic phenomena. 

With the new code, the best estimate 

LOCA (Loss Of Coolant Accident) 

methodology needs to be reestablished. 

For APR1400 LBLOCA (Large 

Break LOCA, APR1000: Advanced 

Power Reactor 1000 MWe), KREM [2] 

has been developed one of the best 

estimate methodology using RELAP5 

code [3, 4]. With the new code, one 

needs to look at the code performance 

to develop best estimate + uncertainty 

method. In this paper, as a part of the 

development effort, we focus on the 

impact of nodalization on the code 

predictions, more specifically, on the 

ECC bypass phenomenon. 

For APR1400 LBLOCA, ECC bypass 

phenomenon is one of the important 

phenomena which would occur in the 

downcomer during the reflood phase 

of LOCA [5]. To study the ECC bypass 

phenomenon, KAERI carried out the 

ECC bypass tests using the MIDAS 

facility [6, 7, 8]. MIDAS simulation is a 

part of the assessment of the KREM. 

One of the important parameters 

for the MIDAS test is ECC bypass fraction. 

The results for each nodalization 

were compared with MIDAS test data. 

The main aim of this study is therefore 

to examine the sensitivity of the 

SPACE code to the number of thermal 

hydraulic channels in the downcomer 

region. 

| | Fig. 1. 

Isometric View of the MIDAS Facility [7]. 

| | Fig. 2. 

Top View of the MIDAS Facility Downcomer [7]. 

2 MIDAS test 

The MIDAS test facility is a steamwater 

separate effect test facility 

which is scaled down from APR1400 

[9]. It is focused on the investigation 

of the ECC bypass phenomenon in the 

downcomer annulus. The test condition 

was determined, based on the 

analysis of the TRAC (Transient 

Reactor Analysis Code) [10]. The 

isometric and top view of the MIDAS 

facility is depicted in Figure 1 and 

Figure 2. 

To investigate the effect of the DVI 

injection nozzle location on the ECC 

bypass fraction, fifteen separate effect 


Sensitivity Analysis of MIDAS Tests Using SPACE Code: Effect of Nodalization ı Shin Eom, Seung-Jong Oh and Aya Diab


tests have been performed with only 

DVI-2 (farthest from the broken cold 

leg), only DVI-4 (closest to the broken 

cold leg), and DVI-2&4 with both 

injection nozzles activated. Table 1 

provides the experimental conditions 

for the 15 tests. 

The bypass fractions of the MIDAS 

experiment for the test conditions are 

presented in the Figure 3. The test 

results show that the ECC bypass 

fraction is highly dependent on the 

injection nozzle location with respect 

to the broken leg as well as the injected 

steam flow rate. 

Injecting through the nozzle closet 

to the broken leg (DVI-4 tests) 

show that the direct bypass fraction 

increases drastically for a steam flow 

rate above 0.7 kg/s. This is expected 

since at a higher steam flow rate, the 

relative speed between the two fluid 

streams becomes higher resulting in a 

higher shear effect. 

On the other hand, injecting 

through the nozzle farthest to the 

broken leg (DVI-2 test) dramatically 

decreases the bypass fraction, and 

accordingly most of the injected ECC 

water penetrates into the lower downcomer. 

This is primarily due to the 

lower interfacial interaction between 

the two streams. As a result of the 

spatial separation, the ECCS stream 

becomes more inertially driven. 

With both nozzles activated 

( DVI-2&4 tests), the bypass ratio 

increases with steam flow rate but 

at a much slower rate as compared 

to that of DVI-4 tests. This may be 

attributed to lower interfacial-interaction 

between the injected steam and 

ECCS stream for the combined case. 

Test 

No. 

Steam 

in (kg/s) 

ECCS Injection 

Nozzle 

KM100 1.7924 DVI-2&4 

KM101 1.6149 DVI-2&4 

KM102 1.3753 DVI-2&4 

KM103 1.1711 DVI-2&4 

KM104 0.0493 DVI-2&4 

KM105 0.9378 DVI-2&4 

KM106 0.8592 DVI-2&4 

KM107 0.8096 DVI-2&4 

KM108 0.7540 DVI-2&4 

KM109 1.8086 DVI-2 

KM110 1.0555 DVI-4 

KM111 0.8995 DVI-4 

KM112 0.7991 DVI-4 

KM113 0.7360 DVI-4 

3 MIDAS Modeling 

for the SPACE Code 

A SPACE model of the MIDAS facility 

is developed with three different 

nodalization schemes as shown in 

Figure 4 to Figure 6. The downcomer 

is modeled as an annulus component 

with 4, 6, and 12 circumferential 

channels. A nodalization sensitivity 

analysis for the ECC bypass phenomenon 

was performed using the SPACE 

code version 3.0. 

For the KREM which has best 

estimate LOCA methodology using 

RELAP5 code, the downcomer was 

represented with 6 channels [4]. The 

comparison with MIDAS test results as 

a part of the code validation showed 

that RELAP5 code over-predicts the 

bypass fraction for low steam flow 

cases while predicts reasonably for 

higher steam flow cases. 

The intact cold legs (CL-1, CL-2, 

and CL-3) are connected to the 

annulus component using a normal 

junction with branch components. A 

time-dependent volume and a 

time-dependent junction were used to 

admit the steam flow rate through 

each cold leg. The broken cold leg 

(CL-4) is connected to the annulus 

component using a normal junction 

with a branch component. 

The DVI nozzle (DVI-4) closest to 

the broken leg is connected to the 

same hydraulic channel as the break 

(CL-4) whereas the DVI nozzle 

(DVI-2) farthest from the break shares 

the same hydraulic channel as the 

intact cold leg (CL-1) as shown in 

Figure 4 to Figure 6. The drain valve 

was modeled using a trip valve 

component which would open if the 

water level of the lower downcomer 

becomes higher than the set point. 

The hot legs, (HL-1 and HL-2) 

which are located between CL-1 and 

CL-2, and between CL-3 and CL-4, 

respectively, are modeled as blunt 

bodies that penetrate the downcomer. 

The flow areas were calculated by 

using the gap width, perimeter, as 

well as other geometric parameters at 

this section to estimate the equivalent 

thermal hydraulic diameter. 

The direct ECCS bypass fraction 

is calculated based on the flow rates 

of ECCS injection, steam injection, 

and drain flow rate at the lower downcomer 

as follows: 

Bypass fraction = 

M Water_out 

M SI_in +M Condensate 

| | Fig. 3. 

ECC Bypass Fraction of MIDAS Tests. 

M Steam_in is the steam injection mass 

flow rate, and M Condensate is the 

condensate mass flow rate calculated 

as follows: 

M Condensate = M Steam_in – M Steam_out 

4 Results and Discussion 

The model predictions of the bypass 

fraction for all three nodalization 

cases (4, 6 and 12 channels) were 

compared to the experimental data. 

The sample standard deviation of the 

differences between measured values 

and predicted values, RMSE (Root 

Mean Square Error), are presented in 

Table 2. 

For the case with DVI-2 injection 

only (KM109), the RMSEs are 

relatively small and acceptable for all 

three cases with 0.056 for 4 channels 

as a representative case. For the 

injection through DVI-4 only (KM110 

~ KM114), the code over-predicts the 

bypass fraction. This is more distinct 

at lower steam flow and for finer 

nodalization (e.g. 12 channels). For 

the cases with injection through both 


KM114 0.6879 DVI-4 

| | Tab. 1. 

Experimental Conditions of MIDAS Tests [7]. 

where, M SI_in is the total ECCS injection 

mass flow rate, M Water_out is the 

discharged liquid mass flow rate, 

| | Fig. 4. 

MIDAS Nodalization Scheme with 4 Channels. 





| | Fig. 5. 


| | Fig. 6. 


Test No. 

Steam Flow Rate 

(kg/s) 

| | Tab. 2. 

RMSE Calculated Results of Bypass Fraction with Measured Data. 

DVI-2 and DVI-4, the steam flow 

rate seems to govern the prediction 

accuracy. In case of high steam flow 

rate tests (≥ 1.1 kg/s), the SPACE 

code predicted the bypass fraction 

well regardless of the number of 

channels chosen. For the low steam 

flow rate tests (≤ 1.1 kg/s), the RMSE 

is ≥ 0.16 as shown in Table 2. More 

detailed examination is presented 

below. 

4.1 Results of High Steam Flow 

Rate Tests 

The results for the high steam flow 

rate tests (KM100 ~ KM103, and 

Number of Channels 

4 6 12 

KM109 1.8086 0.056 0.078 0.005 

KM100 ~ 103 ≥ 1.1 0.017 0.019 0.017 

KM104 ~ 108 

0.161 0.211 0.287 

≤ 1.1 

KM110 ~ 114 0.252 0.334 0.462 

| | Fig. 7. 

Comparison of the Measured and Calculated ECC Bypass Fraction for the 

High Steam Flow Cases. 

KM109) are presented in Figure 7. For 

the high steam flow rate tests, the 

SPACE code predicts the bypass 

fraction relatively well for all 

nodalization cases. 

The liquid flow pattern for the 

KM100 test (highest steam flow rate 

test) of each nodalization case are 

presented in Figure 8 to Figure 10. 

The liquid flow pattern for the all 

nodalization cases are quite similar. 

The direct bypass phenomena occurs 

in the upper region of the downcomer 

as the ECCS flow joins the high 

velocity steam from the intact cold leg 

and is swept away through the broken 

cold leg. In the case of tests with a 

high steam flow rate, the result of 

the 4 channels nodalization is similar 

to that of 6 and 12 channels. Hence, 

the 4 channels representation is considered 

a reasonable approximation. 

4.2 Results of Low Steam Flow 

Rate Tests 

The results for the low steam flow rate 

tests (KM104 ~108 and KM110 ~114) 

are presented in Figure 11. Contrary 

to the high steam flow rate cases, for 

the low steam flow rate tests, the 

SPACE code over-predicts the bypass 

fraction for the all nodalization cases. 

The liquid and vapor flow patterns 

of the 6 channels case for the lowest 

steam flow rate test (KM114) are 

presented in Figure 12 and Figure 13, 

respectively. Most of the liquid 

injected from the DVI nozzle is swept 

with the steam flow through the 

break. The test indicated some downward 

liquid flow at this steam flow 

rate. 

In the SPACE code, the interfacial 

friction model is dependent on the 

flow regime of the control volume. 

Thus, for quantitative agreement with 

the MIDAS experimental measurements, 

the estimation of the flow 

regime has to be properly predicted to 

accurately estimate the bypass flow in 

the upper downcomer. The SPACE 

code selects the annular mist flow 

regime based on the volume average 

conditions, which explains the deviation 

between the code prediction and 

MIDAS tests in the case of low steam 

flow rate. 

4.3 Results of Condensation 

Fraction 

It is worthy to note that for all the 

studied cases, the code under-predicts 

the condensation fraction as shown in 

the Figure 14. The RMSE based on 

calculated condensation fraction with 

the measured condensation fraction 

data are presented in Table 3. The 

under-prediction tendency is more 

distinct for finer nodalization (e.g. 12 

channels) as depicted in Table 3. This 

may clearly be tied to the heat transfer 

correlation which in turn depends on 

the flow regime. Due to mass conservation, 

the lower condensation rate 

leads to over-estimation of the bypass 




| | Fig. 8. 

Liquid Flow Pattern of KM100 Test Calculation 

with 4 Channels Nodalization. 

| | Fig. 9. 



| | Fig. 10. 




| | Fig. 11. 

Comparison of the Measured and Calculated ECC Bypass Fraction for Low 

Steam Flow Cases. 

| | Fig. 12. 



| | Fig. 13. 

Vapor Flow Pattern of KM114 Test Calculation 


fraction. The problem is aggravated 

for the lower steam flow rate tests, 

since the phase change effect overshadows 

the convective effect. It is 

hypothesized that the bypass flow 

may be influenced by the interplay 

between thermal and inertial effects, 

particularly at the lower steam flow 

rate test conditions. 

5 DVI Location Effect for 

the Low Steam Flow Rate 

Test 

As shown in the Figure 4 to Figure 6, 

the DVI channels and the broken 

channel share the same channel. With 

this nodalization, the most of the 

injected liquid flows into the control 

volume directly connected to broken 

cold leg. Since the steam flow for this 

volume is very high, the flow regime 

becomes co-current annular mist flow. 

With co-current annular flow, the 

injected water from the DVI swept 

away to the break. 

To further examine this phenomenon, 

we carried out an additional 

calculation. We selected the 6 channels 

representation. This time, however, 

the DVI-4 is connected to a 

channel next to the channel where 

broken cold leg is connected as shown 

in the Figure 15. The DVI channels 

were separated from the broken channel 

(or cold leg channel), artificially. 

The bypass and condensation 

fraction results of the existing and new 

nodalization cases with 6 channels are 

compared with KM114 test conditions 

in Table 4. Clearly, the new nodalization 

better predicts the bypass and 

condensation fraction. While the 

existing nodalization predicts a bypass 

fraction of 0.714, the new nodalization 

predicts a bypass fraction of 0.091 

with only about 16 % deviation. 

| | Fig. 14. 

Comparison of the Measured and Calculated Condensation Fraction. 

| | Fig. 15. 

New Nodalization Scheme for 6 Channels. 

Test No. 

Steam Flow Rate 

(kg/s) 

Number of Channels 

4 6 12 

Case 

Bypass 

Fraction 

Condensation 

Fraction 

KM109 1.8086 0.029 0.036 0.052 

KM100 ~ 103 ≥ 1.1 0.078 0.094 0.116 

KM104 ~ 108 

0.090 0.113 0.144 

≤ 1.1 

KM110 ~ 114 0.082 0.103 0.138 

| | Tab. 3. 

RMSE Calculated Results of Condensation Fraction with Measured Data. 

Measured value 0.109 0.231 

Existing nodalization 0.714 0.131 

New nodalization 0.091 0.203 

| | Tab. 4. 

Bypass and Condensation Fraction Results Comparison 

in Case of 6 Channels for KM114 Test. 





| | Fig. 16. 


with 6 Channels of New Nodalization. 

Similarly, while the existing nodalization 

predicts a condensation fraction 

of 0.131, the new nodalization predicts 

a condensation fraction of 0.203 with 

only about 12 % deviation. 

The liquid and vapor flow pattern 

diagrams of the 6 channels case for 

the KM114 test are presented in 

Figure 16 and Figure 17 for the new 

nodalization, respectively. The liquid 

flow issuing from DVI-4 becomes 

continuous downward flow as shown 

in Figure 16. This shows the importance 

of proper representation of the 

flow regime. Given that the new 

nodalization does not strictly reflect 

the actual experimental arrangement, 

the proper nodalization scheme needs 

to be further developed. 

6 Conclusions 

In this paper, a nodalization sensitivity 

analysis for the MIDAS test was 

performed using the SPACE code. 

Three cases were modeled: 4, 6, and 

12 channels. 

In the case of high steam flow rate 

with DVI injection from both sides 

tests (KM100 ~ KM103) and DVI-2 

injection test (KM109), the SPACE 

code estimated the bypass fraction 

relatively accurately and the nodalization 

scheme does not affect 

the code results much. From the 

efficiency, 4 channel representation 

is recommended for SPACE code 

nodalization. 

Similar to RELAP5 calculation, the 

SPACE code was unable to accurately 

predict the bypass fraction for the low 

steam flow rate MIDAS tests (KM104 

~ 108 and KM 110 ~ 114) regardless 

of the nodalization used. From a 

safety perspective, over-prediction of 

the bypass flow is conservative for a 

LOCA simulation. 

The over-prediction at low steam 

flow may be attributed to the difficulty 

to correctly represent the flow regime 

in the vicinity of the broken cold leg. 

This led to under-prediction of 

| | Fig. 17. 

Vapor Flow Pattern of KM114 Test Calculation 

with 6 Channels of New Nodalization. 

condensation rate and over-prediction 

of interfacial shear. When the DVI 

channels were horizontally shifted 

with respect to the break channel, the 

SPACE better predicted the bypass 

fraction for the lowest steam flow rate 

MIDAS test (KM114). This fictitious fix 

proves the hypothesis but the result 

should be treated with discretion. 

7 Acknowledgments 

This research was supported by the 

2017 Research Fund of the KINGS 

(KEPCO International Nuclear 

Graduate School), Republic of Korea. 

References 

[1] ***, KHNP, Topical Report on the SPACE 

code for Nuclear Power Plant Design, 

KHNP/TR-0032/2017, 2017. 

[2] S.Y. Lee and C.H. Ban, Code-Accuracy- 

Based Uncertainty Estimation (CABUE) 

Methodology for Large-Break Loss-of- 

Coolant Accidents, Nuclear Technology, 

Vol. 148 Issue 3, pp.335-347, 2004. 

[3] ***, KHNP, Topical Report for the 

LBLOCA Best-Estimate Evaluation 

Methodology of the APR1400 Type 

Nuclear Power Plant, KHNP/TR-0018/ 

2010, 2010. 

[4] S.W. Lee and S.J. Oh, APR1400 Large 

Break Loss of Coolant Accident Analysis 

using KREM methodologies, 2003 KNS 

Autumn Meeting, KNS, 2003. 

[5] S.W. Lee, H.G. Kim, and S.J. Oh, 

Assessment of APR1400 ECCS Capability 

against Large-Break LOCA Scenario 

by RELAP5/MOD3 Code, Nuclear 

Technology, Vol. 158 Issue 3, 

pp.396-407, 2007. 

[6] B.J. Yun, H.K. Cho, T.S. Kwon, C.H. Song, 

J.K. Park, and G.C. Park, Experimental 

Observation on the Hydraulic 

Phenomena in the KNGR Downcomer 

during LBLOCA Reflood Phase, 2000 

KNS Spring Meeting, KNS, 2000. 

[7] ***, KAERI, Direct Vessel Injection Test 

Using the MIDAS Test Facility-ECC Direct 

Bypass Test, MIDAS-QLR-009, 2001. 

[8] W.A. Carbiener and R.A. Cudnik, 

Similitude Considerations for Modeling 

Nuclear Reactor Blowdowns, Tran. Am. 

Nucl. Soc., Vol. 12, pp.361, 1969. 

[9] B.J. Yun et al., Direct ECC Bypass 

Phenomena in the MIDAS Test Facility 

during LBLOCA Reflood Phase, 

KNS Vol. 34, pp.421-432, 2002. 

[10] ***, KAERI, Scaling Analysis of the 

Thermal Hydraulic Test Facility for the 

Large Break LOCA of KNGR, KAERI/ 

TR-1878/2001, 2001. 

Authors 

Shin Eom 

Graduate Student 

Professor Dr. Seung-Jong Oh 

Professor Dr. Aya Diab 

Department of NPP Engineering 

KEPCO International Nuclear 

Graduate School (KINGS) 

Ulsan, Korea 




The Application of Knowledge 

Management and TRIZ for solving the 

Safe Shutdown Capability in Case of Fire 

Alarms in Nuclear Power Plants 

Chia-Nan Wang, Hsin-Po Chen, Ming-Hsien Hsueh and Fong-Li Chin 

1 Introduction The 2011 the Fukushima nuclear disaster in Japan was caused by a failure in the safe shutdown 

system. The severing of power systems incapacitated several of the shutdown devices, thereby hindering the removal of 

excess heat from the reactor. Under these conditions, zirconium on the protective cover of the fuel rods reacted with the 

cooling water to produce hydrogen gas. The resulting explosion fractured the containment building, thereby allowing 

the escape of radioactive materials into the surrounding environment. 

Nuclear power plants designed in 

the U.S. must conform to regulations 

outlined by the Nuclear Regulatory 

Commission (NRC). The safe shutdown 

capabilities of a facility are 

documented in the Final Safety 

Analysis Report (FSAR), which must 

be submitted to authorities prior to 

the licensing of operations. Facility 

upgrades are also subject to approval. 

Operating specifications include 

shut-down procedures to be implemented 

in the event of an earthquake 

or other environmental disaster. In 

1979, the NRC proposed a number of 

fire safety measures [10CFR50 App.R]; 

however, the complexity of nuclear 

facilities has greatly hindered implementation 

and enforcement. Nuclear 

power plants are required to have two 

independent safe shutdown systems, 

either of which must be able to 

manage plant operations during the 

transition from operating phase to 

cold shutdown. The simultaneous 

failure of both of systems would lead 

to a catastrophic collapse of the entire 

system. This study sought to sought to 

improve the safe shutdown performance 

of nuclear power plants in the 

event of fire. We compiled a wide 

range of data pertaining to post-fire 

safe shutdown of nuclear power 

plants, while dealing with each system 

and its components as discrete units. 

Our main objectives were as follows: 

1. To compile a knowledge base 

of issues related to hazards in 

nuclear power plants: The 

knowledge base defines the safe 

shutdown system used in each fire 

zone, describes the components 

used in each system, and organizes 

the shutdown processes in the 

form of a flowchart. 

2. To assess the components of the 

safe shutdown systems using the 

Teoriya Resheniya Izobreatatelskih 

Zadatch (TRIZ) method: 

We defined the attributes and 

parameters of various problems 

associated with safe shutdown 

equipment and developed models 

for each individual problem using 

TRIZ to identify feasible means of 

improvement. 

3. Improve the safety regulations 

of nuclear power plants based 

on case studies and a literature 

review: We formulated a novel 

approach to the analysis of case 

studies with the aim of facilitating 

the identification of omissions 

and flaws in current evaluation 

standards. 

2 Literature review 

Prior to 1974, there were only two 

clauses in the national fire regulations 

(U.S.): 10CFR50 Appendix A (fire 

protection) General Design Criteria 

(GDC) and R.G 1.70.4. In November 

1975, after the fire at Browns Ferry 

Nuclear Power Plant, the NRC 

published the Standard Review Plan 

9.5-1. In May 1976, the BTP APCSB 

9.5-1App.A (Nuclear Power Plant 

Fire Guidelines) came into effect for 

nuclear power plants seeking to obtain 

building permits after July 1 [NRC, 

1976], 1976. In August 1977, the NRC 

published the Generic Letter 77-02 

[USNRC, 1977], addressing issues 

pertaining to administration, the 

regulation of organizations, firefighting 

procedures, and quality 

control measures. In 1980, the NRC 

drew up 10CFR50 Appendix R (fire 

protection program), detailing the 

requirements of all nuclear power 

plants that went into operation prior 

to January 1st 1979. In February 1981, 

the NRC announced 10CFR50.48 

(fire protection) as the standing 

regulations for nuclear power plant 

fire safety [Information Notice, 1984]. 

Compliance with 10 CFR 50 App. R 

was not mandatory for all nuclear 

power plants operating before 

January 1, 1979 (pre-1979 plants); 

however, they had to follow the 

basic design requirements. In contrast, 

nuclear power plants operating 

since January 1, 1979 (post-1979 

plants) have had to comply with BTP 

CMEB 9.5-1, Revision 2 [CRF, 1979] 

In the case study of this paper, an 

operating license was obtained for 

reactor 1 on July 27, 1984. It should 

therefore have been subject to BTP 

CMEB 9.5-1 Rev.2 [July 1981]; however, 

Section 9.5.1 of the FSAR from 

the later Maanshan Nuclear Power 

Plant refers to Appendix A to APCB 

9.5-1 [NRC Branch Technical Position, 

1981]. As a result, both were used 

as references. Taiwan uses the fire 

regulations of 10 CFR 50 Appendix R 

as the basis for fire inspections; 

however, these regulations are somewhat 

rudimentary [TPC, 1999]. 

In U.S. federal regulations 10 

CFR 50 Appendix A, General Design 

Criterion 3 specifies the basic fire 

protection requirements for nuclear 

power plants [CFR, 2012]. For 

example, the design of the fire protection 

system must ensure that even in 

the event of damage of improper use, 

the safety performance would not be 

impaired. Fire protection policy based 

on defense-in-depth is used to protect 

the shutdown system as follows: 

1) preventing the occurrence of fires, 

2) ensuring the rapid detection, control, 

and extinguishing of fires that 

do occur, and 

3) ensuring the normal operation 

of the safe shutdown system if a 

fire cannot be extinguished [NCR, 

1975]. 

95 

OPERATION AND NEW BUILD 


The Application of Knowledge Management and TRIZ for solving the Safe Shutdown Capability in Case of Fire Alarms in Nuclear Power Plants ı Chia-Nan Wang, Hsin-Po Chen, Ming-Hsien Hsueh and Fong-Li Chin


OPERATION AND NEW BUILD 96 

In 10 CFR 50 Appendix R, Section 

III.G.1 are specified the fire protection 

requirements for the emergency 

shutdown of nuclear power plants: 

1. One train of systems necessary to 

achieve and maintain hot shutdown 

conditions from either 

the control room or emergency 

control station(s) is free of fire 

damage. 

2. Systems necessary to achieve and 

maintain cold shutdown from 

either the control room or emergency 

control station(s) can be 

repaired within 72 hours [NRC, 

2007]. 

In 10 CFR 50 Appendix R, Section 

III.G.2 are outlined specific isolation 

requirements for redundant cables 

and safe shutdown systems within the 

same fire compartment: “Except as 

provided for in paragraph G.3 of this 

section, where cables or equipment, 

including associated non-safety 

circuits that could prevent operation 

or cause maloperation due to hot 

shorts, open circuits, or shorts to 

ground, of redundant trains of systems 

necessary to achieve and maintain hot 

shutdown conditions are located 

within the same fire area outside of 

primary containment, one of the 

following means of ensuring that 

one of the redundant trains is free 

of fire damage shall be provided.” 

10 CFR 50 Appendix R, Section 

III.G.3 specifies the situations in 

which fire compartments are required 

to have dedicated safe shutdown 

capabilities involving modification or 

replacement of dedicated cables and/ 

or circuitry. 

Cables, systems and components 

should be independent of area, room, 

zone if the following conditions are 

met: 

1. Where the protection of systems 

whose function is required for hot 

shutdown does not satisfy the 

requirement of paragraph G.2 of 

this section; or 

2. Where redundant trains of systems 

required for hot shutdown located 

in the same fire area may be subject 

to damage from fire suppression 

activities or from the rupture or 

inadvertent operation of fire 

suppression systems. 

3. Furthermore, fire detection and a 

fixed fire suppression system shall 

be installed in the area, room, or 

zone.” 

Guidance IX of the NRC Information 

Notice 84-094 lists the minimum safe 

shutdown monitoring parameters 

accepted by the NRC [NRC Information 

Notice, 1984]. 

NUREG-1852 presents the feasibility 

and reliability criteria [NUREG, 

2007] accepted by the NRC in the 

event that Operator Manual Actions 

(OMAs) are used to perform post-fire 

safe shutdown. 

The above fire protection regulations 

provide the parameters relevant 

to safe shutdown capabilities and 

fire protection. We compared these 

parameters with those of the nuclear 

power plant in our case study to 

identify problems associated with 

safe shutdown capabilities and fire protection. 

However, this is an enormous 

and complex task. Thus, we developed 

an innovative approach to achieve this 

using knowledge management in 

conjunction with TRIZ. 

3 Methodology 

This study sought to improve the safe 

shutdown performance of nuclear 

power plants in the event of fire. 

Knowledge management was first 

used to identify the factors essential 

to safe shutdown. We then sought 

to identify the factors that are not 

adequately addressed in US nuclear 

power regulations. Finally, TRIZ was 

used to guide the formulation of 

recommendations aimed at overcoming 

current regulatory shortcomings. 

3.1 Knowledge management 

and construction of database 

Knowledge management was organized 

into the following phases to 

define core knowledge and construct a 

database for research [Rosner et al., 

1998]. 

Phase 1: Progress from the macroscopic 

system level to the microscopic 

equipment level. 

Phase 2: Identify wiring associated 

with post-fire safe-shutdown. 

Phase 3: Conduct post-fire safe-shutdown 

circuit analysis [Debowski, 

2007]. 

Phase 4: Establish post-fire hot shutdown 

path based on APP.R. 

Phase 5: Construct a distribution of 

post-fire safe hot shutdowns procedures 

throughout the plant. 

Phase 6: Establish basic fire prevention 

database [National Fire Protection 

Association, 2001]. 

3.2 Application of TRIZ to 

improve safe shutdown 

system 

TRIZ is a highly reliable problemsolving 

method, which was developed 

by Altshuller et al. in his review of over 

300,000 patents between 1946 and 

1985 [Altshuller, 1999]. TRIZ is based 

on the concept of abstraction, taking 

an algorithmic approach to the invention 

of new systems and the refinement 

of old systems [Mann, 2007]. 

In this study, we combined 

knowledge management and TRIZ 

in the development of a novel 

method by which to improve safe 

shutdown procedures, as follows 

(comp. Figure 1): 

1. Collect data pertaining to 

current conditions and existing 

designs. 

2. Formulate standards and definitions 

based on existing regulations 

related to post-fire safe 

shutdown. 

3.1. Define and clarify issues. If 

sufficient data is available, proceed 

to Step 4; otherwise, proceed 

to Step 3.2. 

3.2. Search available data and current 

regulations for designs that could 

be improved through knowledge 

management. Compare results 

with the safety conditions stipulated 

in current regulations, and 

then conduct enhancement 

analysis based on the following 

knowledge management techniques: 

(1) establish operating 

standards; (2) identify interdependent 

relationships between 

existing systems; (3) organize 

operational procedures; (4) set 

safe shutdown function codes; 

(5) establish safe shutdown path 

combinations; (6) compare results 

with regulation requirements; 

(7) identify all devices associated 

with post-fire safe shutdown 

(8) set operating status parameters; 

(9) compare results with 

corresponding wire/circuit design 

data of original equipment; 

(10) identify wires/circuits associated 

with post-fire safe shutdown; 

(11) conduct wire/circuit failure 

analysis; (12) compile results of 

wire/circuit analysis in the form 

of a database; (13) establish wire/ 

circuit paths in fire zones. If 

level-by-level comparisons show 

that the existing system complies 

with regulations, then proceed 

to Step 6. 

4. Search through database of 

existing system for instances of 

mismatch with regulations. If the 

database does not meet safety 

requirements, then return to 

Step 1. If the database meets 

safety requirements, then perform 

an assessment of ... 

5. Determine whether non-compliant 

systems affect safe shutdown 

capabilities. 




6. If the current safe shutdown 

capabilities meet or surpass those 

stipulated in the regulations, then 

proceed to Step 8. If the current 

safe shutdown capabilities do 

not meet those stipulated in the 

regulations, then proceed to 

Step 7. 

7. Use TRIZ to search for improvement 

methods, while taking into 

account construction costs and 

probable benefits. 

8. If the current status of the nuclear 

power plant complies with the 

basic safety conditions stipulated 

in the regulations, then it is 

assumed that the plant possesses 

satisfactory safe shutdown capability. 

4 Empirical results 

4.1 Application of knowledge 

management 

We selected a nuclear power plant for 

use as a case study. Fire compartments 

were drawn up according to the floor 

plan and final safety analysis report 

(FSAR) (Table 1). Most nuclear power 

plants include the following: containment 

or drywell building, reactor 

(auxiliary) building, turbine building, 

intake structure (screenhouse), fuel 

building, diesel generator building. In 

principle, if an area is enclosed by 

fire-shielding concrete walls, then 

smaller fire zones can be drawn up 

within the larger fire zone in order to 

differentiate between similar paths. In 

this case, the original fire compartment 

C101 includes numerous rooms. 

ESF 4.16KV SWGR ROOM A was designated 

fire compartment 5 in order to 

re-partition the space according to 

their function. 

Phase 1: Progress from the macroscopic 

system level to the microscopic 

equipment level. 

Step 1: Define the scope of the 

post-fire safe shutdown capacity. 

Shutdown objectives include the 

following: 1. reactivity control; 

2. reactor coolant makeup; 3. reactor 

heat removal; 4. process monitoring; 

5. supporting functions; 6. achieve hot 

Unit 

FL 

No. 

FL 

Code 

Factory 

building 

| | Tab. 1. 

Examples of partitioning fire compartment in nuclear power plant. 

| | Fig. 1. 

Application of knowledge management and TRIZ to improve post-fire safe shutdown performance. 

standby status and maintain systems 

required to (i) prevent fire damage, 

(ii) enable the power unit to last 

through hot standby status for over 

72 hours, and (iii) receive power 

from emergency power system; 

7. achieve cold shutdown status 

and maintain systems required to 

prevent fire damage. The above 

objectives do not cover the following: 

(1) seismic category I criteria, 

(2) single failure criteria, or (3) other 

plant accidents. 

Step 2: Define the core knowledge 

parameters of post-fire safe shutdown 

capacity. 

1) Establish map of interdependence 

among systems employed in 

post-fire safe shutdown. 2) Define 

operating procedures of post-fire safe 

shutdown systems and construct 

operational flowchart. 3) Define 

parameters of post-fire safe shutdown 

functions and construct function code 

list. 4) Identify function code combinations 

required for post-fire safe 

shutdown path and construct path 

combination table. 

Step 3: Refer to existing regulations 

NEI-0001 and RG1.189 of 

US–NRC to confirm that the post-fire 

safe shutdown and wire/circuit 

analysis methods are acceptable. 

First step: Determine Regulatory 

Requirements 

Space 

FL Name 

1 1 C101 CTRL 80' ESSENTIAL CHILLER ROOM A 

1 2 C101 CTRL 80' ESF 4.16KV SWGR ROOM A 

1 3 C101 CTRL 80' ESF SWGR ROOM A 

1 4 C102 CTRL 80' ESSENTIAL CHILLER ROOM B 

1 5 C102 CTRL 80' ESF 4.16KV SWGR ROOM B 

The primary regulations include 

10 CFR 50 Appendix A, General Criterion 

3, and 10 CFR 50 Appendix R. 

Second step: Determine SSD 

Functions, Systems, and Path 

This is meant to ensure that any 

single fire within any fire area in the 

nuclear power plant does not lead to 

incidents such as furnace core meltdown, 

loss of reactor cooling water, or 

damage to the primary containment 

structure. To achieve this objective, 

the safe shutdown functions of the 

reactor must first be confirmed and 

the existing system equipment and 

pipelines in the plant analyzed and 

combined to form a safe shutdown 

path as well as achieve and maintain 

the safe shutdown status of the power 

unit. 

Third step: Select Equipment 

Required for Post-Fire Safe shutdown 

This equipment is used for post-fire 

safe shutdown or to serve as a backup 

in the event of fire-induced malfunctions. 

Fourth step: Select Wires/Circuits 

for Post-Fire Safe shutdown 

These wires/circuits are used for 

post-fire safe shutdown or to serve as a 

backup in the event of fire-induced 

malfunctions 

Below are the basic assumptions 

used in the analysis of post-fire safe 

shutdown capacity: 

1. Only one fire occurs in the plant at 

any one time. 

2. In the event of loss of external 

power due to fire, systems can 

provide backup power for at least 

72 hours. 

3. The only equipment or system 

malfunctions are associated 

directly with the fire. 

4. After the safe shutdown of the 

power unit, there are no additional 

accidents due to plant design 






Drawing 

No. 

including (1) loss-of-coolant accidents 

(LOCA), (2) main steam line 

breaks (MSLB), (3) steam generator 

tube ruptures (SGTR), or (4) 

control rod ejection accidents 

(REA.) 

5. Any wires or equipment in the area 

of a fire that are not protected by 

fire wrap are burned, unless the 

results fire disaster analysis prove 

otherwise. 

6. Fire-induced wire/circuit damage 

can lead to open circuits, short 

circuits, hot shorts, and shorts to 

ground. 

7. The valves, pipelines, tanks, or 

incombustible instrument wires 

affected by the fire do not cause 

damage to the pressure boundary. 

8. Despite fire damage to instruments, 

the pressure boundaries 

of fluids within them are not 

damaged. 

9. Motor-operated valves do not malfunction 

due to fire damage to 

power wires, but they may malfunction 

following fire damage to 

control circuits. 

10. During post-fire safe shutdown, 

power units may be controlled 

manually using existing equipment, 

as long as the fire does not 

directly hinder such operations. 

The scope of the core knowledge 

relating to post-fire safe shutdown 

capacity can be clearly defined and 

verified based on the analytical 

methods proposed in NEI 00-01 Rev. 2 

and the target performance of safe 

shutdown capacity. 

Step 4: Establish inventory of 

post-fire safe shutdown equipment. 

Determine the specifications of 

post-fire safe shutdown equipment 

(Table 2): 1. attributes, 2. operating 

status, and 3. path parameters [NFPA, 

2001]. 

Function Description 

Old System 

Code 

SSD 

Code 

1 RCS BB B1/B2 

2 RCS-ACCUM ISO BH B1/B2 

3 CVCS HHSI BG B5/B6 

4 CVCS HHSI SUP BG BS56 

5 SIS HHSI BH B7/B8 

6 CVCS RCP BG C5/C6 

| | Tab. 2. 

Post-fire safe shutdown system parameters for case study. 

Phase 2: Identify wire/circuits 

associated with post-fire alarm safe 

shutdown. 

Step 1: Identify wires and circuits 

associated used with post-fire safe 

shutdown equipment. 

Using the original design data of 

the plant, list every power wire and 

control wire associated with the 

post-fire safe shutdown equipment. 

Step 2: Determine the specifications 

of all wire/circuits associated 

with post-fire safe shutdown. Set the 

parameters of operating status, 

equipment attributes, and the safe 

shutdown paths to which they belong. 

Step 3: Refer to the existing database, 

control wiring diagram (CWD), 

and control logic diagram (CLD) to 

identify the control wires associated 

with each piece of equipment. 

Step 4: Compile an inventory of 

wires associated with post-fire safe 

shutdown (NEI, 2009). 

A series post-fire safe shutdown 

path (Code: HSD-P1): 

(A1+A3)+(B1+B3+B5+B7+B9)+ 

(D1+E1+F1+G1+H1+I1+J1+K1+ 

L1+M1+N1+P1+S1+U1+V1+ 

W1+X1+Y1.) 

B series post-fire safe shutdown 

path (Code: HSD-P2): 

(A2+A4)+(B2+B4+B6+B8+B10)+ 

(D2+E2+F2+G2+H2+I2+J2+K2+ 

L2+M2+N2+P2+S2+U2+V2+ 

W2+X2+Y2) 

Taking the plant from operating 

to hot shutdown requires that the 

equipment listed above be operational. 

These devices must also be 

included in independent paths 

HSD-P2 or HSD-P1. 

Example of system parameters 

(Table 2) and shutdown path: The 

power for the motor driven auxiliary 

feed water pump (A-1M-AL-P017) in 

auxiliary feed water system of Series A 

(system parameter B3) is supplied by 

Class 1E 4.16kV Bus A-1E-PB-S01 (PB 

system). In post-fire safe shutdown 

operation mode, this bus is powered 

by the emergency diesel generator in 

Series A (system parameter D1). Thus, 

a supply of lubricating oil and a fuel 

(KJ system) must be available for the 

emergency diesel generator. At the 

same time, it is essential that the 125V 

DC electrical system (PK system) 

provide power to the control panel 

of the emergency diesel generator 

A-1J-ZD-P001. The emergency diesel 

generator is uses a jacket water-cooler 

A-1M-KJ-X072 running off of a 

seawater system (EF system); the 

power for the seawater pump A-1M- 

EF-P103, P104 is provided by the 

4.16kV bus A-1E-PB-S01 (PB system.) 

This is an example of the analysis used 

to establish the interdependence of 

systems within a given post-fire safe 

shutdown path. 

Phase 3: Establish an inventory of 

wire/circuits associated with post-fire 


Step 1: Use the wire/circuit inventory 

established in previous phase to 

conduct effect analysis of fire-induced 

wire/circuit failures. Analyze fire- induced 

circuit failures (power, control, 

instrument) associated with each piece 

of equipment, based on inventory of 

equipment used in post-fire safe shutdown. 

These wire/circuits can be 

divided into two categories: those 

necessary to post-fire hot shutdown 

and those necessary to post-fire safe 

shutdown. Single-line diagrams, CLDs, 

and CWDs of post-fire safe shutdown 

equipment in the original design are 

used to investigate fire-induced circuit 

failures, as follows: 

(1) Categorization of wires required 

for post-fire hot shutdown: 

a. Power and control wires required 

for manual operation of equipment 

used in post-fire hot shutdowns 

b. Power and signal wires for instruments 

used in process monitoring 

during post-fire hot shutdown 

c. Wires that could cause the malfunction 

(through fire-induced 

circuit failure) of equipment required 

for post-fire hot shutdowns 

d. Wires that could cause the malfunction 

of components (through 

fire-induced circuit failure) in 

high/low pressure system 

(2) Categorization of wires required 

for post-fire safe shutdown: 

a. Power and control wires required 

for manual operation of equipment 

used in post-fire cold shutdowns 

b. Wires that could cause the malfunction 

(through fire-induced 

circuit failure) of equipment required 

for cold shutdowns 

c. Wires that could cause the malfunction 

of components crucial to 

shutdowns (through fire-induced 

circuit failure) 

Fire-induced circuit-failure parameters 

were established as follows: 

1) fire-induced circuit-failure equipment, 

2) operating status parameters, 

3) fire-induced circuit-failure parameters, 

and 4) wire/circuit attribute 

parameters. 

Step 2: Use the circuit-failure 

parameters to construct a table for the 

analysis of circuits used in post-fire 


Effect analysis of fire-induced 

circuit failures associated with the 

post-fire safe shutdown equipment, 

including open circuits, short circuits, 

hot shorts, and shorts to ground (445 

items in total). This analysis produced 

5,149 results. 




Cable No. 

SSD 

Code 

SSD Equipment 

No. 

| | Tab. 3. 

Examples of post-fire safe shutdown cable paths. 

After referring to the parameters 

associated with post-fire safe shutdown 

equipment in the previous step, 

the fire-induced circuit-failure effect 

parameters and wire/circuit attributes 

were compiled into a post-fire 

safe shutdown circuit analysis table. 

Four types of parameter were 

required: 1) fire-induced circuitfailure 

equipment, 2) operating status 

parameters, 3) fire-induced circuitfailure 

parameters, and 4) wire/ 

circuit attribute parameters. 

Step 3: The regulations stipulate 

special requirements for the wiring 

involved in hot shutdowns; therefore, 

the scope of the core knowledge was 

defined as the wires associated with 

post-fire hot shutdowns. 

Step 4: We establish an inventory 

of the wires involved in post-fire safe 

hot shutdown. 

Phase 4: Establish a path associated 

with post-fire hot shutdown for 

use as a reference based on the special 

requirements in APP.R with regard to 

wires associated with hot shutdown. 

Step 1: Define the scope of core 

knowledge and the wires associated 

with post-fire hot shutdown. 

Step 2: Set the relevant wire/ 

circuit parameters, equipment operating 

status parameters, equipment 

attribute parameters, and safe shutdown 

path parameters. 

Step 3: Refer to the existing wire/ 

circuit layout program SETROUTE in 

the original design to derive the circuit 

layout. The fire zones will need to be 

updated, as the original layout 

program uses the old fire zones. To 

facilitate analysis, the fire zones, 

equipment specifications, safe shutdown 

paths, and operating status 

parameters must be added to the database 

of the wire/circuit layout. 

Step 4: Establish an inventory of 

wire/circuit paths involved in post-fire 


Step 5: Establish the post-fire 

alarm safe hot shutdown path form 

(Table 3). Compile a report of wire/ 

circuit paths involved in post-fire safe 

hot shutdown. The nuclear power 

plant in the case study has two power 

units. Unit 1 contains 1,189 wires and 

17,379 items, whereas Unit 2 contains 

SSD 

Path 

SSD Cable 

Type 

1,184 wires and 17,233 items. Thus, 

there are 2,373 wires associated with 

post-fire safe hot shutdown. The 

organization of the report is based on 

the number system used for the safe 

shutdown equipment, the attribute 

categorization of the wires, their 

origin and destination, the numbering 

of the wire/circuit raceways, and the 

fire zones through which they pass. 

Phase 5: Construct the distribution 

of post-fire safe hot shutdowns 

throughout the entire plant. 

Step 1: Define the scope of the core 

knowledge and the post-fire safe hot 

shutdown path. 

Step 2: Set the fire zones to their 

corresponding parameters. 

Step 3: Based on the wire/circuit 

layout program, identify the fire zones 

through which each wire passes. 

Step 4: Establish the distribution 

of the post-fire hot-shutdown function 

codes and replot the post-fire hotshutdown 

tray routing diagram in 

order to obtain an overview of the safe 

hot shutdown capacity throughout 

the entire plant. 

Example: Series A is presented in 

red and series B in green. The safe 

shutdown cable path in the original 

SETROUTE and corresponding function 

code are used to obtain the safe 

shutdown path and function code of 

each fire containment zone (Table 4). 

Phase 6: Establish a database of 

items pertaining to basic fire prevention. 

Basic fire prevention includes a 

wide range of items: (1) basic data of 

fire zones, (2) firefighting equipment 

in fire zones, (3) fire dampers, (4) fire 

doors, (5) combustion load of fire 

zones, (6) list of fire zones adjacent to 

each fire zone (7), inventory of heat 

generated by all combustible items. 

4.2 Application of TRIZ 

The proposed knowledge management 

approach revealed that fire 

compartments 1 and 17 do not comply 

with some regulations [10 CFR 50.48 

APP.R]. Specifically, Wires involved in 

post-fire safe hot shutdown must not 

pass through the same fire compartment 

without the implementation of 

suitable fire protection measures. The 

FROM No. Raceway No. FZ 

B1EEFHCC8SA H2 B-EF-HV203 HSD-P2 HSD-S 1JZJP061E-F 1 B1EZJG2TSRH 20 

B1EEFHCC8SA H2 B-EF-HV203 HSD-P2 HSD-S 1JZJP061E-F 2 B1EZJG2TUAG 20 

B1EEFHCC8SA H2 B-EF-HV203 HSD-P2 HSD-S 1JZJP061E-F 3 B1EZJG2TUAF 20 

FL FL No. HSD Path No. SSD Path 

1 C101 D1 HSD-P1 

1 C101 H1 HSD-P1 

1 C101 I2 HSD-P2 

1 C101 K2 HSD-P2 

| | Tab. 4. 

Example distribution list of fire alarm safe hot shutdown function codes. 

| | Fig. 2. 

Qualitative analysis model for identification 

of problem. 

| | Fig. 3. 

Standard solutions for eliminating harmful 

effects of fire. 

passage of series A and B wires 

through FZ 1 and FZ 17 renders this 

area vulnerable to fire damage [Hua 

and Yang, 2006]. The structure of this 

problem is modeled in Figure 2. 

Figure 3 presents a qualitative 

field model illustrating the association 

between completeness and damage, 

revealing the first problems to be 

eliminated or controlled in a standard 

solution. 

In this case, the designers used 

XPE/Cl.S.PE cables with heat 

resistance of 90 °C. Their Q value 

(Bench-Scale HRR per Unit Floor 

Area) is 204 kW/m 2 , which means 

that they are classified as safe, even in 






| | Fig. 4. 

Parameter attribute problem model for fire damage to the cable. 

the event of fire; i.e., they have a 

high ignition point and low release 

of heat. The conductivity of the cable 

helps to maintain its structural integrity 

[NUREG, 2010]. The first 

physical contradiction appears when 

the temperature exceeds 100 °C. 

There are four steps that can be taken 

to combat this: spatial separation, 

temporal separation, condition separation, 

and separation of system 

levels. These are used to perform 

separation of fire areas, cable burn 

time, burning conditions, and safe 

shutdown system levels (Figure 4). 

All cables must remain reliable 

along their entire length in order to 

ensure a safe shutdown. The fact that 

fire damage can compromise 

reli ability leads to the second technical 

contradiction. 

We constructed a 39X39 contradiction 

matrix to be compared with 

the 40 Inventive Principles based on 

the problem model established on 

structural attributes and parameter 

attributes. Comparison of temperature 

and reliability resulted in the 

selection of the following inventive 

principles: 

# 3: Local quality 

#10: Preliminary action 

#19: Periodic action 

#35: Parameter changes 

A panel of experts decided to disregard 

Principle #19. Principle #35 

was not applicable because the cables 

had already been laid. Principles #3 

and #10 were implemented for 

reasons outlined in the following: 

Inventive principle #3 (local quality): 

3a. Change an object’s structure from 

uniform to non-uniform, change 

an external environment (or 

external influence) from uniform 

to non-uniform. 

3b. Make each part of an object 

function in conditions most 

suitable to its operation. 

3c. Make each part of an object fulfill a 

different and useful function. 

Improvement requirements and 

feasible methods 

(1) Cables from Series A and B should 

be separated by at least 20 feet. 

(2) Built-in discrete fire detection 

systems should be included in all 

areas. In the original design, FZ 1 

and FZ 17 each had one feedback 

system; however, they are now 

segmented into a feedback loop for 

each area [Generic Letters, 1983]. 

(3) Install close-spaced, open-head 

sprinklers. According to GL 83-33, 

Position 2: “In many plant areas, 

the erection of physical barriers 

between redundant shutdown 

systems is precluded by the location 

of cable trays, HVAC ducts and 

other plant features. In such situations, 

the staff has accepted, in 

concept, the use of an automatic 

fire suppression system which 

No. 

Cable 

No. 

SSD 

Code 

Cable 

Code 

SSD 

Equipment No. 

SSD 

Path 

SSD 

Cable Type 

Raceway 

No. 

Rway 

Code 

1 B1EAPHBC2XA K2 EE6 B-AP-LT201 HSD-P2 HSD-S B1EZJF4TXBA WC 

2 B1EBNHAC2XA K2 EE6 B-BNLT961 HSD-P2 HSD-S B1EZJF4TXBA WC 

3 B1EEFHAC2XA H2 EE6 B-EF-PT201 HSD-P2 HSD-S B1EZJF4TXBA WC 

4 B1EEFHAC2XB H2 EE6 B-EF-PT202 HSD-P2 HSD-S B1EZJF4TXBA WC 

5 B1EEFHCC3EA H2 71M3 B-EF-P105 HSD-P2 HSD-S B1EZJF4TEBA SC 

6 B1EEFHCC3EB H2 71M B-EF-P105 HSD-P2 HSD-S B1EZJF4TEBA SC 

7 B1EEFHCC4EA H2 71M3 B-EF-P106 HSD-P2 HSD-S B1EZJF4TEBA SC 

8 B1EKJHBC3LA D2 938 B-KJ-P147 HSD-P2 HSD-S B1EZJF4TPBA SE 

9 B1EKJHBC4LA D2 938 B-KJ-P148 HSD-P2 HSD-S B1EZJF4TPBA SE 

10 BIEPGHHCEHH E2 91I3 B-1E-PG-S01-07 HSD-P2 HSD-S B1EZJF4TIBA SA 

11 B1EPGHHCEHJ E2 91I3 B-1E-PG-S01-07 HSD-P2 HSD-S B1EZJF4TIBA SA 

12 B1EEFHBCBSB H2 939 B-EF-HV230 HSD-P2 HSD-S B1EZJF4TPBA SE 

13 B1EEFHBCBSD H2 939 B-EF-HV230 HSD-P2 HSD-S B1EZJF4TPBA SE 

14 B1EEFHBCBSE H2 939 B-EF-HV230 HSD-P2 HSD-S B1EZJF4TPBA SE 

15 B1EEFHBCJSA H2 C77 B-EF-HV206 HSD-P2 HSD-S B1EZJF4TPBA SE 

16 B1EEFHCC8SA H2 C27 B-EF-HV203 HSD-P2 HSD-S B1EZJF4TPBA SE 

17 B1EEFHCC8SB H2 C27 B-EF-HV203 HSD-P2 HSD-S B1EZJF4TPBA SE 

18 B1EEFHCC8SC H2 C27 B-EF-HV203 HSD-P2 HSD-S B1EZJF4TPBA SE 




22 B1EEFHCCASA H2 C77 B-EF-HV221 HSD-P2 HSD-S B1EZJF4TPBA SE 

| | Tab. 5. 

Parameter attribute problem model for fire damage to the cable. 




Advertisement 

discharges a “water curtain” across 

the boundary areas separating the 

redundant systems. The staff's 

present position is that such systems 

should feature close-spaced, 

open-head sprinklers with water 

discharge initiated by tripping a 

deluge valve activated by crosszoned 

smoke detectors.” Installation 

of a “water curtain” partition 

within the fire compartment 

ensured that both paths were safe 

for post-fire hot shutdown. 

(4) Install fire separation walls. Specifications: 

1. Fire resistance of 

3 hours. 2. Extending from wall to 

wall and from floor to ceiling. 

3. Fire door with a 3-hour rating to 

facilitate access by personnel. 

4. Air ducts that pass through the 

fire separation wall. A fire damper 

with a 3-hour rating must be 

installed within the section that 

passes through the fire separation 

wall. 5. A sleeve must be added to 

all piping that penetrates the fire 

separation wall. The sleeve must 

be sealed using fire-resistant 

material with a rating of 3 h. 6. The 

cable net passing through the fire 

separation wall must be filled with 

fire-resistant materials with a 

rating of 3 h [Generic Letters, 

1986]. The post-fire hot shutdown 

cable for FZ 17 runs through an 

aisle; therefore, fire separation 

walls are feasible only in FZ 1. 

The definitions and improvement 

plans associated with inventive 

principle #10 are as follows: 

10a: Perform all modifications in 

advance. Rearrange cables (relatively 

low cost.) 

10b: Install items or systems in 

advance to ensure that they are 

ready when and where that may 

be. 

FZ 1 contains mostly Series A cables 

as well as 22 Series B cables. The 

post-fire hot shutdown cable list 

( Table 5) revealed that 11 of the 

cables (number 1-11) can be re-laid 

along new paths, such that only 11 

cables (number 12-22) from Series B 

remain within FZ 1. At this point 10b 

No. 

SSD 

Equipment No. 

Status 

9. Symposium zur 

Endlagerung 

radioaktiver Abfälle 

Vorbereitung auf KONRAD – 

Wege zum G2-Gebinde 

18. – 19. April 2018 in 

Hannover 

Inhalte u. a. 

• KFK – Herausforderung aus Sicht eines 

EVUs 

• Endlager Konrad – Baufortschritt und 

Stand der sicherheitstechnischen 

Überprüfung 

• Aspekte der Endlagerungsbedingungen 

• Entsorgung von Altabfällen 

• Vorgehensweisen bei der stofflichen 

Produktkontrolle 

• Optimierte Prüfung von Antragsunterlagen 

Das detaillierte Programm finden Sie in Kürze 

unter: www.tuev-nord.de/tk-era 

Organisation: 

TÜV NORD Akademie 

Meike Langmann 

E-Mail: mlangmann@tuev-nord.de 

Telefon: 040 8557-2046 


1 B-EF-HV203 ON 



4 B-EF-HV222 OFF 


| | Tab. 6. 

Valve states in hot shutdown mode. 




OPERATION AND NEW BUILD 102 

can be used for OMA for manual 

disconnection or operations. 

To prevent equipment malfunction 

due to fire-induced cable damage, a 

fire alarm in FZ 1W signals the control 

room to initiate the first safe shutdown 

path using Series A cables. 

Similarly, a fire alarm in FZ 1E signals 

the control room to initiate the second 

safe shutdown path using Series B 

cables. On-duty staff must take the 

actions presented in Table 6. 

FZ 17 contains mainly Series B 

cables as well as 11 Series A cables. 

The post-fire hot shutdown cable list 

(Table 7) revealed there is no way to 

re-route the cable paths. At this point 

10b can be used for OMA for manual 

disconnection or operations. 

A fire alarm in FZ 17W signals the 

control room to initiate the first safe 

No. 

| | Fig. 5. 

Conformity to regulations in chart form. 

No. Cable No. SSD 

Code 

SSD 

Equipment No. 

Status 






| | Tab. 8. 

Valve states in hot shutdown mode. 

Cable 

Code 

SSD Equipment 

No. 

SSD 

Path 

shutdown path using Series A cables. 

A fire alarm in FZ 17E signals the 

control room to initiate the second 

safe shutdown path using B cables. 

On-duty staff must take the actions 

presented in Table 8. 

The application of TRIZ requires 

that the following conditions be 

satisfied: At least one of the wire series 

has avoided fire damage. For the sake 

of simplicity, we adopted two inventive 

principles: finding local properties 

and taking preliminary actions. 

Nuclear power regulation 10 CFR 

50.48 APP.R stipulates that any wiring 

essential to post-fire hot shutdowns 

that passes through the same fire zone 

require sufficient shielding to protect 

them from fire for at least three h. 

They must also be separated at least 

20 ft, and automatic fire detection 

and extinguishing systems must be 

installed in the fire zone in question. 

All wiring is expected to comply 

with these regulations; however, prior 

to modifications based on the proposed 

method, 22 of the wires in 

Series B were non-compliant. This 

situation could not be foreseen without 

integration of 850,000 pieces of 

path data via knowledge management. 

Among the 22 wires, 11 were 

re-laid and within a fire compartment, 

thereby reducing the number of 

non-compliant wires to 11. According 

to the principle of preliminary action, 

the remaining 11 wires were deemed 

not to affect post-fire hot shutdown 

performance; therefore, even these 11 

wires can be said to comply with 

regulations. 

Regulations stipulate that the 

control room or emergency control 

station be equipped with a series of 

hot shutdown systems capable of 

maintaining hot shutdown conditions 

in the event of a fire in Fire Zones 1 

and/or 17. 

SSD 

Cable Type 

Raceway 

No. 

1 B1EEFHBCBSB H2 939 B-EF-HV230 HSD-P2 HSD-S B1EZJF4TPBA SE 

2 B1EEFHBCBSD H2 939 B-EF-HV230 HSD-P2 HSD-S B1EZJF4TPBA SE 

3 B1EEFHBCBSE H2 939 B-EF-HV230 HSD-P2 HSD-S B1EZJF4TPBA SE 

4 B1EEFHBCJSA H2 C77 B-EF-HV206 HSD-P2 HSD-S B1EZJF4TPBA SE 







11 B1EEFHCCASA H2 C77 B-EF-HV221 HSD-P2 HSD-S B1EZJF4TPBA SE 

| | Tab. 7. 

Series A cables in fire compartment 17 for safe hot shutdown. 

Rway 

Code 

Fire compartments capable of 

withstanding fire for three hours were 

installed between post-fire safe hot 

shutdown wires. The wires were 

horizontally separated by at least 20 ft 

and automatic fire detection and 

extinguishing systems were installed. 

Following these improvements in Fire 

Zones 1 and 17, the post-fire safe hot 

shutdown wires were in full compliance 

with regulations (Figure 5). 

Number of cables that do not 

comply with regulations ≠ Estimated 

number of cables that do not comply 

with regulations = “Do not comply 

with regulations” 

Number of cables that do not 

comply with regulations = Estimated 

number of cables that do not comply 

with regulations = “Comply with 

regulations” 

Number of non-complaint cables in 

case study nuclear power plant = 0 

5 Conclusions 

This study applied TRIZ and 

knowledge management to an actual 

nuclear power plant in order to 

bring the facility up to regulatory 

minimums. Problems were identified 

using hierarchy analysis in conjunction 

with knowledge management for 

the construction of a database. We 

then identified elements that failed to 

meet current regulations. TRIZ was 

used to identify optimal solutions in 

order to minimize the costs involved 

in making improvements to existing 

nuclear power plants. 

The database of wires and circuits 

essential to post-fire safe shutdown 

operations enables operators to 

identify affected systems and decide 

whether immediate isolation is 

required. The implementation of fire 

zones makes it easy to determine 

whether a zone lies along a safe 

shutdown path. The proposed method 




is able to accurately identify zones 

requiring improvement for fire prevention 

or for other safety concerns. Previous 

regulatory evaluations determined 

only the degree of compliance; 

i.e., they gave no indication of whether 

a safe shutdown could actually be 

achieved. The proposed method helps 

to ensure that safe shutdown can be 

achieved, based on the safety requirements 

stipulated in existing regulations. 

Safe shutdown capability can 

be used as a criterion by which to 

identify the elements that cannot 

feasibly conform to regulations, such 

as areas where automatic fire detection 

and extinguishing systems cannot 

be installed. TRIZ is an innovative 

approach to problem-solving. It provides 

a range of possibilities by which 

to solve problems and the results are 

easily compiled to facilitate training 

procedures. Few existing studies on 

nuclear power plants apply directly to 

real-world cases. Knowledge management 

methods enable the construction 

of a knowledge base, thereby providing 

a means by which to integrate 

implicit and explicit knowledge. Its 

systematic integration of analysis and 

comparison data provide valuable a 

reference to practitioners in the field. 

Parameter settings based on 

current regulatory conditions and 

the use of knowledge management 

models enables quicker and more 

precise identification of the improvements 

required for compliance with 

existing regulations. A fire prevention 

database provides a valuable reference 

for the assessment of fire safety. 

Subsequent tasks include developing 

fire models and automatic analysis 

instruments based on fire dynamics, 

fire load, and fire risk probability, 

which all require such databases. The 

basic fire prevention database in this 

study meets the basic requirements 

for fire analysis and can be used for 

future studies of post-fire phenomena 

in nuclear power plants. The procedure 

outlined in this study provides a 

model for safety assessment of current 

nuclear power plants as well as a 

complete research framework for 

other fire-related research in nuclear 

power plants and even other types of 

safety measures. The nuclear power 

plant studied in this paper features 

three-loop pressurized water reactors. 

Therefore the details of the research 

procedure related to the water 

reactors are not necessarily applicable 

to other types of reactor. Data 

collection, analysis, and comparison 

would have to be performed anew 

to confirm its applicability. 

References 

| | Altshuller, G., Shulyak, L., Rodman, S., 

1999. The Innovation Algorithm: TRIZ, 

Systematic Innovation and Technical 

Creativity. Technical Innovation Ctr.: 

Worcester, MA. 

| | Debowski, S., 2007. Knowledge 

Management. Wiley India Pvt. Ltd. 

| | Generic Letters GL 83-33, Position 2, 

1983. Water Curtain, October, 1983. 

| | Generic Letters GL 86-10, Position 3.6.2, 

1986. Fire Stop, April, 1986.(1.) NRC 

BTP APCSB 9.5-1 App. A , (1976) 

Fire Protection guide for Nuclear Power 

Plants, May, 1986. 

| | Hua, Z., Yang, J., Coulibaly, S., Zhang, B., 

2006. Integration TRIZ with problemsolving 

tools: a literature review from 

1995 to 2006. International Journal of 

Business Innovation and Research 1: 

111-128. 

| | Information Notice 84-09, 1984. 

Lessons Learned from NRC Inspections 

of Fire Protection Safe Shutdown 

Systems (10 CFR 50, Appendix R). 

| | Mann, D., 2007. Hands-on Systematic 

Innovation. IFR Press: Clevedon, UK. 

| | National Fire Protection Association 

805, Performance-based Standard for 

Fire Protection for Light Water Reactor 

Electric Generating Plants, 2001 Edition. 

| | NEI 00-01, Rev.2, 2009. Guidance for 

Post Fire Safe Shut Down Circuit Analysis. 

| | NFPA805 National Fire Protection 

Association 805, Performance-based 

Standard for Fire Protection for Light 

Water Reactor Electric Generating 

Plants, 2001 Edition. 

| | NRC Branch Technical Position (BTP) 

9.5-1, 1981. Guidelines For Fire 

Protection For Nuclear Power Plants, 

CMEB, July 1981. 

| | NRC BTP APCSB 9.5-1 App. A , 1976. 

Fire Protection guide for Nuclear Power 

Plants. 

| | NRC Standard Review Plan 9.5-1, 1975. 

Fire Protection Program, November, 

1975. 

| | NRC, 1979. 10 CFR 50 Appendix R to 

Part 50 – Fire Protection Program For 

Nuclear Power Facilities Operating Prior 

To January 1. 

| | NRC, 1984. Information Notice 84-094 

Guidance IX. 

| | NRC, 2007. RG 1.189, Rev. 2 Section 

5.3, Fire Protection of Safe-Shutdown 

Capabilities. 

| | NRC, 2012. 10 CFR 50 Appendix A to 

Part 50, General Design Criterion 3. 

| | NUREG-1852, 2007. Demonstrating the 

Feasibility and Reliability of Operator 

Manual Actions in Response to Fire, 

Final Report, October, 2007. 

| | NUREG-1924, 2010. Electric Raceway 

Fire Barrier Systems in U.S. Nuclear. 

| | Rosner, D., Grote, B., Hartman, K, 

Hofling, B, Guericke, O., 1998. From 

natural language documents to 

sharable product knowledge: a 

knowledge engineering approach. In: 

Borghoff U.M., Pareschi, R. (Eds.), 

Information technology for knowledge 

management, pp. 35–51, Springer 

Verlag. 

| | Society of Fire Protection Engineers, 

2003. SFPE Hand Book. 

| | TPC Maanshan Nuclear Power Plant, 

1999. Final Safety Analyze Report. 

| | TRIZ: A New Approach to Innovative 

Engineering and Problem Solving, 1996 

AME Annual Conference in Milwaukee, 

WI, November 5-8. 

| | USNRC Generic Letter 77-02, 1977. Fire 

Protection Functional Responsibilities, 

Administrative Control and Quality 

Assurance. 

Authors 

Chia-Nan Wang 

Hsin-Po Chen 

Fong-Li Chin 

Ming-Hsien Hsueh 

Department of Industrial 

Engineering and Management 

National Kaohsiung University 

of Applied Sciences 

No.415, Jiangong Rd., Sanmin Dist., 

Kaohsiung City 807 

Taiwan, China 

Department of Industrial 

Engineering and Management 

National Kaohsiung University 

of Applied Sciences 

No.415, Jiangong Rd., Sanmin Dist., 

Kaohsiung City 807 

Taiwan, China 





104 

DECOMMISSIONING AND WASTE MANAGEMENT 


in Salt Solutions 

Bernhard Kienzler 

Introduction For many years, in Germany POLLUX canisters were considered as reference concept for spent 

nuclear fuel disposal casks. The cask consists of the shielding cask with a screwed-in lid and the inner cask with bolted 

primary and welded secondary lid. The spent fuel should be inserted in the final disposal cask in bins. The cylindrical 

wall and bottom of the inner cask consist of fine-grained steel 15 MnNi 6.3. The thickness of the cylindrical wall was 

designed according to the mechanical and shielding requirements and was 160 mm thick. The primary lid of the inner 

cask was also made of fine-grained steel. This lid was designed to keep the sealing function prior to and during the 

welding of the secondary lid. A plate made of neutron-moderating and absorbing materials (carbon/boron mixture) 

was attached to the primary lid. The secondary lid is designed as a welded lid. The base body of the shielding cask 

consisted of ductile cast iron (GGG 40). The wall thickness was designed according to the requirements for the shielding 

and was 265 mm thick. The weight of the POLLUX cask was 65 Mg [1]. The whole POLLUX cask consisted of actively 

corroding steels. 

The corrosion behavior of the POLLUX 

materials in salt solution for temperatures 

up to 200°C were investigated 

[2]. Both materials showed high corrosion 

rates especially at elevated 

temperatures and frequently the question 

was asked why not using alloyed 

steels. In fact, alloyed steels are developed 

to be corrosion resistant, and the 

steels are widely used especially for 

corrosion-resistant applications. 

Alloyed steels such as stainless 

steels do not readily corrode, rust or 

stain in contact with water as finegrained 

or cast iron steels. However, 

the alloyed steels are not fully stainproof 

in low-oxygen or high-salinity 

environments. There are various 

grades and surface finishes of stainless 

steel to suit the environment the 

alloy must endure. Stainless steel is 

used where both the properties of 

steel and corrosion resistance are 

required. 

Stainless steels differ from carbon 

steel by the amount of chromium 

present. Unprotected carbon steel 

rusts when exposed to air and 

moisture. The iron oxide film has 

lower density than steel, the film 

expands and tends to flake and fall 

away. In comparison, stainless steels 

contain sufficient chromium to 

undergo passivation, forming an inert 

film of chromium oxide on the surface. 

This layer prevents further corrosion 

by blocking oxygen diffusion to the 

steel surface and stops corrosion from 

spreading into the bulk of the metal. 

Passivation occurs only if the proportion 

of chromium is high enough 

and oxygen is present. 

In the scope of corrosion studies 

of high-level waste canister materials, 

the corrosion behavior of several 

alloyed materials was investigated. 

The materials comprised nickel based 

alloys (Hastelloy C22 and C4), and 

chromium-nickel steels. Furthermore, 

titanium alloys and copper-nickel 

alloys were taken into the investigations. 

These alloys are not covered in 

this contribution. 

The recommendations of the 

German High-Level Waste Commission 

[3] are reflected in the German law for 

amendment of the site selection law 

(passed by the German Parliament, 

March 23, 2017 [4]). Especially the 

maximum temperature condition has 

been changed. The maximum temperature 

at the canister surfaces is now 

limited to 100 °C, and the retrievability 

of the wastes during the operational 

phase of the disposal and the recoverability 

of the wastes for a period of 500 

years is need to be taken into account. 

Corrosion mechanisms 

of alloyed steels 

The corrosion resistance of stainless 

steel (Cr-Ni steel) known under 

atmospheric conditions depends on 

the chromium content of the alloy. 

Chromium leads to the formation of a 

passive layer, the so-called chromium 

oxide skin, which spontaneously 

forms in air and protects the material 

underneath from corrosion. By 

alloying different chromium and 

molybdenum fractions, the corrosion 

resistance can be adjusted to the 

environmental conditions. The low 

corrosion rates of Cr-Ni steels are due 

to the build-up of passive layers (oxide 

layers) on the surface, which are 

re-established under the conditions of 

low-concentrated solutions. 

The stability of container materials 

in a deep underground disposal is 

influenced by various uniform and 

local corrosion processes. These 

processes are controlled by the local 

geochemical conditions, in particular 

pH, redox potential and chloride 

concentration. Iron and steels are not 

thermodynamically stable in contact 

with water or saline solution. A 

number of different corrosion processes 

are described depending on a 

variety of factors [5]. For metals, two 

types of corrosion occur: general and 

localized corrosion. 

• General or uniform corrosion 

results in a relatively uniform mass 

loss over the entire area of the 

sample. General corrosion effects 

are predictable. Cast irons and 

steels corrode uniformly when 

exposed to open atmospheres, soils 

and natural waters as well as in salt 

solutions. 

• Localized corrosion occurs at discrete 

sites on the metal surface. 

The areas immediately adjacent to 

the localized corrosion normally 

corrode to a much lesser extent. 

These types of corrosion are less 

common in atmospheric exposure 

than in immersion exposures. 

Corrosion activity at localized 

corrosion sites may vary with 

changes of the water composition, 

defects in passivation layers, 

changes in contaminants or 

pollutants, changes in the electrolyte 

and by formation of 

galvanic cells. The predominant 

forms of localized corrosion are 

pitting and crevice corrosion. 

• Pitting corrosion is especially 

prevalent in metals that form a 

protective oxide layer. Pitting 

can be initiated on an open, 

freely-exposed surface or at 

imperfections in the passivation 

layer. Deep, even fully penetrating 

pits can develop with 


Corrosion Processes of Alloyed Steels in Salt Solutions ı Bernhard Kienzler


only a relatively small amount 

of metal loss. Pitting can occur 

isolated or as group of pits 

which may coalesce to form a 

large area of damage. 

• Crevice corrosion occurs in 

crevices where the environment 

differs from the surrounding 

bulk environment. The different 

environments result in 

corrosion because of differences 

in concentration (e.g., 

oxygen, pH, and ferric ions). If 

there is an oxygen concentration 

difference, corrosion will 

proceed at crevices where less 

oxygen is available than in the 

environment surrounding the 

crevice. Crevices are formed 

when two surfaces are in 

proximity to one another, such 

as when two metal surfaces are 

in close contact. 

• Contact (galvanic) corrosion 

occur when different metals are 

in contact in a common electrolyte. 

At current flows between 

the two metals, the less noble 

metal (the anode) corrodes at a 

faster rate than would have 

occurred if the metals were not 

in contact. In this case, the rate 

of corrosion depends on the 

relative areas of the metals in 

contact and the composition 

(conductivity) of the electrolyte. 

• Stress corrosion cracking 

(SCC) requires the simultaneous 

presence of tensile 

stresses (effect of external loads 

or welding / bending) and 

specific environmental factors. 

• Intergranular attack is caused 

by carbon diffusion to the grain 

boundaries and precipitation as 

chromium carbide. This effect 

removes chromium from the 

metal phase (solid solution) 

leaving a lower chromium 

content adjacent to the grain 

boundaries. 

Especially in environments with high 

chloride concentrations, chloride 

promotes the breakdown of the oxide 

layer. In the presence of chloride ions, 

oxygen can be displaced by chloride 

ions in the oxide layer of the passivated 

metal. The addition of further 

chloride ions results in a region which 

is no longer protected by the oxide 

layer. This site now offers an attack 

point for further corrosion. Under 

favorable circumstances, a so-called 

re-passivation may occur: the chloride 

ion is displaced again by oxygen, and 

the protective oxide layer is “repaired” 

again. Otherwise, the pitting corrosion 

continues. The rate of displacement 

of oxygen by chloride in the 

passivation layer is the measure of the 

incubation period for the occurrence 

of local corrosion processes. The 

following mechanisms effect pitting 

corrosion [6]: 

• The dissolved oxygen concentration 

outside of the pit is considerably 

higher than in the hole. The 

low oxygen concentration in the 

pit hinders re-passivation of the 

metal. 

• The small pit forms an anode, the 

remaining surface represents the 

cathode. The corrosion rate is 

determined by the ratio of the 

cathode to anode area. 

• The metal dissolves according 

Me n+ + H 2 O + k MeOH (n-1)+ + 

H + , reducing the pH. 

• Critical potential must exceed a 

certain critical potential value. In 

salt solution, the critical potential 

is defined by E pit = A + B log [Cl − ] 

with Cl − is the bulk chloride 

concentration. B is generally in 

the range 60-90 mV [7]. Critical 

pitting potentials (E pit ) of 1.4301 

Cr-Ni steel (type 304, UNS S30400) 

are reported by Yashiro et al [8] as 

a function of temperature (373 K 

to 523 K) and chloride (Cl − ) 

concentration (0.01 to 2 mol/kg- 

H 2 O). Steady polarization tests 

were performed at discrete intervals 

around Epit. Results were 

expressed by E pit = A − B log [Cl − ]. 

In regard to temperature dependency, 

the constant A decreased 

with temperature, while B was 

almost constant up to 448 K. 

• In the presence of Cl − , the dissolved 

metal in the pit reacts with 

chloride forming iron chlorides 

which hydrolyses (FeCl 2 +H 2 O vk 

FeClOH + Cl − + H + ) and reduce 

the pH. 

The actual water consumption for 

pitting corrosion is substantially lower 

than in the case of uniform surface 

corrosion of unalloyed steels. Carbon 

steels also shows a passivation in the 

alkaline environment, e.g. at pH > 12 

in concrete constructions [9]. 

In contrast to alloyed steels, 

unalloyed carbon steels do not build 

up a protective layer under low or 

slightly basic pH conditions, since 

the alloying element chromium is 

missing. Under acidic to basic pH 

conditions voluminous iron oxides / 

iron hydroxides are formed, which 

generally do not adhere to the underlying 

material. Therefore, the steel is 

not protected but the oxidation is 

maintained under the influence of 

moisture and oxygen. This reaction 

observed in the unalloyed steels is 

referred to as an active corrosion 

process in which iron reacts to iron 

oxide/hydroxide. Numerous experiments 

have shown that the active 

corrosion of the unalloyed steels is 

uniform and at a largely constant rate 

[10–16]. This behavior allows predicting 

the mass loss or thickness 

reduction of the disposal cask to a 

certain degree. 

The corrosion experiments reported 

here were performed in salt 

solutions. Under reducing conditions 

as they prevail in a deep geological 

disposal, the corrosion process of 

carbon steel consumes water and 

generates hydrogen. During the corrosion 

process, dissolved iron reacts 

with the aqueous medium forming 

ferrous hydroxides with divalent iron 

(Fe II ). At 7 < pH < 9, the observed 

solid corrosion products are magnetite 

(Fe 3 O 4 ) and amorphous iron 

hydroxides. At sufficiently low redox 

potentials (absence of oxygen) in 

chloride solutions, Cl − ions react with 

amorphous iron hydroxides forming 

the reaction product “green rusts”. 

This compound has the formula 

[Fe II 3Fe III (OH) 8 ]Cl×H 2 O and can be 

formed at [Cl − ]/[OH − ] > 1 [17]. It 

consists of both Fe II and trivalent iron 

(Fe III ). In contact with oxygen, green 

rust transforms quickly to magnetite. 

In the presence of Mg-rich brines, 

(Fe,Mg)(OH) 2 and Fe(OH) 2 Cl compounds 

were found and characterized 

[18]. 

Materials and methods 

When the corrosion experiments were 

started, the boundary conditions for 

the research on container materials 

for highly radioactive waste resulted 

from the requirements defined by 

pouring the molten highly radioactive 

glass directly into the canister, apply 

the necessary welding and decontamination 

of the containers and by 

the requirement for transport, interim 

storage and final disposal. For the 

POLLUX canister, the influence of the 

production and sealing of a final 

storage canister was considered, and 

U-shaped samples, welded samples 

using different welding procedures, as 

well as contact samples were prepared 

for the experiments. In particular, to 

assess the influence of the welding on 

the corrosion processes, different 

treatments of the samples were 

applied, including the delivery state, 

heat-treated samples, welded and 

subsequently heat-treated samples. 

DECOMMISSIONING AND WASTE MANAGEMENT 105 





Material 

Material 

description 

A comprehensive description of the 

sample shape and treatment has been 

published [19]. Further experiments 

included contact samples where 

different steels were screwed together 

in close contact and corrosion tests 

under γ irradiation. The whole suite of 

steels under investigations ale listed in 

Table 1. 

Two different sample types were 

produced to test the materials for 

mass loss, pitting corrosion, crack 

corrosion and stress corrosion 

Material 

number 

Density 

g/cm 3 

Ni based alloys Hastelloy C4 Ni Mo 16 Cr 16 Ti 2.4610 8.669 

Ti alloys Titan – Palladium Ti 99.7 – Pd 

Ti 99.7 - Pd EG 

Fe based alloys Fine-grained steel FStE 255 

TStE 460 

15 Mn Ni 6.3 

DC 01 / St 12 

ST 37-2 

Cr-Ni steel 

Cu alloys 

Nodular cast steel 

Ni-Resist D2 

Ni-Resist D4 

Nirosta 

GGG 40.3 

GGG-Ni Cr 20.2 

GGG-Ni Si Cr 30.55 

X2CrNi19-11 

Cu.99 

Cu-Ni 70/30 

Cu-Ni 90/10 

Ni alloys Nickel 99.9 

Ni/Cu 70/30 

| | Tab. 1. 

Metal alloys for construction of waste canisters under investigation at KIT-INE. 

3.7025 

3.7035 

1.0566 

1.8915 

1.6210 

1.0330 

1.0038 

0.7043 

0.7660 

0.7680 

1.4833 

1.4306 

4.0000 

4.7000 

4.9000 

2.4068 

2.4360 

4.593 

4.593 

7.814 

7.671 

7.512 

7.85 

7.856 

6.955 

7.36 

7.596 

8.022 

7.956 

9.198 

8.866 

8.998 

8.48 

8.51 

cracking (SCC). For the determination 

of the mass loss, sheet metal specimens 

with the dimensions 40 mm × 

20 mm were cut in the respectively 

available sheet thicknesses. The mass 

loss was determined only in the case 

of samples in the delivery condition. 

The susceptibility to pitting corrosion 

as well as the susceptibility to crack 

corrosion could be assessed also. 

The Ni-Resist steels have been 

included in the investigation program 

because these steels are specified for 

handling salt solutions such as sea 

water. Lower uniform corrosion rates 

were expected as in the case of fine 

grained steel. After the exposure time, 

the samples were recovered from the 

corrosion medium and the specimens 

were cleaned from the adhering salts 

and corrosion products by pickling in 

suitable solutions according to ASTM 

guidelines [20]. Then the specimens 

were cleaned in alcohol and examined 

for general and local corrosions as 

well as for stress corrosion cracking. 

The general corrosion (integral corrosion 

rate) was calculated from the 

integral weight losses determined by 

gravimetry and from the respective 

material densities. The specimens 

were examined for local corrosion and 

stress corrosion cracking by microscopic 

evaluation and with the help of 

metallographic cross-sections, measurements 

of pit depths and surface 

profiles. 

Results and discussion 

General corrosion 

Due to the fact that localized corrosion 

processes are observed in the 

experiments, the mass loss rate is used 

for comparisons. The general corrosion 

rate relies on uniform corrosion 

of the surfaces and is not considered 

reasonably for alloyed steel. Figure 1 

and Figure 2 show the mass loss and 

the corresponding mass loss rates for 

a) mass loss 

b) mass loss rate 

| | Fig. 1. 

Measured mass loss and mass loss rates of Hastelloy in MgCl 2 -rich (red) and NaCl solution (blue) as function of time at various temperatures. 

a) mass loss 


| | Fig. 2. 

Measured mass loss and mass loss rates of Cr-Ni steels (1.4306 and 1.4388) in MgCl 2 -rich (red) and NaCl solution (blue) as function of time at 150 °C. 



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a) mass loss 


| | Fig. 3. 

Measured mass loss and mass loss rates of fine-grained steel (1.6210) in MgCl 2 -rich (red) and NaCl solution (blue) as function of time at 150 °C. 

Hastelloy and for the two Cr-Ni steels. 

The Hastelloy experiments covered a 

temperature range between 90 °C and 

170 °C, whereas the CR-NI steels were 

investigated at 150 °C, only. 

For Hastelloy, all mass loss rates 

were found below 12 g m -2 yr. -1 showing 

no distinct time dependence. For 

the experiments with Cr-Ni steels, the 

initial mass loss rates decreased and 

remained for the long term below 

15 g m -2 yr. -1 . The effect of the solution 

type on the mass loss rates for Cr-Ni 

steels was not significant. Also, the 

differences of the mass loss and mass 

loss rates between 1.4306 and 1.4833 

steels were marginal. Concerning the 

temperature effect of the general 

corrosion of Hastelloy, relatively high 

mass losses were found at 90°C after 

676 days. At higher temperatures, the 

exposure period remained below 500 

days. The reason for the increased 

mass losses could be explained by 

crevice corrosion of the Hastelloy C22 

in MgCl 2 rich solution showing pit 

depths of about 200 µm. The scatter 

of mass losses is correlated to local 

corrosion processes. 

For comparison, the mass loss and 

mass loss rates of the fine-grained 

steel 1.6210 is shown in Figure 3. In 

this case, the mass loss rates were by a 

factor of 50 higher in comparison to 

the Cr-Ni steel in NaCl solutions and 

by a factor about 100 higher in MgCl 2 

solution after about 500 days (150 °C). 

The uniform mass loss rates of 

the Ni-Resist steels were found in 

the range of the Cr-Ni steels at 

20 ± 7 g m 2 yr. 1 for steel 0.7660 

and 12 ± 9 g m -2 yr. -1 for 0.7680, 

respectively. These values are also 

significantly lower in comparison to 

the fine-grained steel 1.6210. 

| | Fig. 4. 

Crevice corrosion in Hastelloy C22 after 676 

days in MgCl 2 rich solution at 90 °C showing 

depths of about 200 µm. 

Local corrosion phenomena 

The breakdown of passivity (the 

breaching of the protective barrier 

provided by the passive film) initiates 

the most damaging kinds of corrosion, 

the localized forms of corrosion, 

pitting, crevice corrosion, intergranular 

attack, and stress corrosion. The 

induction period for pitting corrosion 

starts with the initiation of the breakdown 

process by the introduction of 

breakdown conditions and ends when 

the localized corrosion density begins 

to rise. Unfortunately, electrochemical 

corrosion studies were applied 

only for carbon steel and the influence 

of chemical species in brines have 

been investigated [21]. For this 

reason, corrosion potential for pitting 

corrosion have not been determined 

for the investigated alloyed steels. 

In brine media, localized corrosion 

has been investigated over the complete 

range of chloride concentrations. 

The Cl- concentration, however, 

is not as critical as pH and temperature, 

since the attack can occur at any 

concentration over the minimum 

value. Factors such as incubation 

time, severity, and frequency of 

occurrence can be influenced by the 

concentration. 

Localized corrosion was observed 

for all alloyed steels. In the case of 

Hastelloy C22, the first pits occurred 

after 275 days in the MgCl 2 rich 

solution at 90 °C. These pits had 

depths of about 10 µm. After 552 

days, the depths increased to 20 µm 

and after 676 days, a pit’s depth of 

200 µm was found in a crevice. In the 

MgCl 2 solution 2, even deeper pits 

were detected. In NaCl solution, after 

552 days, the pit’s depth amounted to 

16 µm. 

The average pit depths as function 

of time in the steels 1.4306 and 1.4833 

are shown in Figure 5. 

In contrast to the observations 

for Hastelloy, the depths of the pits 

were significantly deeper after about 

3 months. The pits showed relative 

a) Steel 1.4306 at 150°C 

| | Fig. 5. 

Average pit depths determined in untreated Cr-Ni steels as function of time. 

b) Steel 1.4833 at 150°C 




a) Stress corrosion cracking in the heat 

affected zone of a welding seam: 

TSS Experiment at Asse salt mine 

temperature: 180 °C 

Duration: about 11 years. 

| | Fig. 6. 

Localized corrosion phenomena of steel 1.4306: Stress corrosion cracking along grain boundaries. 

high depth variations. The immersion 

tests were terminated after about 500 

days, therefore an increase in the pit’s 

depths as determined in the case of 

Hastelloy was not observed. The 

average pit depth of both steel was 

found in the range of 30 to 40 µm. 

The steels 1.4306 and 1.4833 

showed significant stress corrosion 

cracking at 150 °C (tests at 90 °C were 

not performed). Figure 6 shows polished 

micrographs of steel 1.4306 

specimen in contact with dry rock salt 

(a) and immersed in NaCl solution. 

Localized corrosion was found in 

both cases, even in the almost dry 

con ditions established in the TSS 

experiment performed in the Asse salt 

mine [22]. The penetration depths 

of the cracks were measured in 

the mm range. Contact samples in 

MgCl 2 solution showed even more 

pronounced stress corrosion cracking 

[2]. 

With Hastelloy C4 corrosion tests 

under γ-irradiation of 10 Gy/h were 

performed (fuel element storage 

pool at Dido test reactor at the 

Research Center Juelich). Different 

types of samples were examined: 

plane samples as delivered, plane 

samples with removed oxide layer 

on the surface, U-shaped welded 

samples, crevice samples, and samples 

| | Fig. 7. 

Intergranular corrosion in a Ni-Resist D4 

sample after 776 days in MgCl 2 solution 

at 90 °C. 

b) Stress corrosion cracking of a plane 

specimen of steel 1.4306 after 422 days 

in NaCl solution at 150°C. 

with different welding procedures 

such as tungsten inert gas welding 

(TIG) or electron beam welding (EB). 

Significant deviation of the observed 

mass losses in comparison to test without 

irradiation were not found. 

Almost all Ni-Resist steel samples 

showed intergranular corrosion effects 

(Figure 7). These referred to samples 

as delivered and to crevice samples. 

Summary and conclusions 

The results of the corrosion experiments 

with Cr-Ni steels, Hastelloy and 

the Ni-Resist materials revealed a 

significantly lower general corrosion 

rate (mass loss rate) in comparison to 

the fine-grained steels. On the other 

hand, these materials were subdued 

to localized corrosion processes such 

as pitting corrosion, crevice corrosion, 

intergranular corrosion and stress 

corrosion cracking. The local corrosion 

processes were enhanced in 

welded or in contact specimen. In 

many cases, the localized corrosion 

phenomena were found only after 

certain incubation periods. Especially 

in the case of Hastelloy, the incubation 

period was about 9 months at 90 °C in 

MgCl 2 solution and the pitting corrosion 

rate was relatively high. Stress 

corrosion cracking by intergranular 

corrosion of the Cr-Ni steels penetrated 

deep into the materials. Intergranular 

corrosion was also found in 

the Ni-Resist steels. 

As a consequence of the occurrence 

of localized corrosion processes 

as well as the unpredictable incubation 

times of these processes, one 

might understand the decision to 

apply uniformly corroding steels for 

waste canisters, even if the general 

corrosion rate would be by a factor up 

to 1,000 higher. 

The mass loss is proportional to the 

hydrogen produced under reducing 

conditions in a deep disposal. A 

POLLUX cask has a surface area of 

about 30 m 2 . Under extreme conditions, 

15 kg of steel could be corroded 

per year in NaCl solution, forming 

360 mol H 2 per year (8 m 3 standard 

conditions). Hydrogen keeps a reducing 

environment, however, by 

increasing pressure it acts as driving 

force for gas, solution and contaminant 

transport. Internationally efforts 

are undertaken to reduce the potential 

amount of hydrogen produced by 

corrosion phenomena. 

Based on the measurements 

reported in this contribution, Cr-Ni 

steels seem not to provide a reasonable 

solution for a long-lived stable 

waste package. Even, if the hydrogen 

production is reduced, the long-term 

sealing function of these steels is 

unclear. Under the almost dry condition 

of the in-situ experiment (TSS) 

in the Asse mine, stress corrosion 

cracking in the heat affected zone of 

a welding seam of Cr-Ni steel was 

observed after 11 years at 180 °C. 

Acknowledgment 

The corrosion studies of canister 

materials for heat producing wastes 

cover exclusively the research performed 

by Dr. Emmanuel Smailos and 

his working group. Until his retirement 

in 2004, Dr. Smailos was responsible 

for the corrosion studies of 

various materials at the Institute for 

Nuclear Waste Disposal (INE). 

References 

[1] H. Lahr, H.-O. Willax, and H. Spilker, 

Conditioning of spent fuel for interim 

and final storagein the pilote conditioning 

plant (PKA) at Gorleben, in 

International Symposium on Storage of 

Spent Fuel from Power Reactors, 

Vienna, Austria, 9-13 November 1998, 

1998. 

[2] E. Smailos and B. Fiehn, Korrosionsuntersuchungen 

an der Werkstoffkombination 

des POLLUX-Behaelters 

zur direkten Endlagerung abgebrannter 

Brennelemente in Steinsalz formationen, 

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4552, 1989. 

[3] Kommission Lagerung hoch radioaktiver 

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[4] Gesetz zur Fortentwicklung des 

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Standortes für ein Endlager für Wärme 

entwickelnde radioaktive Abfälle und 

anderer Gesetze, 2017. 

[5] Uhligs corrosion handbook, 3 rd ed 

( Online-Ausg.) ed. Hoboken, N.J: Wiley, 

2011. 






[6] R. Newman, Pitting Corrosion of Metals, 

Electrochem. Soc. Interface Vol. 19, 

pp. 33-38, 2010 

[7] J. R. Galvele, Transport processes and 

the mechanism of pitting of metals, 

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pp. 464-474 1976. 

[8] H. Yashiro, K. Tanno, S. Koshiyama, and 

K. Akashi, Critical Pitting Potentials for 

Type 304 Stainless Steel in High- 

Temperature Chloride Solutions 

Corrosion, Vol. 52, pp. 109-114, 1996. 

[9] George R. Brubaker and P. B. P. Phipps, 

Corrosion chemistry, Washington, D.C.: 

American Chemical Society, 1979. 

[10] E. Smailos, W. .Schwarzkopf, R. Köster, 

and K. H. Gruenthaler, Advanced 

corrosion studies on selected packaging 

materials for disposal of HLW canisters 

in rock salt, in Corrosion Problems 

Related to Nuclear Waste Disposal: 

A Working Party Report, European 

Federation of Corrosion, Ed., ed: The 

Institute of Materials, 1992, pp. 23-31. 

[11] E. Smailos, W. Schwarzkopf , B. Kienzler, 

and K. R., Corrosion of Carbon-Steel 

Containers for Heat-Generating Nuclear 

Waste in Brine Environments Relevant 

for a Rock-Salt Repository, in Scientific 

Basis for Nuclear Waste Management: 

Proc.of the 15th Internat.Symp., 

Strasbourg, November 4-7, 1991, 1992, 

pp. 399-406. 

[12] E. Smailos, Corrosion of high-level 

waste packaging materials in disposal 

relevant brines, Nuclear Technology, 

Vol. 104, pp. 343-350, 1993. 

[13] E. Smailos, I. Azkarate, J. A. Gago, 

P. van Iseghem, B. Kursten, and 

T. McMenamin, Corrosion on metallic 

HLW container materials, in Fourth 

European Conference on Management 

and Disposal of Radioactive Waste, 

1997, pp. 209-223. 

[14] E. Smailos, A. Martínez-Esparza, 

B. Kursten, G. Marx, and I. Azkarate., 

Corrosion evaluation of metallic 

materials for long-lived HLW/spent 

fuel disposal containers, Forschungszentrum 

Karlsruhe, FZKA 6285, 1999. 

[15] E. Smailos, M. A. Cunado, I. Azkarate, 

B. Kursten, and G. Marx, Long-term 

performance of candidate materials for 

HLW/spent fuel disposal containers, 

Forschungszentrum Karlsruhe, Wissenschaftliche 

Berichte, FZKA-6809, 2003. 

[16] E. Smailos, Influence of gamma 

radiation on the corrosion of carbon 

steel, heat-generating nuclear waste 

packaging in salt brines, IAEA, Wien 

IAEA TECDOC-1316 Effects of Radiaton 

and Environmental Factors on the 

Durability of Materials in Spent Fuel 

Storage and Disposal, 1995. 

[17] A. Raharinaivo, G. Arliguie, 

T. Chaussadent, G. Grimaldi, V. Pollet, 

and G. Taché, La corrosion et la 

protection des aciers dans le béton, 

Paris: Presses de l'École Nationale des 

Ponts et Chaussées, 1998. 

[18] B. Grambow, E. Smailos, H. Geckeis, 

R. Müller, and H. Hentschel, Sorption 

and reduction of uranium(VI) on iron 

corrosion products under reducing saline 

conditions, Radiochimica Acta, Vol. 

74, pp. 149-154, 1996. 

[19] E. Smailos, R. Köster, and 

W. Schwarzkopf, Korrosionsuntersuchungen 

an Verpackungsmaterialien 

für Hochaktive Abfälle, European Appl. 

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pp. 175-222, 1983. 

[20] ASTM G 1- 72, Recommended Practice 

for Preparing, Cleaning and Evaluation 

of Corrosion Test Specimens, Annual 

Book of ASTM Standards, Vol. Part 10, 

p. 489, 1974. 

[21] A. M. Farvaque-Bera and E. Smailos, 

Electrochemical Corrosion Studies on a 

Seleted Carbon Steel for Application in 

Nuclear Waste Disposal Containers: 

Influence of Chemical Species in Brines 

on Corrosion, Kernforschungszentrum 

Karlsruhe, KfK-5354, 1994. 

[22] W. Bechthold, E. Smailos, 

S. Heusermann, W. Bollingfehr, 

B. B. Sabet, T. Rothfuchs, P. Kamlot, 

J. G. Olivella, and F. D. Hansen, 

Back filling and sealing of underground 

repositories for radioactive waste in 

salt (Bambus II project). Final report, 

European Commission, EUR-20621-EN, 

2004. 

Author 

Dr. Bernhard Kienzler 

Karlsruhe Institute of 

Technology (KIT) 

Institut für Nukleare 

Entsorgung (INE) 

Hermann-von-Helmholtz Platz 1 

76344 Eggenstein-Leopoldshafen 

Germany 

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Design and Development of a Radioeco 

logical Domestic User Friendly Code 

for Calculation of Radiation Doses and 

Concentration due to Airborn Radionuclides 

Release During the Accidental 

and Normal Operation in Nuclear 

Installations 

A. Haghighi Shad, D. Masti, M. Athari Allaf, K. Sepanloo, S.A.H. Feghhi and R. Khodadadi 

1.1 Introduction Though nuclear power is a good source of energy and is not generally a threat, a major reactor 

accident can lead to a catastrophe for people and the environment. The major health and environmental threat would 

be due to the escape of the fission products into the atmosphere. There have been instances of nuclear reactor accidents 

like the heavy water cooled and moderated reactor at Chalk River in Canada in 1952, the graphite moderated gas cooled 

reactor at Sellafield in Britain in 1957, the boiling water reactor at Idaho Falls in US in 1961, the pressurized water 

reactor on Three Mile Island in the US in 1979, the graphite moderated water cooled reactor at Chernobyl in Ukraine in 

1986, the sodium cooled fast breeder reactor at Monju in Japan in 1995 [Makhijani, 1996] and the boiling water reactor 

at Fukushima Daiichi NPP in Japan following an earthquake and tsunami in 2011. Among them, Chernobyl and 

Fukushima completely changed the human perception of radiation risk. On April 26, 1986, USSR suffered a major 

accident, which was followed by an extensive release to the atmosphere of large quantities of radioactive materials. An 

explosion and fire released huge quantities of radioactive particles into the atmosphere, which spread over much of the 

western USSR and Europe. The Chernobyl disaster was one of the two maximum classified event (level 7) on the 

International Nuclear Event Scale (the other being the Fukushima Daiichi nuclear disaster happened in 2011) and was 

the worst nuclear power plant accident in history in terms of cost and the resulting deaths. The battle to contain the 

contamination and avert a greater catastrophe ultimately involved over 500,000 workers and cost an estimated 

18 billion rubles. During the accident itself, 31 people died, and long-term effects such as cancers and deformities are 

still being accounted for. Unfortunately, the other severe accident happened on March 11, 2011; a powerful earthquake 

(magnitude 9.0) hit off the east coast of Japan. The tsunami triggered by the earthquake surged over the east coast of 

the Tohoku region, including Fukushima. The Fukushima Daiichi NPP’s cooling ability was lost and reactors were heavily 

damaged. Owing to controlled venting and an unexpected hydrogen explosion, a large amount of radioactive material 

was released into the environment. Consequently, many residents living around the NPP were exposed to radiation. 

In almost every respect, the consequences of the Chernobyl accident clearly exceeded those of the Fukushima accident. 

In both accidents, most of the radioactivity released was due to volatile radionuclides (noble gases, iodine, caesium, and 

tellurium) [G. Steinhauser, A. Brandl, T. E. Johnson, 2014]. 

111 

RESEARCH AND INNOVATION 

1.2 The context 

The objective of the paper is to develop 

a domestic user friendly dynamic 

radio logical dose and model for accidental 

atmospheric release of radionuclides 

and normal operation from a 

nuclear facility, which has been coupled 

with a long-range atmospheric 

transport and Gaussian dispersion 

model. The research in this study is 

based on (i) atmospheric dispersion of 

radionuclides, (ii) dose and risk model 

development, (iii) validation of the 

model with FSAR of typically 

WWER-1000 Reactor. Models to 

represent the transport of radionuclides 

following atmospheric tests 

of nuclear weapons were developed 

during the 1950s and 1960s. Though 

radio nuclides have been released into 

the environment during routine operational 

conditions of nuclear facilities, 

accidents and nuclear weapons tests, 

the KIANA Advance Computational 

Computer Code model that was developed 

for this study was planned to 

predict all of radiation doses and risks 

in the case of a nuclear accident and 

normal operation in nuclear installations. 

The novelties in this research are 

to couple a KIANA Advance Computational 

Computer Code dynamic dose 

and risk model with a long-range 

atmospheric transport model to predict 

the radiological consequences due 

to accidental releases and normal 

operation in nuclear installations, and 

to perform the model simulation for 

NPP sites in IRAN territory and with 

another site specification data as far as 

it can be acquired. Most of the mechanisms 

and phenomena considered in 

each of the existing dose and risk 

calculation and environmental transfer 

models have been compiled in the 

newly developed single 

KIANA Advance Computational 

Computer Code to lead detailed modelling. 

An uncertainty and sensitivity 

analysis can also part of the study to 

determine the most influential parameters 

and their uncertainties on the 

results for users (if applicable). A huge 

amount of data, such as radioactivity 

concentration in food, pasture and 

doses, regarding the consequences 

of nuclear power plants’ accidents 

and normal conditions in literature 

was used for the development of 

Computer Code and its validation. 


Design and Development of a Radio eco logical Domestic User Friendly Code for Calculation of Radiation Doses and Concentration due to Airborn Radio nuclides Release 

A. Haghighi Shad, D. Masti, M. Athari Allaf, K. Sepanloo, S.A.H. Feghhi and R. Khodadadi


RESEARCH AND INNOVATION 112 

1.3 The innovation 

The main features of this software and 

study can be summarized as follows: 

Exposure from all pathways is 

included- Ingestion pathways are 

modelled in such a detailed way 

that, translocation, -transfer between 

soil-plant, and feed-animal, food processing 

and storage, weathering, and 

dilution in the plant are all taken into 

account. Time dependency in radionuclide 

transfer in the environment 

considering food harvesting, sowing 

times, feeding regimes, and the 

growing up of a person are all taken 

into account. Individual doses for 

maximum and average individuals 

and for four age groups are calculated. 

Doses in the case of implementation 

of countermeasures are calculated. 

Collective doses for big cities can be 

calculated. Two different methods for 

stochastic risk modelling are applied. 

A probabilistic module has also 

been developed; namely, uncertainty 

analysis can be performed (if applicable).This 

study is regarded as unique 

since. The model algorithms, which 

the KIANA Advance Computational 

Computer Code developed for this 

study was based on IAEA safety report 

series [Müller, H. and Pröhl, G., 1993], 

has been modified; the KIANA 

Advance Computational Computer 

Code to be able to calculate inhalation 

doses from resuspension, individual 

doses in terms of both average and 

maximum habits, collective doses and 

late risks, and to utilize the recent 

knowledge in the dose and risk assessment 

area to the extent possible, such 

as dose conversion factors and risk 

coefficients etc. 

The long-range transport model, 

which the code/software developed 

for this study was coupled with, 

was also upgraded to increase the 

number of pollutants modelled to 

provide us easiness. Besides, extensive 

uncertainty and sensitivity analyses 

associated with 96 parameters have 

been performed for this study. The 

meteorological module in the existing 

environmental emergency response 

system is associated with 3-day- 

Domestic forecast meteorological 

data acquired through the State 

Meteorological Directorate. The dispersion 

model is the Developed AIREM 

and DOZAE M model that has the 

capability to predict trajectories, 

concentration, and deposition patterns 

in the case of nuclear accidents and 

normal operations. However, doses, 

risks, and activities in the food chain 

are not calculated with the existing 

system in IRAN. Since the newly 

developed KIANA Advance Computational 

Computer Code for this 

study is compatible with the existing 

system's dispersion code, it can easily 

be integrated into it. 

2.1 Atmospheric dispersion 

models 

Numerous radiation dose calculation 

tools have been developed over the 

years. They calculate trajectories, 

atmospheric transport and dispersion, 

age-dependent radiation doses, early 

and late health risks, monetary costs 

of the accidents, doses in the case 

of implementation of emergency 

actions, collective health risk, uncertainty 

analysis etc. Atmospheric 

dispersion methods in these tools 

can be based on simple Gaussian or 

numerical approaches. Short-range 

dispersion models usually use 

straight-line Gaussian plume model. 

These models are appropriate if the 

release is from a source that has 

dimensions, which are small compared 

to the distances at which concentrations 

are to be estimated. For 

example, for the distances out to 

5-10 km from the source point, if the 

terrain is relatively flat and has 

uniform surface conditions in all 

directions and if the atmospheric 

conditions at the time and location of 

the release completely control the 

transport and diffusion of material 

in the atmosphere short-range 

atmospheric dispersion models are 

preferred. Gaussian dispersion equations 

should be used to estimate concentrations 

up to the 80 km from the 

source under ideal conditions of flat 

terrain and no spatial variations of the 

wind field. Consequently, for a countrywide 

dispersion simulation, due to 

topo graphy and dispersion area, the 

straight-line Gaussian models can not 

be appropriate tools. Therefore, longrange 

atmospheric dispersion models 

are used in this paper. Dose assessment 

methodology in some aforementioned 

short range codes neglects 

ingestion pathway and calculation 

of doses in the late phase of the accident. 

These are coupled with simple 

radiation dose modelling algorithm, 

including only inhalation and external 

radiation pathways i.e. HotSpot, 

RASCAL and RTARC [Homann, S. G., 

2010, Mcguire, S. A., Ramsdell, Jr., J. V. 

and Athey, G. F., 2007, Stubna M. and 

Kusovska Z. 1993] All radiation dose 

exposure pathways can be seen in 

Figure 1. 

Since short range codes generally 

calculate short-term doses incurred 

immediately after the accident and 

recommend emergency protective 

actions, such as intervention, sheltering 

and iodine pills, and long-term 

effects incurred from the ingestion 

pathway are not generally calculated 

with these types of codes. Some of 

the codes having a Gaussian plume 

methodology calculates ingestion 

doses, but not in a dynamic or 

| | Fig. 1. 

Radiation Dose Exposure Pathways in KIANA Advance Computational Computer Code. 





comprehensive way for real time 

releases i.e. GENII [Napier 2002]. 

Long- range atmospheric transport 

models, on the other hand, generally 

focus on the calculation of the trajectories, 

atmospheric transport and 

dispersion, and are used for real time 

emergency preparedness purposes. 

These numerical models use multiple 

wind measurements in both the horizontal 

and vertical directions, and 

include terrain effects and vertical 

and horizontal wind shear. They also 

treat the parameter variables more 

realistically, such as surface roughness, 

deposition and variable atmospheric 

stability. Numerical modelling 

is widely used to study long-range 

airborne transport and deposition of 

radioactive matter after a hypothetical 

accident and normal operations. 

Ladas, Mesos, and Derma are those 

having long-range atmospheric 

transport and dispersion algorithm 

[ Draxler, R.R., and G.D. Hess, 1997, 

Suh et al., 2006, 2008, 2009, Apsimon, 

H.M.; Goddard, A.J.H.; Wrigley, J., 

1985 and Sørensen, 1998; Sørensen et 

al., 2007]. Generally, these types of 

long-range dispersion codes are integrated 

with environmental transfer 

models to predict activity in the 

environment and the resulting doses. 

2.2 Radioecological models 

Two general classes of radioecological 

models have evolved; dynamic (transient) 

and equilibrium (steady state). 

Both describe the environment in 

terms of various „compartments” such 

as plant types, animal food products’ 

types and soil layers. Some environmental 

media may be described in 

terms of more than one compartment, 

such as the roots, branches and trunk. 

When the equations are evaluated for 

sufficiently long times with unvarying 

values of the inputs and rate constants, 

the ratios of the concentrations 

of the radionuclides in the various 

compartments approach constant 

values. The system is then considered 

to be in equilibrium or in a steady 

state. These „quasi-equilibrium models” 

do not account for changes in 

plant biomass, livestock feeding 

regimes, or in growth and differential 

uptake of radioactive progeny during 

food chain transport. They are generally 

not appropriate for the assessment 

of critical short-term impacts 

from acute fallout events that may 

occur during the different times of the 

year and for applications related to 

the development of criteria for the 

implementation of actions. In the late 

1970’s the dynamic radioecological 

models started to emerge and led to a 

number of different such models. 

Since dynamic food chain transport 

models themselves are normally 

rather complex and require significant 

computing times most of the codes 

[e.g. Slaper et al., 1994, Hermann et 

al., 1984, Napier et al., 1988] neglect 

radiation exposure changes due to 

seasonal variations of radionuclides 

in the environment and human 

behaviour. For more realistic dose 

calculations, time dependency of 

the radionuclide transfer processes 

should be taken into account, leading 

to a dynamic modelling. Lots of radiological 

data are necessary for dynamic 

ingestion pathway modelling. After 

the significant parameters are determined 

with respect to their effects on 

the results by sensitivity analysis 

these data may be derived locally to 

lead to realistic modelling, PARATI, 

PATWHWAY, Ecosys-87, SPADE 

(quasi- equilibrium), COMIDA and 

DYNACON are some dynamic dose 

models for modelling environmental 

transfer of radionuclides in the food 

chain [Rochedo et.al. 1996, Whicker 

and Kirchner, 1987, Müller, H., Pröhl, 

G., 1993, Johnson and Mitchell, 1993; 

Mitchell, 1999, Abbott, M.L., Rood, 

A.S., 1993, Hwang, W.T., Lee, G.C. Suh, 

K.S. E.H. Kim]. 

Since equilibrium in the model 

compartments (between vegetation, 

soil, and animal products) is not 

reached for a long time, it is essential 

to consider seasonality in the growing 

cycle of crops, feeding practices of 

domestic animals, and dietary habits. 

However, because of the temporal 

resolution demanded for the output, a 

great deal of information is required 

as input to this type of model, and 

extensive computer resources are 

required for the implementation. 

By using assumptions of quasiequilibrium 

(that is, relatively small 

changes from year to year in local 

conditions), the dynamic models may 

be simplified into equilibrium models. 

Knowledge of the contamination level 

of radionuclides in foodstuffs, including 

crops and animal products is 

essential information for deciding the 

implementation of protective actions. 

The degree of contamination can be 

evaluated through a model prediction 

from the amount of radionuclides 

deposited on the ground, as well as 

through direct measurements of 

radionuclides in foodstuffs. In developing 

systems for emergency preparedness 

as well as providing for 

rapid decision-making relating to 

foodstuffs, the characterization of 

action plans based on model predictions 

are likely to be appropriate. In 

the case of short-term deposition of 

radionuclides after a nuclear accident, 

the radionuclide concentration in 

foodstuffs is strongly dependent on 

the date (or season) when the deposition 

occurs, and on the time after the 

deposition due to factors such as 

crop growth and biokinetics of radionuclides 

ingested by the animals. 

Therefore, these dynamic environmental 

transfer models are generally 

implemented in a real time emergency 

or decision support systems, which 

are used before and during an ongoing 

emergency and provide sound 

basis countermeasures. In some radioecological 

models, such as COMIDA, 

CRLP and TERNIRBU [Brown, J. and 

Simmonds, J., R.,1995, KrcgewskiP., 

1989, Kanyar, B., Fulop N., TERNIRBU, 

1996] soil compartment is modelled 

in such a way that it is divided into 

many layers: surface layer, root layer, 

and deep soil layer, etc.. The code 

developed for this study took AIREM, 

DOZAE M & S. R.S of IAEA models as 

reference. The data library for unlimited 

isotopes is available in the new 

software (sub routines). All natural 

phenomena important for the ingestion 

pathway modelling is taken into 

consideration in the new algorithm 

and model. Whereas, time dependent 

translocation, layered soil compartment, 

wet interception, and mushroom 

pathway are not available in the 

current model. Generally, the computer 

models developed for the prediction 

of routine releases from NPPs 

are based on the annual average concentrations 

of radionuclides in air 

and on the ground. However, for NPP 

routine atmospheric releases a 

dynamic model coupled with a longrange 

transport code was developed 

in another study [Kocar, C., 2003]. In 

the current study, to address the 

unique features of modelling operational 

radiological consequences of 

nuclear power plants, a few new 

algorithm based on the dynamic 

radioecological model had been 

considered. Different from the aforementioned 

dynamic model [Müller, H. 

and Pröhl, G., 1993], transfer mechanisms 

of C-14 and H-3 were coded and 

multi-location food supply and interregional 

moves of people in the computational 

domain were permitted. 

In this study, inhalation doses from 

both passages of the cloud and resuspension 

of deposited activity are 

calculated and accidental releases are 

simulated, but the previous one is for 

operational releases are modelled and 







| | Fig. 2. 

Summary of Code Algorithms. 

H-3 and C-14 releases which are of 

great significance for operational 

releases are modelled. In this study, 

individual doses are calculated for 

two different habits of the people in 

term of food consumption and gamma 

reduction. 

3.1 KIANA advance 

computational computer 

code structure 

A deterministic dose calculation 

model called KIANA Advance Computational 

Computer Code has been 

developed for this study. For the dose 

assessment, all exposure pathways 

have been implemented as follows: 

Transfer of radionuclides through 

food chains and the subsequent 

internal exposures of humans due to 

ingestion of contaminated foodstuffs- 

Internal exposure due to inhalation of 

radionuclides during passage of cloud 

and from resuspension of deposited 

radionuclides- External exposure 

from radionuclides in the passing 

cloud- External exposure from radionuclides 

deposited on the ground. The 

design of the KIANA Code is flexible 

such that it can be adopted anywhere 

for any nuclear power plant/nuclear 

installation site with suitable modifications 

to the database. 

3.2 Ingestion pathway 

Ingestion pathway calculations in 

KIANA Advance Computational 

Computer Code take into account 

the following process and data: 

Yield of grass and agricultural food 

products. Harvesting and sowing time 

of grass and agricultural products. 

Translocation within plants. Interception. 

Weathering from plant surfaces. 

Dilution of radionuclide concentrations 

due to plant growth. Uptake 

by plant roots. Migration within the 

soil and Plant contamination due to 

resuspended soil. Different livestock 

feeding regimes. Storage times for 

fodder and human food products. 

Changes in radionuclide concentrations 

due to food processing. Age 

dependent ingestion dose coefficients 

for the public are taken from ICRP 72 

[1996]. Dose coefficients for 3 months 

infant, 5 year old children, 15 years 

old teen and adult are used. ICRP 

ingestion dose conversion factors take 

into account integration period of 

50 years for adults and 70 year for 

children. Input data to the ingestion 

modelling is the time integrated 

air concentrations, and deposited 

activity from any dispersion model or 

measured data. Ingestion of tap water 

and aquatic food products are not 

considered in KIANA Advance Computational 

Computer Code. 

3.3 Activity concentration 

of plant products 

The contamination of plant products 

as a function of time results from the 

direct contamination of the leaves and 

the activity transfer from the soil by 

root uptake and resuspension: 

C i (t) = C i,f (t) + C i,r (t) 

C i (t); total contamination 

of plant type i, 

C i,f (t); contamination of plant type i 

due to foliar uptake, 

Ci,r(t); contamination of plant type i 

due to root uptake 

Pasture and 13 different plant products, 

i.e. corn cobs, spring and winter 

wheat, spring and winter barley, rye, 

fruits, berries, and root, fruit and leafy 

vegetables, potatoes and beet can be 

modelled by KIANA Advance Computational 


3.4 Foliar uptake 

of radionuclides: 

Calculation of the contamination of 

plants must distinguish between 

plants that are used totally (leafy vegetables 

and grass) and plants of which 

only a special part is used. The activity 

concentration at time after the deposition 

is determined by the initial contamination 

of the plant and activity 

loss due to weathering effects (rain, 

wind) and radioactive decay and 

growth dilution. For plants that are 

totally consumed growth, excluding 

pasture grass, growth is implicitly 

considered because the activity deposited 

onto leaves is related to the 

yield at harvest. Interception factor is 

defined as the ratio of the activity initially 

retained by the standing vegetation 

immediately subsequent to the 

deposition event to the total activity 

deposited. Radionuclides to agricultural 

plants may be intercepted by dry 

process, wet process, or a combination 

of both. The interception fraction is 

dependent on the plant intensity in 

the area, stage of development of the 

plant, and generally leaf area of the 

crops. In the present model, a single 

coefficient was used and interception 

factors for grass and other plants were 

taken from DoseCAL code; the interception 

factor for grass and, fruits and 

vegetables is assumed to be 0.3 and 

for the grain and cereals it is 0.005. 

The activity concentration at the time 

of harvest is given 

(3.8) 

C i,f (t); concentration of activity in 

plant type i at time of harvest, 

f i ; interception factor 

for plant type i, 

A i ; total deposition (Bq.m –2 ) of 

plant type i at time of harvest, 

λ w ; loss rate (d –1 ) 

due to weathering, 

λ r ; decay rate (d –1 ), 

Δt; time span between deposition 

and harvest (d) 

The approach for pasture grass is 

different because of its continuous 

harvest. Here, the decrease in activity 

due to growth dilution is explicitly 

considered. 

C g,f (t); activity concentration 

(Bq.kg –1 ) in grass at time t 

after deposition, 

f g ; interception factor for grass, 

A g ; total activity deposited onto 

grass (Bq.m –2 ) 

Y g ; yield of grass at time of 

deposition (kg.m –2 ) 

a; fraction of activity translocated 

tot the root zone, 

λ b ; dilution rate by increase 

of biomass (d –1 ), 

λ t ; rate of activity decrease (d –1 ) 

due to translocation to the 

root zone 

For the weathering rate constant λw; a 

value equivalent to a half-life 14 d is 

taken from Farmland code (NRPB, 

1995) and for rate of activity decrease 

due to translocation to the root zone 

λt; 1.16x10-2 d-1 with a contribution 

fraction a= 0.05 using different measurement 

of grass contamination after 

the Chernobyl accident are assumed 

[Pröhl, 1990]. For plants that are only 

partly used for animal feeding or 

human consumption the translocation 





from leaves to the edible part of the 

plant has to be considered. This process 

strongly depends on the physiological 

behaviour of the element 

considered. It is important for mobile 

elements such as caesium, iodine, 

tellurium whereas for immobile 

elements including strontium, barium, 

zirconium, niobium, ruthenium, 

cerium, plutonium only direct deposition 

onto edible parts of the plants 

play role. Translocation process is 

quantified by translocation factor Ti, 

which is defined as the fraction of the 

activity deposited on the foliage being 

transferred to the edible parts of the 

plant until 

harvest. It is dependent on the 

element, plant type and time between 

deposition and harvest. Translocation 

factors for agricultural food products 

for caesium, strontium and other 

elements were taken from IAEA 

TRS-472 (2010). Translocation factors 

for only the ripening stage is applied 

in KIANA Advance Computational 


3.5 Root uptake 

of radionuclides 

The estimation of the root uptake of 

radionuclides assumes that the radionuclides 

are well mixed within the entire 

rooting zone. The concentration 

of activity due to root uptake is calculated 

from the concentration of activity 

in the soil using transfer factor TFi 

that gives the ratio of concentration of 

activity in plants (fresh weight) and 

soil (dry weight) 

C i,r (t) = TF i C s (t) 

C i, r (t); concentration of activity 

(Bq/kg) in plant type i due to 

root uptake at time t after the 

deposition, 

TF i ; soil-plant transfer factor for 

plant type i, 

Cs(t); concentration of activity 

(Bq/kg) in the root zone of 

soil at time t 

The soil conditions which soil-plant 

transfer factors are based are often 

characterised by a low pH value together 

with a high organic content, 

and low contents of clay, potassium 

and calcium. Such soils are frequently 

found in upland areas, Scandinavia, 

and parts of Eastern Europe. (Pröhl, 

G., and Müller, H., 1993) The concentration 

of activity in the root zone of 

soil is given by; 

A s ; total deposition to soil 

(Bq.m –2 ) 

L; depth of root zone (m) 

ρ; density of soil (kg.m –3 ) 

λ s ; rate of activity decrease due 

to migration out of the root 

zone 

λ r ; rate of fixation (d –1 ) 

The migration rate λ s is estimated 

according to; 

v a ; velocity of percolation water 

in soil (m.a –1 ) 

K d ; distribution coefficient 

(cm 3 .g –1 ) 

θ; water content of soil (g.g –1 ) 

3.6 Contamination 

of animal products 

The contamination of animal products 

results from the activity intake of 

the animals and the kinetics of the 

radionuclides within the animals. 

Inhalation of radionuclides by the 

animals is not considered; this pathway 

may be relevant for milk contamination 

in certain cases, but it is 

unimportant for resulting doses. The 

amount of activity ingested by the 

animals is calculated from the concentration 

of activity in the different 

foodstuffs and the feeding rates; 

A a,m (t); activity intake rate of the 

animal m (Bq.d –1 ), 

K m ; number of different feedstuffs 

fed to the animal m, 

C k (t); activity concentration 

(Bq.kg –1 ) in feedstuffs k, 

I k,m (t); feeding rate (kg.d –1 ) for 

feedstuffs k and animal m 

Soil ingestion is also considered in 


Code. Soil intake of animals 

varies widely depending on the 

grazing management and the condition 

of the pasture. If the feeding of 

mechanically prepared hay and silage 

during winter and an intensive 

grazing regime on well fertilized 

pasture are assumed a mean annual 

intake of 2.5% of the grass dry matter 

intake seems to be appropriate. This 

nuclide independent value is equivalent 

to soil-plant transfer factor of 

5x10-3 and it is added to the transfer 

and resuspension factor in KIANA 


Code. This means that for all elements 

with a transfer factor lower than this 

value, soil eating is the dominating 

long term pathway for the contamination 

for milk and meat from grazing 

cattle, presuming that resorption in 

the gut is the same for soil-bound and 

plant incorporated radionuclides. 

Seven different animal products, 

namely cow, sheep and goat milk, and 

lamb, beef cattle, egg and chicken, 

can be modelled by KIANA Advance 

Computational Computer Code. 

Transfer of radionuclides from fodder 

into animal products is calculated as 

follows: 

C m (t); activity concentration 

in animal product m at time t, 

TF m ; transfer factor (d.kg –1 ) 

for animal product m, 

j; number of biological transfer 

rates, 

a mj ; fraction of biological transfer 

rates, 

λ b,mj ; biological transfer rate j (d –1 ) 

for animal product m 

For sheep and goat milk transfer 

factors 10 times higher than for cow 

milk are assumed. For lamb, goat’s 

meat, and chicken, the transfer was 

estimated from the feed-beef transfer 

factor by applying correction factors 

for the lower body mass. Correction 

factors are 3 for lamb, and goat’s meat 

and 100 for chicken. [Müller, H. and 

Pröhl, G., 1993] Biological turnover 

rates of animal products were taken 

from DOZAE M, AIREM and DoseCAL. 

3.7 The processing and 

storage of foodstuffs 

The processing and storage of foodstuffs 

in order to take advantage of the 

radioactive decay and dilution during 

these processes are taken into account 

in the model. The enrichment of minerals 

in the outer layers of grains and 

the fractionation in the milling products 

is considered. Besides, the radioactive 

decay during processing and 

storage is taken into account. The storage 

presumes the stability of the foodstuffs 

or the possibility to convert the 

foodstuffs into stable products. Storage 

times are considered to be mean 

time between the harvest and beginning 

of product consumption. Concentration 

of activity in products is 

calculated from the raw product by 

the following relation: 

C k (t) = C ko (t–t pk )P k exp(–λ t pk ) 







| | Fig. 3. 

Code Algorithms of contamination of plant products as a function of time results from the direct contamination of the leaves and the activity transfer from the soil 

by root uptake and re-suspension that used in construction of KIANA Advance Computational Computer Code. 

| | Fig. 4. 

Code Algorithms calculation of Inhalation doses for each incremental time 

step (in days) that used in construction of KIANA Advance Computational 


C k (t); activity concentration 

(Bq/kg) in product k ready 

for consumption at time t, 

C ko ; activity concentration 

(Bq/kg) in raw product 

at time t, 

P k ; processing factor 

for product k, 

λ r ; radioactive decay constant 

(d –1 ), 

t pk ; storage and processing 

time (d) for product k 

3.8 Activity intake and 

exposure 

The intake of activity by humans is 

calculated from the time-dependent 

concentrations of activity in foodstuffs 

and the human consumption rate: 

A h (t); human intake rate (Bq.d –1 ) 

of activity, 

C k (t); concentration of activity 

(Bq.kg –1 ) of foodstuff k, 

V k (t); consumption rate (kg.d –1 ) 

of foodstuff k 

The foodstuffs are assumed to be 

locally produced. Food consumption 

data that is very important for 

calculating dose exposure by ingestion 

pathway is different depending 

on where people live. Country specific 

data on consumption of food products 

have been used to lead to realistic 

modelling. The dose Ding(t) due to 

ingestion of contaminated foodstuffs 

within time t after the deposition, is 

given by the following; 

D ing (t); ingestion dose (Sv) 

DF; age dependent dose factor 

for ingestion (Sv.Bq –1 ) 

4 Total dose calculation 


Code calculates yearly doses for 

each age group and for each sector – 

segment after the accident. Agricultural 

food products' activities are 

calculated at each year's harvest, 

grass and animal products' activities 

are calculated on a monthly basis. 

All aforementioned pathways are 

included in dose calculations as shown 

below: 

Dose total = Dose inhalation + Dose ingestion + 

Dose cloudshine + Dose groundshine 






| | Fig. 5. 

Code Algorithms calculation of Activity concentration of plant products Root uptake of radionuclides that used in construction of KIANA Advance Computational 


| | Fig. 6. 

Code Algorithms concentration, activity intake rate of the animal m (Bq. d -1 ), that used in construction of KIANA Advance Computational Computer Code. 

Dose total ; total dose (Sv) 

Dose inhalation ; inhalation dose (Sv) 

Dose ingestion ; ingestion dose (Sv) 

Dose cloudshine ; cloudshine dose (Sv) 

A person is assumed to be as an infant 

up to 1 year, as a child up to 9 years, as 

teen up to 16 years and as an adult up 

to 70 years; namely when calculating 

long term doses after the accident 

growing up of a person is taken into 

account in terms of his/her food 

consumption habits, sensitivity to 

doses and occupancy factors. 

4.1 Calculation of collective 

doses 

The impact of an accident on the 

population as a whole depends not 

only on the deposition, atmospheric 

activity levels and dose obtained, 

but also on the population living in 






| | Fig. 7. 

Code Algorithms of Concentration of activity in products is calculated from 

the raw product that used in construction of KIANA Advance Computational 


| | Fig. 8. 

Code Algorithms intake of activity by humans is calculated from the 

time-dependant concentrations of activity in foodstuffs and the human 

consumption rate that used in construction of KIANA Advance 

Computational Computer Code (upper part of the diagram). 

Code Algorithms for dose Ding(t) due to ingestion of contaminated 

foodstuffs within time t after the deposition, is given by the following 

that used in construction of KIANA Advance Computational Computer 

Code (lower part of the diagram). 

that particular area. For example 

the deposition, atmospheric activity 

levels, dose obtained and individual 

health risk, due to any NPP accident, 

may be very high, but these high 

values may not mean anything if there 

is no one living there. Consequently, 

better representation of the collective 

doses or risk of an accident, nuclear 

and nonnuclear, can be obtained by 

multiplying the individual dose or 

health risk by the number of people 

living in the receptor. For this study, 

average values all over the geographical 

regions were taken into 

account, since data does not vary 

considerably over the regions. On 

the other hand. Transfer factors for 

animal- feeds and soil-plants, and fixation 

rates, distribution coefficients, 

translocation factors, dose conversion 

factors and metabolic turnover rates 

in animals for all related isotopes, and 

processing factors and storage days 

for food products, weathering rates, 

interception factors and soil density, 

water content of soil, percolation 

water velocity, dilution factor of 

the grass, depth of root zone, the 

references in which Cs-137 default 

values were taken for validation study, 

were used in KIANA Advance Computational 

Computer Code during 

simulation of the case studies. Since 

most of these data are not dependent 

on location. 

5 Result and discussions 

Dispersion of radionuclides is also an 

application area of KIANA Advance 

Computational Computer Code. User 

supplied inputs for KIANA Advance 

Computational Computer Code calculations 

are pollutant species 

characteristics, emission parameters, 

gridded meteorological fields and 

output deposition grid definitions. 

The horizontal deformation of the 

wind field, the wind shear, and the 

vertical diffusivity profile are used to 

compute the dispersion rate. Gridded 

meteorological data are required for 

regular time intervals. The meteorological 

data fields may be provided on 

one of the different vertical coordinate 

system: Pressure-sigma, pressure 

absolute, terrain-sigma or a hybrid 

absolute-pressure-sigma The doses 

and time dependant radioactivity concentration 

values in the food products 

and pasture grass predicted by KIANA 


Code have been compared with those 

of different codes (AIREM,DOZA) 

which participated in assessment task, 

and data measured in Boshehr, and 

Finland after Chernobyl accident. 

Radionuclide 

Activity (Bq) 

Sr-89 

8.5E+09 

Kr-90 

6.7E+13 

Rb-90 

6.4E+13 

Sr-90 

2.2E+07 

Sr-91 

2.6E+11 

Sr-92 

2.1E+11 

Mo-99 

1.1E+09 

Ru-103 

9.3E+08 

Ru-106 

1.3E+07 

Ru-106 

1.3E+07 

Ru-106 

1.3E+07 

Te-131 

9.3E+10 

I-131 3.1E+13 

Te-132 

1.2E+10 

I-132 8.3E+13 

Te-133 

1.6E+11 

I-133 6.8E+13 

Xe-133 

1.7E+13 

I-134 6.3E13 

Cs-134 

1.8E+12 

I-135 5.1E+13 

Xe-135 

1.1E+13 

Cs-137 

2.8E+12 

Xe-138 

4.6E+13 

C-138 4.9E+13 

Ba-139 

9.9E+11 

Ba-140 

1.1E+10 

La-140 

1.4E+09 

141-Ce 

1.8E+09 

Ce-144 

2.0E+08 

Br-84 

1.5E+13 

Kr-85m 

1.2E+13 

Kr-85 

3.3E+09 

Br-87 

3.7E+13 

Kr-87 

3.9E+13 

Kr-88 

4.9E+13 

Rb-88 

4.9E+13 

Kr-89 

6.7E+13 

Rb-89 

7.1E+13 

Pr-144 

1.8E+08 

Zr-95 

1.2E+09 

Nb-95 

1.2E+07 

Zr-97 

7.4E+10 

Nb-97 

6.7E+10 

Na-24 

2.7E+11 

K-42 1.2E+12 

Fe-59 

1.9E+07 

Co-58 

7.4E+07 

Cr-51 

1.4E+08 

Mn-54 

1.9E+07 

Co-60 

2.0E+08 

Activities (Bq) 

I-131 3.1E+11 

I-132 8.4E+11 

I-133 6.9E+11 

I-134 6.3E+11 

I-135 5.1E+11 

| | Tab. 1. 

Radionuclide release to environment after 

severe accident at typically WWER-1000 NPP 

such as Boushehr. 

Those codes are dynamic (timedependent), 

and only one of them; i.e. 

DoseCAL, is quasi-equilibrium. Since 


Code is developed as dynamic 

software (such as DoseCAL), only 

dynamic codes' results are presented 

for comparison. KIANA Advance 

Computational Computer Code has a 






| | Fig. 9. 

Code Algorithms for calculation of ground level air concentration at downwind distance x in the sector) p (Bq/m3), when the source and receptor on the same 

building surface that used in construction of KIANA Advance Computational Computer Code. 

capability to make simulation with 

seven pollutants at a time at most. 

Since some more radionuclides 

considered being most important in 

terms of their effects in the environment 

are used to represent accidental 

release of radionuclides in the literature, 

HYSPLIT model's source code 

has been modified to simulate more 

pollutants to provide us easiness for 

this study. 

In this study, dry deposition velocity 

is assumed to be a constant for 

each radionuclide and surface type. 

the dry deposition velocity values for 

agricultural surface type were used in 

our simulations. To strengthen our 

assumption, size of the particles 

released into environment in the case 

of a nuclear accident was also investigated. 

Release height is another 

important parameter for subsequent 

dispersion modelling in KIANA 


Code. Literature studies show that 

variations of the initial plume rise 

below the mixing height only slightly 

affect the results outside the local 

scale, whereas plume rise above that 

level led to significantly changed patterns 

with relatively little depositions 

on the local and meso-scales. Thus, 

a release into the atmospheric 

boundary level compared with a 

release to the free troposphere leads 

to large differences in the deposition 

patterns and lifetimes (a week or 

more) of radionuclides within the 

atmosphere. Release height was 

assumed as a line source between 

50-100 meter considering all the accident 

type, release points in the reactor 

and plume rise. In 1986, there was a 

recommendation to postpone the 

open field sowing of lettuce, spinach 

and other fast growing vegetables. 

Although it is not clear to what extent 

this recommendation was implemented 

across all regions, the fact 

that KIANA Advance Computational 

Computer Code did not account for 

any delay in sowing. However, only 

root uptake for leafy vegetables was 

taken into account in DoseCAL. Leafy 

vegetables activities predicted by 


Code are within the uncertainty 

band of the measured values and the 

best of all other code results. The 

probability for T-test for is 0.834, 

which is close to one. The differences 

between the predictions of the codes 

which participated in VAMP exercise, 

may be raised from misinterpretation 

of site-specific information; namely 

taking into account different assumptions, 

or using different soil-plant and 

feed-animal transfer factors as stated 

in IAEA TECDOC-904 (1996). Inhalation 

and external doses predicted 

by KIANA Advance Computational 

Computer Code as the as the DoseCAL 

calculations are rather consistent 

compared to other codes' predictions. 

Ingestion doses predicted by KIANA 


Code, on the other hand, is lower 

compared to the other codes. Since in 

ingestion module of KIANA Advance 

computational Computer Code, mushroom, 

fish, game animals are not taken 

into account, whereas other food 

products, i.e. fruits, root and fruit vegetables, 

eggs have been considered as 

default. it is almost equal to beef consumption, 

and most of the ingestion 

doses calculated by most of the models 

participated in validation exercise 

were incurred from fish consumption. 

Hence, the difference in ingestion 

dose prediction in KIANA Advance 

Computational Computer Code can be 

attributed to fish pathway. Ingestion 

doses are highly dependent on consumption 

rates as seen from the differences 

between the doses for average 

and maximum individuals. Inhalation 

doses are the highest for the children, 

though the highest inhalation DCFs 

are of infants, breathing rates for 

the children are higher than for the 

infants. Inhalation dose for teens and 






adults are lower than children, since 

DCF’s for radioisotopes considered in 

case study for children are higher than 

those for adults except caesium isotopes. 

External doses are calculated 

for infants and others (child, teen and 

adult). Although DCF’s for infants are 

1.5 times higher than the others, the 

correction factor for shielding is lower 

for infants than others, hence external 

doses are lower for infants. External 

ground doses are lower for infants 

too, as far as the years passed after the 

accident is concerned. In the case of 

implementation of countermeasures 

on food consumption restrictions in 

the first year after the accident, the 

ingestion and total doses for average 

individuals for all age groups can be 

predicted by KIANA Advance computational 

Computer Code. . the most 

dose contributing isotopes are Cs-134, 

Cs-137 and I-131 in the first year after 

the accident. In the long term, Cs-134 

and Cs-137 (Table 1) remain in the 

environment due to their long radioactive 

half-lives. The dose consequence 

of Xe-133 is the least amongst 

others due to its very short half-life, 

i.e. 5.25 days and its inertness. Lifetime 

doses incurred from Cs-137, 

Cs-134 and I-131 are more than 95% 

of total doses. Ingestion doses are the 

highest for the infant, child, adult and 

teen; respectively in the first year after 

the accident since the ingestion DCF 

for I-131 for the infants is the highest. 

Infant ingestion doses remain the 

highest as years pass after the accident, 

since infant's growing up is 

taken into account and their food 

consumption increases when they are 

growing. 

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M.G., Rechard, R.E, Rudeen, D.K. and 

Swift, EN.: 1993. Uncertainty and 

Sensitivity Analysis Results Obtained in 

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Authors 

A. Haghighi Shad 

PhD in Nuclear Energy Eng 

Department of Nuclear Eng. 

Science and Research Branch 

of Islamic Azad University 

Tehran, Iran 

D. Masti 

Assistant of Prof. Azad University 

of Boushehr 

Boushehr NPP 

Manager of Research and 

Development in BNPP-1 

M. Athari Allaf 

Assistant of Prof. 




Tehran, Iran 

K. Sepanloo 

Associate of Prof. Reactor and 

nuclear safety school 

Nuclear Science and Technology 

Research Institute (NSTRI) 

Tehran, Iran 

S.A.H. Feghhi 

Prof. Shahid Beheshti University 

of Tehran 


Deputy Manager of execution and 

Research in Nuclear Eng. Faculty 

Tehran, Iran 

R. Khodadadi 

Consultant 



Tehran, Iran 

121 

EVENTS 

Event Report: Vertiefungskurs 2017: 

Zukunftsmanagement – zentrale 

Lösungsansätze für Kernanlagen 

Matthias Rey 

Zukunftsmanagement erfordert sorgfältige Planung und Wissen darüber, welche Optionen zur Verfügung 

stehen, wieweit Optimierungen sinnvoll sind und welche Maßnahmen und Prozessänderungen sich allenfalls bereits 

anderswo bewährt haben. Der Vertiefungskurs 2017 des Nuklearforums Schweiz nahm diese Thematik auf. Im Zentrum 

standen am ersten Kurstag Lösungsansätze zum Optimieren von Systembetrieb und Instandhaltung. Am zweiten Tag 

standen die Mitarbeitenden in seiner sich verändernden Umwelt im Fokus. Als Novum wurden dieses Jahr an beiden 

Nachmittagen die Themen der Inputreferate des Vormittags in Workshops vertieft diskutiert. 

Der neue Präsident der Kommission 

für Ausbildungsfragen des Nuklearforum 

Schweiz, Thomas Kohler, 

begrüßte die Teilnehmenden und 

wies auf das neue Format mit den 

Workshops hin, das aufgrund der 

Feedbacks zu vergangenen Kursen 

eingeführt worden ist. 

Optimierung von Systembetrieb 

und Instandhaltung 

In der Einleitung zum ersten Block 

wies Andreas Pfeiffer, Leiter des Kernkraftwerks 

Leibstadt, darauf hin, dass 

in der Schweiz bald das letzte KKW 

im deutschsprachigen Raum stehen 

dürfte. Die Betreiber stünden unter 

Druck seitens der Politik, müssten ihre 

Koten optimieren und sähen sich 

mit einer schrumpfenden Lieferantenbasis 

konfrontiert. 

Wie die ABB ihre Lieferanten 

bewirtschaftet legte Nikolaus Gäbler, 

Head of Supply Chain Management 

der Business Unit Grid Automation, 

dar. In der Schweiz gibt es ihm zufolge 

praktisch nur noch hoch spezialisierte 

Anbieter. Zudem sei die Supply 

Chain im Servicegeschäft besonders, 

charakterisiert durch ihre Kurzfristigkeit, 

wenig Beständigkeit und viele 

Sonderwünsche. Damit „die linke 

Hand genau weiss, was die rechte 

tut“, habe die ABB weltweit ein 

IT-Tool für das Lieferantenmanagement 

eingeführt, in das sämtliche 

Anfragen und Offerten eingetragen 

werden. Um langfristige Partnerschaften 

zu schaffen müsse man auch 

das Zwischenmenschliche berücksichtigen 

und sich manchmal mit 

Lieferanten treffen, ohne dass dabei 

gleich ein Geschäft entsteht. Um 

Kosten und Prozesse zu optimieren 

oder Abhängigkeiten zu reduzieren, 

kommt laut Gäbler vor, dass die ABB 

einen Lieferanten gleich komplett 

übernimmt. 

Im zweiten Referat zum Thema 

Reverse Engineering zeigte Florian 

Kanoffsky von der KSB AG auf, was ein 

Unternehmen tun kann, wenn seine 

Ersatzteile nicht mehr geliefert 

werden. Wenn sich kein anderer 

Lieferant findet und der Austausch der 

entsprechenden Komponenten keine 

Option ist, können Teile nachgebaut 

werden, was dann eben als «Reverse 

Engineering» bezeichnet wird. 

Kanoffsky beschrieb den typischen 

Ablauf solcher Aufträge von der 

Vermessung über das Erstellen von 

3D- und Guss-Modellen bis zur 

Endbear beitung. Bei der Planung 

müsse gerade in der Nuklearbranche 

den Genehmigungsprozessen, der 

Zeich nungsfreigabe sowie den Prüfungen 

und Abnahmen genug Zeit 

| | Vertiefungskurs 2017, wie gewohnt im Hotel Arte in Olten 

beige messen werden. Auch rechtliche 

Aspekte wie Patente und allenfalls 

Geheimhaltungsklauseln für Zeichnungen 

und Pläne in bestehenden 

Verträgen gelte es unbedingt zu 

beachten. 

Theoretische Ansätze, 

Fallstudien und Erfahrungsberichte 

Mit dem Referat von Giovanni 

Sansavini vom Reliability and Risk 

Engineering Laboratory der ETH Zürich 

zu Importance Measures ging es 

anschließend von der Praxis in die 

Theorie. Importance Measures quantifizieren 

die Bedeutung von Komponenten 

oder Ereignissen bei der 

Beurteilung der Systemperformance. 

Sie seien eine große praktische Hilfe 

für Systemdesigner und -manager, 

Events 

Event Report: Nuklearforum Schweiz – Future Management – Key Solutions for Nuclear Facilities ı Matthias Rey


122 

EVENTS 

da sie Schwachstellen im System 

aufzuspüren helfen und Richtlinien 

für die Verbesserung liefern. 

Danach ging es wieder in Richtung 

Praxis, genau gesagt zur Probabilistischen 

Sicherheitsanalyse (PSA) 

in KKW. Dusko Kancev, Fachverantwortlicher 

PSA Modellentwicklung 

und Sicherheitsindikatoren des Kernkraftwerks 

Gösgen, zeigte anhand 

einer PSA-Fallstudie, wie in KKW die 

Überwachungsanforderungen unter 

Berücksichtigung der Ausrüstungsalterung 

optimiert werden können. 

Mit dem verwendeten Modell kann 

die Alterung von Komponenten 

explizit, und nicht wie bei der 

„ traditionellen“ PSA stationär, dargestellt 

und letztendlich die Überwachungsintervalle 

der untersuchten 

Komponenten optimiert werden. 

Mit dem letzten Vortrag vor der Mittagspause 

folgte dann der erste, am 

Vertiefungskurs mittlerweile traditionelle 

Blick über den Tellerrand. 

Ronald Meier, Sektionsleiter Technische 

Organisation Zürich des Bundesamts für 

Zivilluftfahrt, stellte optimierte Instandhaltungsstrategien 

für den Langzeitbetrieb 

vor. Er ging auf Aspekte wie 

Ersatzteilstrategien und Lagerhaltung 

ein. Punkto Ersatzteile zahlen sich 

große Flotten des gleichen Flugzeugtyps 

sowie die Zusammenarbeit mit 

anderen Fluggesellschaften aus. Auch 

bei der Lagerhaltung spielt das sogenannte 

Pooling eine zunehmende 

Rolle, ebenso das Auslagern von 

Ersatzteillagern und die Tendenz zu 

zentralen größeren Lagern und nur 

kleinen Lagern vor Ort. Sowohl bei der 

Diversifizierung der Zulieferer als auch 

bei Reparaturen durch Eigenpersonal 

sind Überprüfungen und Zulassungen 

durch die Behörden nötig. Dass auch 

die Ausbildung streng reguliert ist und 

entsprechend lange dauert, führt zusammen 

mit eher kleinen Löhnen bei 

großer Verantwortung zu gewissen 

Nachwuchsproblemen in der Flugzeuginstandhaltung. 

Ein weiteres 

Problem stellen gefälschte oder nicht 

zugelassene Ersatzteile dar. 

Diskussion in Gruppen 

Am Nachmittag fand dann die besagte 

Premiere mit vier zeitgleich laufenden 

Workshops statt, für die sich die Teilnehmenden 

im Vorfeld angemeldet 

hatten. Eine Gruppe beschäftigte 

sich mit der Frage, was verlängerte 

Betriebszyklen für die Instandhaltung 

bedeuten. Lagerhaltung 

und Bestellkontrakte: vorbeugende 

Instandhaltung oder ‹run to 

failure›? lautete das Thema des 

zweiten Workshops. Die dritte Gruppe 

| | Diskussion im Workshop... | | ... und Präsentation im Plenum 

befasste sich mit der System Health 

und Systemzustandsberichten hinsichtlich 

des Kostenoptimierungspotenzials. 

Im vierten Workshop 

ging es um Möglichkeiten der Wertschöpfung 

und Belastungen der 

technischen Systeme, die der 

Lastfolge betrieb von KKW mit sich 

bringen kann. Der erste Kurstag 

endete mit der Präsentation der 

Resultate aus den einzelnen Workshops 

unter der Leitung von Michael 

Dost, Leiter des Kernkraftwerks 

Beznau, endete der erste Kurstag. 

Kompetenzanpassung 

und -transfer 

Den zweiten Tag des Vertiefungskurses 

eröffnete Martin Saxer, Leiter 

des Kernkraftwerks Mühleberg, mit 

dem Hinweis auf die Bedeutung der 

Menschen und ihrer Kompetenzen 

für das Zukunftsmanagement. Das 

erste Referat von Frank Sommer, 

Senior Vice President, Center of Competence 

Operations der PreussenElektra 

GmbH, erläuterter die Herausforderungen 

und Erfahrungen bei 

organisatorischen Veränderungen. 

Sommer erläuterte, wie der Energiekonzern 

E.ON entstanden ist und wie 

daraus letztlich die PreussenElektra 

hervorging. Er beleuchtete die Auswirkungen 

großer organisatorischer 

Veränderungen und Neuausrichtungen 

auf die Mitarbeitenden. In 

der Vergangenheit habe die große 

Herausforderung in der Integration 

von Kraftwerken etablierter Unternehmen 

aus verschiedenen Ländern 

in ein Großunternehmen bestanden. 

Dagegen stehe heute der Erhalt der 

Kompetenz für den sicheren Betrieb 

der Anlagen bis zur Stilllegung im 

Fokus. Um Sicherheit für die Mitarbeitenden 

zu erreichen und einen 

wirtschaftlichen Rückbau sicherzustellen 

sei es enorm wichtig, Nachbetrieb 

und Rückbau frühzeitig zu 

planen. Die Entwicklung eines internationalen 

Geschäfts schaffe in 

diesem Zusammenhang Perspektiven 

für die Belegschaften. 

Der darauf folgende Vortrag von 

Christer Johansson, Deputy Director 

Maintenance der Forsmarkskraftgrupp 

AB bei Vattenfall, stand unter ganz 

anderen Vorzeichen, da er von Strategien 

zur Laufzeitverlängerung 

handelte. Neben Strategien bei der 

Instandhaltung ging Jonansson vertieft 

auf den Kompetenzerhalt beim 

Personal ein. In Forsmark wird zum 

Beispiel wo immer möglich jüngeres 

Personal mit weniger Erfahrung 

zusammen mit langjährigeren Mitarbeitenden 

eingesetzt, oft auch unter 

Miteinbezug von Lieferanten. Darüber 

hinaus sei das Vorhandensein von 

Designregeln, Komponentenspezifikationen, 

Testergebnissen und weiterer 

Dokumentationen sowie das Wissen, 

wie die Komponenten im System 

funktionieren, Grundvoraussetzung 

für den Kompetenzerhalt. 

Know-how-Management und 

Know-why-Management in der 

Nuklearindustrie lautete der Titel 

des Beitrags von Tomas Hahn, Vice 

President Products and Projects der 

Areva GmbH. Die aktuelle wirtschaftliche 

Lage der europäischen 

Energieindustrie führt laut Hahn zu 

einem immer geringeren Volumen 

an Engineering- Aufgaben. Die neuen 

Schwerpunkte lägen beim Lebensdauer 

management und Modernisierungen, 

der Erhöhung von Sicherheitsstandards 

sowie der Weiterentwicklung 

des Stands von Wissenschaft 

und Technik. Die langen Laufzeiten 

von Kernkraftwerken bedingen den 

Transfer von Know-how von einer 

Generation von Ingenieuren zur 

nächsten. Daneben sei auch das Knowwhy-Training 

von großer Bedeutung, 

also die Vermittlung von Basis hintergrund 

wissen wie bestimmte Anlagen- 

Designs, Sicherheitskonzepte mit den 

Forderungen nach Redun danzen, 

Standards etc. und nicht zuletzt die 

Interaktion zwischen den verschiedenen 

Reaktorsystemen. Damit seien 

die Voraussetzungen gegeben, um 

komplexe technische Fragestellungen 

in einem anspruchsvollen Genehmigungsumfeld 

unter schwierigen 

Markt bedingungen professionell zu 

bearbeiten und die Bedürfnisse der 

Kunden zu befrie digen. 

Events 

Event Report: Nuklearforum Schweiz – Future Management – Key Solutions for Nuclear Facilities ı Matthias Rey


Sinkende Verfügbarkeit 

und steigender Bedarf 

Auf die Bedeutung des Kompetenzmanagements 

für die Aufsicht angesichts 

der aktuellen Entwicklungen in 

der Kerntechnik ging anschließend 

Holger Knissel, Fachspezialist Mensch 

und Organisation beim Eidgenössischen 

Nuklearsicherheitsinspektorat 

(Ensi), ein. In der Nuklearindustrie 

stehe aktuell eine sinkende Kompetenzverfügbarkeit 

einem steigenden 

Bedarf gegenüber. Gründe für die 

sinkende Verfügbarkeit sind der 

Generationswechsel in den Betriebsorganisationen, 

die wegen der politischen 

Randbedingungen abnehmende 

Attraktivität die zu Rekrutierungsproblemen 

führt, sowie der 

Kostendruck aufgrund der wirtschaftlichen 

Lage. Auf der anderen Seite 

nehme der Kompetenzbedarf zu, weil 

das Spektrum an benötigten Kompetenzen 

aufgrund der technologischen 

Entwicklungen immer breiter wird, 

weil die Alterung der Anlagen neue 

Fragestellungen aufwirft und weil der 

Support der Zulieferer abnimmt. 

Daraus folgerte Knissel, dass eine 

Kompetenzlücke zu entstehen droht. 

Dem könne und müsse mit aktivem 

Kompetenzmanagement entgegengewirkt 

werden. 

Der nächste Beitrag stellte einen 

weiteren Ausflug in die Aviatik dar: 

Nutzbarmachen von Erfahrungen 

aus ‹near misses› von Stefan Oser, 

Leiter Technical Training der Swiss International 

Air Lines Ltd. Er ging unter 

anderem der Frage nach, wie ein 

gesundes und vernünftiges Maß an Anleitungen, 

Checklisten und sonstiger 

Dokumentation für Instandhaltungsund 

Reparaturarbeiten aussieht und 

wie man die Leute dazu bringt, Vorkommnisse 

und Ab weichungen zu 

melden. Anhand von Erlebnissen aus 

seiner persönlichen Karriere legte er 

dar, wie wichtig lebenslanges Lernen, 

insbesondere aus Fehlern, ist. 

Freiheit der Forschung 

gewährleistet 

Für das letzte Inputreferat des diesjährigen 

Vertiefungskurses zeigte 

Horst-Michael Prasser von der ETH 

Zürich auf, was für den langfristigen 

Kompetenzerhalt in der Schweiz 

nötig ist. Er betonte eingangs, dass 

das neue Energiegesetz keine Einschränkungen 

für die Nuklearforschung 

beinhalte und das die Freiheit 

der Forschung gewährleistet sei. Auch 

gebe es keine spezifischen Budgetkürzungen 

für die Nuklearforschung 

am Paul Scherrer Institut PSI und die 

Professuren an den eidgenössischen 

Hochschulen. Weiter brauche es 

Kompetenzerhalt und Kompetenzentwicklung 

bei Kerntechnikern, angehenden 

Kerntechnikern sowie auch 

Kernenergiegegnern, denn ein profunder 

Disput über Kerntechnik sei 

eine objektive Notwendigkeit unserer 

Zeit. Offene, proaktive Kommunikation 

auch zu Problemen sei unerlässlich, 

ebenso wie breit angelegte 

Forschung und Bildung. 

| | Horst-Michael Prasser :«Ein profunder 

Disput über Kerntechnik ist eine objektive 

Notwendigkeit unserer Zeit.» 

Am Nachmittag beschäftigten sich 

drei Workshop-Gruppen unter der 

Leitung von Vertretern der Kernkraftwerke 

Gösgen, Mühleberg und 

Leibstadt mit der Frage nach dem 

richtigen Maß beim Erkennen und 

Melden von Befunden. Der vierte 

Workshop thematisierte den Kulturwandel 

und den Umgang mit Multinationalität 

in Kernkraftwerken. Die 

Ergebnisse wurden ebenfalls wieder 

im Plenum präsentiert und diskutiert, 

dieses Mal moderiert von Herbert 

Meinecke, dem Leiter des Kernkraftwerks 

Gösgen. Der Geschäftsführer 

des Nuklearforums, Beat Bechtold, verabschiedete 

anschließend die Teilnehmenden 

des Vertiefungskurses 

mit dem Hinweis, dass dieser von nun 

an voraussichtlich im Zweijahres- 

Rhythmus stattfindet. 

Author 

Matthias Rey 

Nuklearforum Schweiz / 

Forum nucléaire suisse 

Frohburgstrasse 20 

4600 Olten, Switzerland 

123 

KTG INSIDE 

Inside 

KTG: Wichtige Terminhinweise 

in eigener Sache 

Ankündigungen zum Vortag unserer diesjährigen Jahrestagung, 

dem 49 th Annual Meeting on Nuclear Technology 

(AMNT 2018) vom 29. bis 30. Mai 2018 im Estrel-Hotel, 

Berlin: 

33 

KTG-Mitgliederversammlung 

• Wann? Montag, 28. Mai 2018, 16.00 Uhr 

• Wo? Estrel Convention Center, Raum IV 

(2. OG), Sonnenallee 225, 12057 Berlin 

33 

Verleihung des Karl-Wirtz-Preises 


• Wo? Estrel Convention Center, Raum IV 

(2. OG), Sonnenallee 225, 12057 Berlin 

33 

Get-together der KTG (auch für Nicht-Mitglieder) 


• Wo? Estrel Convention Center, Leaf, 

Sonnenallee 225, 12057 Berlin 

KTG Fachgruppe Thermo- und 

Fluiddynamik 

Die KTG Fachgruppe Thermo- und Fluiddynamik 

beschäftigt sich mit 

• der Entwicklung, Validierung und Anwendung von 

Methoden und Computerprogrammen zur Berechnung 

von Strömungsvorgängen im Reaktorkühlkreislauf 

(RKL) sowie dem Containment, 

• der zur Validierung der Rechenmethoden erforderlichen 

Experimente einschließlich der Entwicklung von 

Messtechniken sowie 

• der Bestimmung von analytischen sowie experimentellen 

Unsicherheiten. 

Methoden, Computerprogramme und Experimente werden 

u.a. in kerntechnischen Verfahren genutzt, um Nachweise 

zu führen (Hersteller und Betreiber) oder unabhängig zu 

prüfen (Behörden und Gutachter) und die Einhaltung 

von Anforderungen aus dem kerntechnischen Regelwerk 

aufzuzeigen. Aktuelle Themen, die derzeit im Fokus der 

Fachgruppe stehen, sind die Weiterentwicklung und Validierung 

von eindimensionalen Systemcodes zur Simulation 

KTG Inside


124 

KTG INSIDE 

von innovativen Reaktorkonzepten mit passiven Sicherheitsmerkmalen, 

die Ertüchtigung von Computational Fluid 

Dynamic (CFD) Methoden zur Berechnung mehrphasiger 

Strömungszustände, die Entwicklung von Methoden zur 

Durchführung von Sensitivitäts- und Unsicherheitsanalysen 

für CFD Analysen. Des Weiteren werden aktuelle Fragestellungen 

zur technisch-wissenschaftlichen Absicherung 

des ver bleibenden Betriebs deutscher Kernkraftwerke und 

Forschungsreaktoren in der Fachgruppe aufgegriffen. 

Hierzu zählen u.a. Themen wie die mögliche Beeinträchtigung 

der Kernkühlung durch Isoliermaterial und oder 

Zinkboraten oder das sog. Neutronenflussrauschen. 

Der Vorstand der Fachgruppe, die derzeit um die 200 

Mitglieder besitzt, besteht derzeit aus 5 Personen. Dies sind 

Dr.-Ing. Andreas Schaffrath (Gesellschaft für Anlagen- und 

Reaktorsicherheit) gGmbH, der aktuell der Sprecher der 

Fachgruppe ist, Dipl.-Ing. Sören Alt (Hochschule Zittau, 

Görlitz), Dr.-Ing. Ingo Ganzmann (AREVA GmbH) und Prof. 

Dr.-Ing. Eckhart Laurien (IKE Stuttgart). Kassenwart und 

Kommunikationsbeauftragter der Fachgruppe ist Dr.-Ing. 

Jürgen Sydow (TÜV NORD Systems GmbH). Die Fach gruppe 

arbeitet – sofern dies thematisch erforderlich ist – interdisziplinär 

mit anderen Fachgruppen der KTG zusammen und 

organisiert z.B. gemeinsame Fach sitzungen auf dem jährlich 

stattfindendem Annual Meeting on Nuclear Technology 

(AMNT), KTG Fachtage oder Vortragsveranstaltungen. Die 

KTG Fachgruppe Thermo- und Fluiddynamik aktualisiert 

kontinuierlich ihren Internetauftritt. 

Die letzte große Veranstaltung der Fachgruppe war der 

Ende 2016 in Karlsruhe zusammen mit den Fachgruppen 

Reaktorphysik und Berechnungsmethoden und Reaktorsicherheit 

durchgeführte, 2-tägige Fachtag zu Aktuellen 

Themen der Reaktorsicherheit. Thematische Schwerpunkte 

des Fachtages waren neue Erkenntnisse aus den Bereichen 

Neutronenphysik, Anlagenbetrieb, BE-Lagerbecken, sowie 

Sensitivität, Entwicklung und Validierung von Codes 

sowie Tools zur Berechnung von Unsicherheiten und 

Sensitivitäten. Abgerundet wurde der Fachtag durch einen 

geselligen Abend. Der Fachtag war mit über 60 Teilnehmern 

gut besucht. Über den Fachtag wurde in der atw 

(International Journal for Nuclear Power, Heft 10, 2016) 

sowie der Kerntechnik (Heft 5, 2016) berichtet. Darüber 

hinaus wurden diverse Beiträge des Fachtages im Heft 3, 

2017 der Kerntechnik veröffentlicht. 

Aktuell engagieren sich zahlreiche Mitglieder substantiell 

an der Vorbereitung des AMNT 2018. Sie sind u.a. im 

Programmausschuss oder verschiedenen Auswahlausschüssen 

vertreten. Die Fachgruppe hat u.a. die Fokussitzung 

Safety of Advanced Nuclear Power Plants vor bereitet, 

in der zunächst ausgewählte Experten über aktuelle kerntechnische 

Entwicklungen in UK und China berichten. Es 

folgt dann ein Vortrag über eine Initiative der OECD/NEA 

zur Untersuchung und Bewertung thermohydraulischer 

Aspekte sog. passiver Sicherheitssysteme. Im Anschluss 

wird dann ein Vertreter des Bundesministeriums für 

Wirtschaft und Energie (BMWi) eine Übersicht über die 

derzeit in Deutschland durchgeführten Arbeiten im Bereich 

der Reaktorsicherheitsforschung geben. Es folgen abschließend 

zwei Vorträge, in denen herausragende BMWi 

finanzierte Forschungsarbeiten zu experimentellen und 

analytischen Untersuchungen passiver Systeme zur Beherrschung 

von Auslegungsstörfällen vorgestellt werden. 

Zusätzlich wurden bereits für die technischen Sitzungen 

des Key Topic Outstanding Know-How & Suitainable 

Innovations die eingereichten Abstracts gereviewt. 

Für das Jahr 2018 ist bereits – neben den üblichen 

Aktivitäten zur Vorbereitung des AMNT 2019 – zusammen 

mit der Sektion Süd eine Vortragsveranstaltung bei der 

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) 

gGmbH zu dem Thema Erweiterung der GRS Rechenkette 

für fortschrittliche Reaktoren geplant. 

Zu allen zuvor genannten Aktivitäten hoffen wir auf 

eine rege Teilnahme. 

Dr.-Ing. Andreas Schaffrath 

Sprecher der KTG Fachgruppe Thermo- und Fluiddynamik 

Kernfusion: Eine kleine Fachgruppe 

für ein Thema mit viel Zukunft 

Die Fachgruppe Kernfusion der KTG wurde erst 1997 

gegründet, also zu einer Zeit, als die KTG bereits 28 Jahre 

alt war. Derzeit hat sie 60 Mitglieder. Ihre thematischen 

Schwerpunkte liegen auf Fusionstechnologie und Plasmaphysik. 

Der Gründer und erste Sprecher der Fachgruppe war Dr. 

Gert Spannagel (FZK). Der Staffelstab wurde 2002 weitergegeben 

an Michael Gehring (Babcock Noell), der ihn über 

10 Jahre lang hochhielt. Ich selbst erhielt ihn dann 2013. 

Obwohl die Fachgruppe Kernfusion in der KTG von 

Anfang an eine etwas kleinere Fachgruppe war, hat sie 

doch immer wieder durch ihre Aktionen und ihre Präsenz 

auf der Jahrestagung munter zum Leben und Programm 

der KTG beigetragen. In vielen Technischen und Fach- 

Sitzungen auf den Jahrestagungen verfolgte und kommunizierte 

sie die Weiterentwicklung der Kernfusionstechnologie 

und machte von Anfang an klar, dass Kerntechnik 

eben mehr ist als die technische Beherrschung der Kernspaltung. 

Zu den Highlights der Vergangenheit gehörte 

sicherlich auf der JK 2007 der Plenarvortrag von Kaname 

Ikeda, damals erster „Director-General“ des ITER-Projekts. 

Auch in jüngerer Vergangenheit wurden interessante 

Aktivitäten entwickelt. So konnten wir 2016 Prof. Robert 

Wolf für einen Plenarvortrag auf der AMNT zum Thema 

Wendelstein 7-X gewinnen. Der W7X hatte erst wenige 

Monate zuvor sein erstes Plasma gesehen und gezeigt, dass 

er über ein nahezu perfektes Magnetfeld verfügt. Und 

2017 organisierten wir im Anschluss an die AMNT eine 

Exkursion nach Greifswald, um uns diesen W7X mal selbst 

anzusehen. Dass dabei Dr. Spannagel zu den Expeditionsteilnehmern 

gehörte, hat mich besonders gefreut. Vor Ort 

in Greifswald gab uns Prof. H.-S. Bosch einen umfassenden 

Einblick in die Besonderheiten des W7X, seine Inbetriebnahme, 

über die Ergebnisse der ersten Betriebsphase 

und die Pläne zum weiteren Projektverlauf. Anschließend 

erklärte uns Dr.-Ing. L. Wegener die Besonderheiten 

und Herausforderungen des W7X-Projekts hinsichtlich 

Konstruktion, Organisation und Projektmanagement. So 

waren wir schon vor der eigentlichen Führung beeindruckt 

und sensibilisiert für das, was wir anschließend auch aus 

der Nähe zu sehen bekamen. 

Neben den Aktionen und dem „Blick über den Tellerrand“ 

der Kernspaltungstechnik, den wir bieten, stellt 

unsere Fachgruppe aber auch einen Link dar zu anderen 

Fusions-orientierten Körperschaften wie den deutschen 

Fusionslaboren (IPP, KIT und FZJ), dem deutschen ITER 

Industrie Forum (dIIF), dem Europäischen Fusion Industry 

Innovation Forum (FIIF) und anderen. 

Und was wir in der Zukunft vorhaben? Schließen Sie 

sich unserer Fachgruppe an und wünschen Sie sich etwas! 

Dr. Thomas Mull 

Sprecher der KTG Fachgruppe Kernfusion 

KTG Inside


KTG-Sektion Ost: 

Exkursion 

Die KTG-Exkursion 2017 der Sektion Ost führte uns in das 

mitteldeutsche Braunkohlegebiet zur MIBRAG südlich von 

Leipzig. Die erste Station war das Braunkohlekraftwerk 

Deuben mit der angeschossenen Brikettfabrik. Das Kraftwerk 

stammt aus den 1930er Jahren. Der erste Eindruck 

des Kraftwerkskomplexes überraschte uns mit der gelungenen 

Architektur der erhaltenen Industrie gebäude in Ziegelbauweise. 

Nach dem Besuch des Leitstandes konnten 

wir im Kraftwerksgebäude in einen stillgelegten Braunkohle-Feuerungskessel 

einsteigen und erhielten anschaulich 

einen Einblick in die Funktionsweise und die technischen 

Herausforderungen des Kraftwerks. Beim anschließenden 

Rundgang durch den Generatorsaal erfuhren wir, 

dass ein Großteil der erzeugten Energie für die Großgeräte 

des angeschlossenen Tagebaus und für den Transport der 

Braunkohle mit Förderbändern und E-Loks benötigt wird. 

Leider konnte die geplante Besichtigung der Brikettfabrik 

wegen eines Stillstandes nicht stattfinden. 

freigesetzten elementaren Quecksilbers führte zum 

Auftragseingang. Heute werden unter anderem quecksilberhaltige 

Schlämme und Rückstände mit natürlicher 

Radioaktivität behandelt. Diese Rückstände in Form 

von Schlämmen entstehen beispielsweise bei der Erdgasförderung. 

Die Schlämme werden thermisch behandelt 

und dabei Quecksilber gewonnen, das dann hochrein 

vermarktet wird. Die Rückstände mit natürlicher Radioaktivität 

werden immobilisiert und auf spezielle Deponien 

verbracht. Bei einem Rundgang durch die Produktionshallen 

wurden uns anschaulich die Technologien beim 

Metallrecycling erläutert. 

Mit vielen neuen Eindrücken, die auf uns in den zwei 

Tagen einwirkten, haben wir dann die Heimreise angetreten. 

Besonderer Dank geht an die Mitarbeiter der beiden 

Firmen für die intensive und offene Betreuung während 

der Führungen. 

B. Standfuß et al. 

Zwischen Forschung, Rückbau 

und Entsorgung – aktuelle Aufgaben 

in der Kerntechnik 

125 

KTG INSIDE 

| | KTG-Sektion Ost: Exkursion 2017 

Im Tagebau Profen konnten wir uns von der Besucherplattform 

aus einen Überblick über die Ausmaße des 

Tagebaus verschaffen. Am Nachmittag fuhren wir dann 

zum Tagebau Schleenhain. Nach dem Besuch der Kaue und 

des Leitstandes des Tagebaues fuhren wir im Besucherbus 

im Tagebau direkt bis an die Schaufelradbagger, die Eimerkettenbagger 

und die kilometerlangen Bandanlagen. Es 

wurde erläutert, dass die Sanierung der Tagebauflächen 

nach der Verfüllung noch sechs Jahre vom Tagebauunternehmen 

durchgeführt wird. Mehrere Anpflanzungen und 

Fruchtfolgen garantieren, dass danach das Gelände wieder 

mit hoher Ackerzahl landwirtschaftlich ertragreich 

genutzt werden kann. In Gesprächen mit MIBRAG- 

Mitarbeitern wurde von diesen die technikfeindliche 

Einstellung von zunehmenden Teilen der Gesellschaft 

bedauert, die bei der Einstellung des Braunkohlentagebaus 

allein in Mitteldeutschland mehrere zehntausend 

Arbeitsplätze kosten würde. 

Mit einem geselligen Abend und Diskussionen, 

beendeten wir den sehr informativen Tag. 

Am nächsten besuchten wir die Gesellschaft für 

Metallrecycling Leipzig (GMR) in der Produktionsstätte 

Espenhain. Während eines informativen Einführungsvortrages 

erhielten wir einen Einblick in die Tätigkeitsfelder 

der Firma. Der Ursprung der Firma stammt aus einem 

Auftrag zur schadlosen Vernichtung von Munition für 

Sturmgewehre der ehemaligen NVA; insbesondere die 

Alleinstellung in der BRD mit der Fähigkeit zur Rückhaltung 

des bei der Verbrennung von Knallquecksilber 

Nachwuchstagung der Jungen Generation 

der KTG vom 8. – 10. November 2017 

Deutschland war über Jahrzehnte führend in der Entwicklung 

der Kerntechnik und dem sicheren und 

wirtschaft lichen Betrieb kerntechnischer Anlagen. Seit 

dem im Jahr 2011 beschlossenen beschleunigten Ausstieg 

aus der Kernenergienutzung ist die Hälfte der Zeit vergangen, 

bis das letzte deutsche Kernkraftwerk vom Netz 

genommen werden soll. 

Knapp 50 Teilnehmer waren der Einladung der Jungen 

Generation der KTG zur Nachwuchstagung nach Karlsruhe 

gefolgt. Der Campus Nord des Karlsruher Institut 

für Technologie (KIT), das frühere Forschungszentrum 

Karlsruhe, war seit den 1950er Jahren eine der Hauptstützen 

der kerntechnischen Entwicklung Deutschlands. 

Viele kerntechnische Forschungsrichtungen mit ihren 

Versuchs-, Pilot- und Forschungsanlagen, aber auch Einrichtungen 

der kerntechnischen Industrie waren hier 

beheimatet, einige sind es bis heute. Wie im restlichen 

Land stehen auch hier die Zeichen auf Rückbau – zum 

einen, weil einige Anlagen unterdessen das Ende ihrer 

Nutzungszeit erreicht haben, zum anderen aber auch, 

weil Rückbau, Entsorgung und Endlagerung wichtige 

Forschungsthemen sind. 

Als Teil der „Energiewende“ wird der anstehende 

Rückbau der Kernkraftwerke immer konkreter – Grund 

genug, sich direkt bei einem Elektroenergieerzeuger zu 

informieren, wie die Unternehmen damit umgehen. Die 

Teilnehmer konnten der Einladung in die EnBW-Zentrale in 

Karlsruhe folgen, um dort bei einem Get-together in 

entspannter Atmosphäre eine kurze Ansprache des 

Geschäftsführers der EnBW Kernkraft GmbH, Herrn Jörg 

Michels, zu hören. Seinen Worten zufolge hat EnBW den 

Rückbau auch seiner noch im Leistungsbetrieb befind lichen 

KKW zeitlich und monetär auskömmlich geplant. Er ermunterte 

die Teilnehmer ausdrücklich, optimistisch in die 

Zukunft zu sehen! Dieser Optimismus fußt auf mehreren 

Gründen: Der Rückbau der KKW wird nicht innerhalb einer 

Dekade abgeschlossen sein. Weiterhin erwirbt man in 

einem Rückbauprojekt Kompetenzen, die sich mühelos auf 

KTG Inside


126 

KTG INSIDE 

| | KONRAD-Container am Haken – 

Umlagerung im Zwischenlager 

der KTE auf dem Campus Nord 

| | Der Master-Slave-Manipulator – 

was so leicht aussieht, ist dann 

doch recht schwer... 

Projekte abseits der Kernenergie erzeugung anwenden 

lassen – das Lösen ingenieur technischer 

Anforderungen sowie enge Termin- und Kostenkontrolle 

erfordern alle Projekte, ob innerhalb 

oder außerhalb kerntechnischer Anwendungen! 

Schlussendlich erhält man innerhalb eines 

Rückbau projekts, welches verschiedenste Gewerke 

und Industriezweige 

mit- und nebeneinander tätig werden lässt, 

einen hohen Grad an Vernetzung mit verschiedenen 

Branchen. 

Der folgende Tag begann früh. Der Bus 

brachte die Teilnehmer vom Tagungshotel zum 

KIT Campus Nord. Nach einer Vorstellung des 

KIT, erfuhren wir, an welchen Stellen der 

Rückbau hinsichtlich eingesetzter Technik 

nicht nur Handwerk ist, sondern durchaus 

auch Aufgaben für die ingenieurtechnische 

Wissenschaft bereithält. Nach der Vorstellung 

der aktuellen und früheren Aufgaben des Instituts 

für Nukleare Entsorgung (INE), wo im 

Rahmen gesellschaftlicher Vorsorgeforschung 

grundlegende und anwendungsorientierte 

FuE-Arbeiten zur sicheren Ent sorgung radioaktiver 

Abfälle durchgeführt sowie Fragestellungen 

zum Rückbau kerntechnischer Anlagen 

thematisiert werden, starteten Besichtigungen. 

Im INE-Kontrollbereich wurden Details zu 

endlagerungsvorbereitenden Untersuchungen 

an Brenn elementen, zur Aktiniden forschung und über 

Möglichkeiten der Laserspek troskopie vorgestellt. 

An den INE-Beamlines der Synchrotron Radiation 

Facility „KARA“ erfuhren wir Details zu den Möglichkeiten 

und Anwendungsgebieten der Bildgebung mittels 

Röntgenstrahlung. 

Der Nachmittag gehörte ausführlichen Besichtigungen 

an den Anlagen der Kerntechnische Entsorgung Karlsruhe 

GmbH (KTE) am KIT Campus Nord – Zwischenlager, Wiederaufarbeitungsanlage 

(WAK) und Mehrzweckforschungsreaktor 

(MZFR). 

Das Zwischenlager beeindruckte durch seine Dimensionen. 

Interessant zu sehen, mit welchen Untersuchungsmethoden 

Reststoffe nach Eingang kontrolliert und 

qualifiziert werden. Die angewendeten Verfahren kommen 

auch bei der Nachqualifizierung älterer Reststoffe zum 

Einsatz. Parallel wird ein hoher Aufwand bei der Pflege der 

älteren Gebinde und der Konditionierung von Gebinden 

für das Endlager KONRAD betrieben. 

In der WAK erhielten wir Einblick in die Vorbereitungen 

des fernhantierten Rückbaus der Bereiche, die für einen 

manuellen Rückbau nicht zugänglich sind. Die Ortsdosisleistung 

ist insbesondere in der Verglasungsanlage so hoch, dass 

auch technische Geräte nach begrenzter Einsatzdauer beeinträchtigt 

werden bzw. versagen. Teils müssen hier zur 

Steuerung der Rückbauwerkzeuge Techniken und Verfahren 

etabliert werden, die in der Form bisher noch nirgends 

zum Einsatz kamen. 

Am MZFR konnten wir ein Kernkraftwerk in seinen 

„späten Jahren“ erleben. Die Führung brachte uns zu 

vielen interessanten Orten innerhalb dieses im Wesentlichen 

bis auf die Ge bäudestruktur entkernten Gebäudes. 

Neben letzten Rückbauarbeiten ist man dort mit dem 

messtech nischen Nachweis der Freigabefähigkeit, die zur 

Freigabe des Gebäudes gemäß § 29 StrlSchV führen soll, 

befasst. Interessant, wie anspruchsvoll auch oder gerade 

solche letzten Schritte sind, wo nicht mehr der Schutz der 

Person vor der Direktstrahlung, sondern der Nachweis der 

Kontaminations freiheit im Vordergrund steht. 

Der Tag wurde mit einem gemütlichen Beisammensein 

bei Speis und Trank abge rundet. Dabei waren Zeit und 

Gelegenheit, neue Kontakte zu knüpfen oder bestehende 

Kontakte zu vertiefen. 

Auch der letzte Tagungstag begann früh. Nach den 

intensiven Eindrücken des Vortags zu Rückbau und 

nuklearer Reststoffwirtschaft stand nun der wirtschaftliche 

und politische Rahmen des Rückbaus im Fokus. 

Zuerst wurde die Rolle der EU hinsichtlich wissenschaftlicher 

und politischer Unterstützung thematisiert. 

Zwei weitere Vorträge zeigten an praktischen 

Beispielen, was in Kernkraftwerken nach der Abschaltung 

passiert. Anhand der Kernkraftwerke Philippsburg und 

Neckarwestheim wurde gezeigt, wie das Management von 

Reststoffen vom Rückbau über den Transport bis hin zur 

Rezyklierung ineinandergreift – einfach gesagt: „Was 

passiert mit einem Kernkraftwerk nach der Abschaltung?“. 

Im Folgenden wurde berichtet, wie in den Blöcken A 

und B des Kernkraftwerks Biblis in Umsetzung erteilter 

Stilllegungs- und Rückbaugenehmigungen erste Abbaumaßnahmen 

durchgeführt werden. Geplant ist hier, auch 

im Unterschied zu den Anlagen in Philippsburg und 

Neckarwestheim, die Abbau- und Reststoffbearbeitungstätigkeiten 

innerhalb der bestehenden Gebäude durchzuführen. 

| | Tagungsteilnehmer auf Besichtigungstour am Institut für Technische Physik 

Im Anschluss wurde die Kostenschätzung von Stilllegungs- 

und Rückbaumaßnahmen thematisiert. Wichtig 

dabei ist, dass die Gesamtkosten mindestens zutreffend, 

jedoch keinesfalls zu niedrig geschätzt werden. Diese Verpflichtung 

ergibt sich nicht zuletzt aus den Bestimmungen 

zur Entsorgung von Kernkraftwerken bzw. -anlagen. 

Zugleich sind steuerrechtliche Vorgaben zu beachten, da 

Rückstellungen den steuerpflichtigen Gewinn mindern. 

Nicht zuletzt vor dem Hintergrund steigender Preise und 

teils unsicherer gesetzlicher Rahmenbedingungen haben 

die Betreiber ein vitales Interesse, dass die Gesamtkosten 

des Rückbaus ausreichend abgeschätzt werden. 

Den Abschluss des Vortragsteils am Vormittag bildeten 

Einblicke in die automatisierte Zerlegung von RDB- 

Einbauten mittels Unterwasser-Robotertechnik. Von der 

Ertüchtigung des Basisgeräts zur Unterwasserfähigkeit 

über die Erarbeitung eines Interventionskonzepts, der Entwicklung 

eines „Masterarms“ für die Werkzeugaufnahme, 

der Entwicklung eines Werkzeugwechselsystems bis zur 

Ausarbeitung von Schutzmechanismen ist dabei ein breites 

Spektrum von Herausforderungen zu bestehen, bevor der 

erste Einsatz stattfinden kann. 

KTG Inside


Gestärkt vom Mittagessen wurde die Tagung am 

Nachmittag mit der Besichtigung der Spultestanlage 

TOSKA des Instituts für Technische Physik (ITEP) am KIT, 

einer Anlage, in der große supraleitende Magnete für die 

Fusion getestet werden, sehr erfolgreich beendet. 

Dank an dieser Stelle allen Vortragenden und Organisatoren 

des KIT und KTE für die sehr guten Führungen und 

die perfekte Organisation des Besichtigungsnachmittags 

als auch allen weiteren Vortragenden aus der Industrie! 

Unser Dank gilt weiterhin allen Organisatoren, die 

erhebliche Teile ihrer Freizeit für das Zustandekommen 

und die Ausgestaltung der Tagung geopfert haben. Weiterhin 

danken wir unseren Arbeitgebern, Helfern sowie 

direkten und indirekten Sponsoren und Unterstützern. 

Ohne ihr Wirken hätte die Tagung nicht zu einem Erfolg 

werden können. 

Sven Jansen 

Im Namen des Vorstands der Jungen Generation der KTG 

MINT pink: WiN dabei 

Am 20.11.2017 fand im Körber-Forum in Hamburg der 

Programmabschluss von „MINT pink“ statt. MINT pink ist 

ein schulübergreifendes Programm, das ausgewählte 

Schülerinnen der Mittelstufe für die Wahl eines naturwissenschaftlichen 

Profils in der Oberstufe ermutigt 

und Studien-, Arbeits- und Karrieremöglichkeiten 

im Mathe matik- Informatik-Naturwissenschaft-Technik- 

Bereich auf zeigt. 

| | MINT pink: WiN dabei. Chantal Greul stellt ihren Beruf in der Kerntechnik 

Schülerinnen vor. 

Chantal Greul durfte als Role Model über 90 Mädchen 

den Beruf der Ingenieurin in der Kerntechnik vorstellen. 

Nach einer kurzen Vorstellung des eigenen Lebenslaufes 

und des Arbeitsalltages in einer kerntechnischen Anlage, 

durften die Schülerinnen in kleinen Gesprächsrunden 

ihre Fragen stellen. Diese reichten von allgemeinen Fragen 

bis hin zu spezifischen Fachfragen rund um Kernenergie, 

Rückbau und Endlagerung in Deutschland. Die Veranstaltung 

war eine interessante Gelegenheit mit potenziellem 

Nachwuchs in Kontakt zu kommen und sie über 

die spannende Arbeit in der Kernenergiebranche zu 

informieren. Auch die Programmauswertung zeigt den 

Erfolg des MINT-pink-Programmes. Vor Programmstart 

konnten sich 27 % der Mädchen vorstellen, das Physikoder 

Chemieprofil in der Oberstufe zu wählen. Nach 

Programmende waren es 45 %. 

KTG Inside 

Verantwortlich 

für den Inhalt: 

Die Autoren. 

Lektorat: 

Sibille Wingens, 

Kerntechnische 

Gesellschaft e. V. 

(KTG) 


10115 Berlin 

T: +49 30 498555-50 

F: +49 30 498555-51 

E-Mail: s.wingens@ 

ktg.org 

127 

KTG INSIDE 

Yvonne Broy 

www.ktg.org 

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KTG Inside


Top 

IAEA: Solving the back end: 

Finland’s key to the final 

disposal of spent nuclear fuel 

(iaea) Countries operating nuclear 

power plants store their spent nuclear 

fuel either at reactor sites or away 

from them. Spent fuel can be dangerous 

to people and the environment 

if not properly managed; therefore, 

a publicly acceptable, permanent 

solution for its disposal is needed. 

While a number of countries are 

considering deep geological disposal 

repositories, Finland is the only 

country that has begun the construction 

of a repository for the final 

disposal of its spent nuclear fuel. 

At a depth of 400 to 450 metres 

and with about 70 km of tunnels and 

shafts, the ONKALO repository in 

Olkiluoto on Finland’s west coast 

will house copper canisters filled 

with spent fuel from nuclear power 

reactors. It is expected to receive 

waste for about 100 years, after which 

time it will be sealed. 

“Since the decision was made 

40 years ago on the overall waste 

management strategy and on a deep 

geological repository as the primary 

option for spent nuclear fuel, all the 

stakeholders have stood by it,” said 

Tiina Jalonen, Senior Vice President 

for Development at Posiva, the company 

in charge of the project. “Governments 

and people have changed, 

but the decision and the vision for the 

future have remained the same.” 

Another reason why Finland’s 

model has worked is the timely 

involvement of all the stakeholders in 

the project, who worked as one team, 

targeting the same goal. 

“The roles between the different 

stakeholders have been clear. The 

decision makers have developed 

legislation in parallel to introducing 

nuclear energy, and the Radiation and 

Nuclear Safety Authority of Finland 

(STUK) has developed safety guides, 

regulations and competences to 

review and inspect our documentation 

and applications,” said Jalonen. 

Moreover, involving STUK from the 

beginning was crucial to building the 

trust in the project. “It wouldn’t have 

worked if any of the stakeholders were 

missing from the process,” explained 

Petteri Tiippana, Director General at 

STUK. “Active participation of the safety 

regulator provided the local community 

with additional assurances.” 

In fact, public acceptance was 

crucial for the success of the project. 

The selection of the Olkiluoto site 

–home to three nuclear reactors – as 

the repository site was made, not only 

for the geological suitability of this 

area, but also for the acceptance of the 

people living there. Finland conducted 

many studies about local and 

national attitudes toward the project, 

which showed that people living 

around nuclear power plants tend to 

have more trust in nuclear projects. 

“Trust has been one cornerstone 

in being able to proceed according to 

the Government’s schedule,” Jalonen 

said. “Building trust has required 

extensive and open communication 

with local people, the authority and 

the decision makers.” 

The project is based on the “multiple 

barriers” concept, which aims to 

provide needed containment and 

isolation to prevent spent fuel from 

leaking and spreading, according to 

Posiva. The combination of bedrock, 

disposal canisters surrounded by clay, 

tunnels filled with clay containing 

backfilling materials and plugging the 

tunnel’s mouth will all serve as protective 

multiple barriers. 

Who’s next? 

Two other countries have made progress 

towards building repositories for 

high-level radioactive waste or spent 

fuel declared as waste. In June 2016, 

the Swedish Radiation Safety Authority 

endorsed the licence application 

for the future spent fuel deep geological 

repository at Forsmark. Review by 

the Swedish Land and Environment 

Court for environmental licencing of 

the project started in September 2017. 

In France, the licence application 

for the deep geological disposal 

facility, Cigéo, is under preparation; it 

is planned to be submitted by the end 

of 2018, with construction starting in 

2020. The pilot phase of disposal 

could start as soon as 2025. It will 

contain waste from the reprocessing 

of spent fuel from France’s current 

fleet of nuclear power plants and 

other long-lived radioactive waste. 

The science 

High-Level Radioactive Waste (HLW) 

is produced from the burning of 

uranium fuel in nuclear power reactors. 

It is of two kinds: spent fuel, 

declared as waste and ready for 

disposal, or waste resulting from the 

reprocessing of spent fuel. 

Due to its high radioactivity and 

very long half-life (the time it takes 

for a radioactive substance to lose half 

its radioactivity), HLW has to be well 

contained and isolated from the human 

environment. Intensive research 

has identified the suitability of various 

rock types to host deep geological repositories 

and engineered barrier systems 

to isolate the waste. These repositories 

are constructed in suitable geological 

formations at a depth of several 

hundred meters and designed to 

contain high-level waste for hundreds 

of thousands of years. 

| | www.iaea.org 

Reactors 

Georgia’s commitment to 

new nuclear a win for US 

economy, environment 

(nei) In response to the announcement 

that the Georgia Public Service Commission 

unanimously approved an 

order allowing continued construction 

of two additional reactors at Plant 

Vogtle, the following is a statement by 

NEI President and CEO Maria Korsnick. 

“Completing the Plant Vogtle expansion 

is good for America on many 

levels, especially in terms of our 

national security, our commitment to a 

cleaner environment, and energy 

diversity. In addition to the thousands 

of workers who will cheer this decision, 

these nuclear facilities when 

completed will produce decades worth 

of clean, reliable power and provide 

billions of dollars in economic benefits. 

“Demonstrating we can build and 

complete new nuclear plants here in 

America will help us regain our 

leader ship in a technology we invented. 

America’s pre-eminence in 

nuclear energy makes our country 

safer because it allows us to influence 

and control how this technology is 

used around the world.” 

| | www.nei.com 

Finnish cities to explore Small 

Modular Reactors for district 

heating 

(nucnet) The Finnish cities of Helsinki, 

Espoo and Kirkkonummi have begun 

studies to find out if it would be 

feasible to replace coal and natural 

gas in district heating with small 

modular nuclear reactors (SMRs), the 

environmental group Ecomodernist 

Society of Finland said. The society 

said a feasibility study will be carried 

out into the potential for SMRs to 

replace fossil fuel-burning in cities 

around the Helsinki metropolitan 

area. Several advanced SMRs are in 

development and coming to market by 

2030 that could meet the specifications, 

the society said. Most of the 

district heating in Finland is produced 

129 

NEWS 

News


130 

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 

by burning coal, natural gas, wood 

fuels and peat. While many Finnish 

cities have progressive climate policies 

and goals, they have struggled to 

decarbonise heating and liquid fuels, 

the society said. Rauli Partanen, 

vice-chair of the society and an independent 

energy analyst and author, 

said there are “significant economic 

possibilities” in producing combined 

heat and power (CHP) with nuclear 

reactors. He said: “With CHP, the 

reactor could produce roughly twice 

the value per installed capacity compared 

with just electricity production, 

while at the same time decarbonising 

heat production.” He said nuclear 

is great for baseload needs, but 

with advanced technologies such as 

high temperature reactors and high 

temperature electrolysis, nuclear can 

also be used to decarbonise not just 

electricity, heat but also transportation 

fuels and many industries”. 

| | www.vtt.fi 

EDF ‘Cannot build new 

reactors in France without 

guarantees’ 

(nucnet) French state-controlled 

utility EDF can no longer build new 

nuclear reactors in France without 

state support, chief executive officer 

Jean-Bernard Levy was quoted as 

saying in an interview with the Ouest 

France daily newspaper. Asked when 

EDF could build new reactors at home, 

Mr Levy said: “Henceforth, we cannot 

build new reactors without adequate 

regulation providing guaranteed 

income”. He said the Flamanville-3 

EPR project under construction in 

northern France began at a time of 

high power prices and that now all 

power sources, nuclear as well as 

renewables, need to get “the same 

visibility on sales prices”. For its 

project to build two EPRs at Hinkley 

Point in the UK, EDF obtained an 

EU-approved state-guaranteed price 

of £92.5 per MWh over 35 years, 

which is above current market prices. 

The government of French president 

Emmanuel Macron is planning to close 

old reactors to reduce the share of nuclear 

energy in French power generation 

to 50% by around 2035 from 75 % 

today. Mr Levy said EDF expects to get 

approval to load nuclear fuel at 

Flamanville-3 at the end of 2018. 

| | www.edf.com 

Japan’s Regulator: 

Kashiwazaki Kariwa-6 and -7 

meet new safety standards 

(nucnet) Units 6 and 7 of the 

Kashiwazaki Kariwa nuclear power 

station in Niigata Prefecture, northwestern 

Japan, meet new regulatory 

standards imposed after the March 

2011 Fukushima-Daiichi accident, the 

Nuclear Regulation Authority said. 

The two units, owned and operated 

by Tokyo Electric Power Company 

(Tepco) are the first boiling water 

reactors to meet the new standards. 

Tepco also owns the Fukushima- 

Daiichi station. 

Kashiwazaki Kariwa was not affected 

by the March 2011 earthquake and 

tsunami which damaged Fukushima- 

Daiichi, although the station’s seven 

reactors had all been offline for up to 

three years following a 2007 earthquake 

which damaged the site but did 

not damage the reactors themselves. 

While the units were offline, work 

was carried out to improve the 

facility’s earthquake resistance. 

Accord ing to JAIF, the governor of 

Niigata Prefecture, Ryuichi Yoneyama, 

has said he will not discuss restarting 

the two units until further information 

about nuclear incidents and their 

impact on public health is made available. 

Both units are 1,315-MW BWRS. 

Kashiwazaki Kariwa-6 began commercial 

operation in November 1996 and 

Kashiwazaki Kariwa-7 in July 1997. 

Tokyo-based nuclear industry 

group the Japan Atomic Industrial 

Forum said 14 nuclear units have now 

been approved by the NRA as meeting 

the new standards. They are 

Kashiwazaki Kariwa-6 and -7, 

Operating Results October 2017 (corrigendum, atw 1 (2018) p. 58) 

Plant name 

Type 

Nominal 

capacity 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 

Energy generated, gross 

[MWh] 

Time availability 

[%] 

Energy availability 

[%] *) Energy utilisation 

[%] *) 

Month Year Since Month Year Month Year Month Year 

commissioning 

KBR Brokdorf DWR 1480 1410 745 937 223 3 903 011 338 316 924 100.00 41.98 93.94 39.11 84.59 35.98 

KKE Emsland 4) DWR 1406 1335 745 1 004 762 9 304 398 333 303 977 100.00 91.93 99.93 91.77 95.81 90.70 

KWG Grohnde DWR 1430 1360 745 970 799 8 126 396 365 069 095 100.00 87.01 94.85 83.35 90.42 77.21 

KRB B Gundremmingen 4) SWR 1344 1284 745 778 570 8 351 414 330 004 358 100.00 91.83 100.00 90.98 76.78 84.52 

KRB C Gundremmingen SWR 1344 1288 745 968 428 7 990 831 318 640 904 100.00 85.41 99.83 83.30 96.32 81.02 

KKI-2 Isar DWR 1485 1410 745 1 073 129 9 378 353 339 453 163 100.00 89.84 99.71 89.37 96.66 86.22 

KKP-2 Philippsburg DWR 1468 1402 745 1 046 248 5 745 846 353 059 535 100.00 55.80 99.92 55.72 94.15 52.80 

GKN-II Neckarwestheim DWR 1400 1310 745 1 011 300 8 549 400 318 131 734 100.00 86.71 99.50 86.46 97.13 83.84 

Operating Results November 2017 

Plant name 

Type 

Nominal 

capacity 

gross 

[MW] 

net 

[MW] 

Operating 

time 

generator 

[h] 

Energy generated, gross 

[MWh] 

Time availability 

[%] 

Energy availability Energy utilisation 

[%] *) [%] *) 

Month Year Since Month Year Month Year Month Year 

commissioning 

KBR Brokdorf DWR 1480 1410 720 942 685 4 845 695 339 259 608 100.00 47.20 93.97 44.04 88.16 40.67 

KKE Emsland 4) DWR 1406 1335 720 1 017 448 10 321 846 334 321 425 100.00 92.65 99.77 92.49 100.61 91.59 

KWG Grohnde DWR 1430 1360 446 586 675 8 713 070 365 655 769 61.99 84.77 56.84 80.97 56.57 75.35 

KRB B Gundremmingen 4) SWR 1344 1284 720 701 347 9 052 761 330 705 705 100.00 92.56 98.85 91.69 71.32 83.33 

KRB C Gundremmingen SWR 1344 1288 720 956 516 8 947 347 319 597 420 100.00 86.72 98.05 84.62 98.30 82.57 

KKI-2 Isar DWR 1485 1410 720 1 061 544 10 439 897 340 514 707 100.00 90.75 100.00 90.33 99.05 87.37 

KKP-2 Philippsburg DWR 1468 1402 720 1 042 562 6 788 408 354 102 097 100.00 59.77 100.00 59.69 97.08 56.77 

GKN-II Neckarwestheim DWR 1400 1310 720 996 000 9 545 400 319 127 734 100.00 87.91 98.51 87.54 99.08 85.21 

News


Mihama 3, Takahama-1, -2, -3 and -4, 

and Ohi-3 and -4, Ikata-3, Genkai-3 

and -4 and Sendai-1 and -2. 

All of Japan’s 48 reactors were shut 

between 2011 and 2012 after the 

Fukushima-Daiichi accident. Five 

units have resumed commercial operation. 

They are: Takahama-3 and -4, 

Ikata-3 and Sendai-1 and -2. 

Before the Fukushima-Daiichi 

accident Japan had generated around 

30% of its electricity with plans to 

increase the share to 40%. According 

to the International Atomic Energy 

Agency Japan’s nuclear share in 2016 

was 2.15%. 

| | www.tepco.co.jp, www.nsr.go.jp 

Slovakia: Mochovce-3 startup 

target of end 2018 is realistic 

(se) The schedule for the completion 

of the third and fourth units of the 

Mochovce nuclear power station in 

Slovakia is realistic, with Unit 3 likely 

to begin commercial operation at the 

end of 2018 and Unit 4 at the end of 

2019, regulator UJD said. According 

to utility Slovenské Elektrárne, fuel 

will be loaded into Unit 3 in July 2018. 

Preparations have begun for the start 

of a cold pressure test of the primary 

circuit at Unit 3, local media reports 

said. In June 2016 the utility said construction 

work at Unit 3 was “more 

than 92%” finished, with Unit 4 at 

75%. Mochovce-3 and -4 are both 

440-MW pressurised water reactors of 

the Russian VVER V-213 design. 

| | www.seas.sk 

Plans for UK new nuclear 

move forward as regulator 

approves design for UK-ABWR 

(nucnet) Plans for two new nuclear 

power stations in the UK have taken a 

crucial step forward as UK regulators 

approved the design of the reactor 

technology for the projects. The Office 

for Nuclear Regulation gave the green 

light today for the UK Advanced 

Boiling Water Reactor (UK-ABWR), 

designed by Hitachi-GE. The ONR 

said the design is suitable for construction 

in the UK, marking the end 

of a five-year regulatory process. 

Horizon Nuclear Power is proposing 

to build and operate two of these 

reactors in Wylfa Newydd on Anglesey 

and Oldbury-on-Severn in Gloucestershire. 

Duncan Hawthorne, Horizon’s 

chief executive, said: “This is a huge 

milestone for Horizon and a major 

leap forward for us in bringing 

much-needed new nuclear power to 

the UK.” Horizon said today that 

“steady progress” is being made with 

the Hitachi-backed Wylfa Newydd 

project, including the submission of 

the site licence application and completion 

of a third public consultation. 

Attention will now turn to financing 

the Wylfa Newydd project. Earlier this 

year Horizon said: “We have always 

been clear that we are looking to bring 

other investors into Horizon. Based on 

the strengths of our project, we are in 

positive discussions with a number of 

parties but we will not be commenting 

on the process whilst it is ongoing.” 

| | www.onr.ork.uk, 

www.hitachi-hgne-uk-abwr.co.uk 

Company News 

Framatome pursues the 

industrial and technological 

adventure of the nuclear 

energy business 

(framatome) New NP, a subsidiary of 

AREVA NP, becomes Framatome, a 

company whose capital is owned by 

the EDF group (75.5%), Mitsubishi 

Heavy Industries (MHI – 19.5%) and 

Assystem (5 %). 

Framatome confirms its recognized 

manufacturer’s ambition: being the 

sup plier of safe and competitive nuclear 

solutions, supporting its electrical 

utility customers all over the world. 

Framatome, 14,000 employees 

worldwide 

Framatome employees have recognized 

skills, a know-how that was 

forged over the long history of the 

company and that has enabled us to 

build outstanding industrial success in 

France and internationally. Framatome 

places its faith in the expertise 

of the women and the men who are at 

its very core: this expertise underpins 

the company’s strategy and is key to 

serving the needs of its customers and 

furthering the success of the nuclear 

industry. 

In the words of Bernard Fontana, 

Chairman of the Managing Board and 

Chief Executive Officer of Framatome, 

“Framatome possesses unique knowhow 

in an industry that today is and 

will remain key for a low-carbon 

energy mix. Our employees in France 

and around the world have been able 

to face considerable challenges in 

recent years. As we emerge from this 

transition phase, I share their pride 

and I want to thank them for all 

the work they have accomplished. 

Steeped in a rich heritage, Framatome 

is today one of the reference players in 

the nuclear sector worldwide, benefiting 

from unparalleled operating 

feedback. Our ambition is delivering a 

level of industrial excellence that is 

recognized by our customers.” 

Proud of its core business expertise 

as designer, supplier and installer of 

nuclear steam supply systems Framatome 

contributes to the design of 

power plants, supplies the nuclear 

steam supply system, designs and 

manufactures components and fuels, 

integrates the instrumentation and 

control systems and carries out the 

maintenance of in-service nuclear 

reactors. It delivers its high-performance 

products and services to 

customers all over the world. 

Framatome is a technology company, 

holding around 3,500 patents 

covering some 680 inventions, which 

serve the most demanding needs of its 

customers who number among the 

key international energy leaders. 

Framatome operates on more than 

250 reactors worldwide. 

An internationally-focused strategy 

of development and industrial excellence 

Framatome has the determination 

to go further in terms of industrial 

excellence, leveraging five strategic 

axes: proven and sustainable expertise, 

performance in delivering, an 

agile and adaptive organization, safe 

and competitive solutions and international 

development. With an existing 

global fleet of some 440 reactors 

representing output of around 

390 GWe in 31 countries, and with 

new nuclear capacity on its way, the 

nuclear market presents opportunities 

in the areas of components, fuel, retrofits 

and services. (18191512) 

| | www.framatome.com 

Brookfield to acquire Westinghouse 

Electric Company 

(westn) Westinghouse Electric Company, 

the global leader in nuclear 

technology, fuel and services, has 

agreed to be acquired by Brookfield 

Business Partners L.P. (NYSE:BBU) 

(TSX:BBU.UN) together with institutional 

partners (collectively, “Brookfield”) 

for approximately $ 4.6 billion. 

The purchase price for substan tially 

all of the global business of Westinghouse 

Electric Company LLC and its 

affiliated debtors and debtors- in-posses 

sion (collectively “Westinghouse”) 

excludes cash, but includes the assumption 

of certain pension, environmental 

and other operating obligations. 

“Brookfield’s acquisition of Westinghouse 

reaffirms our position as the 

leader of the global nuclear industry,” 

said Westinghouse President & CEO 

José Emeterio Gutiérrez. “Our transformation 

and strategic restructuring 

131 

NEWS 

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132 

NEWS 

process is creating a stronger, stable, 

and more streamlined global Westinghouse 

business, for the benefit of our 

customers and employees.” 

Brookfield’s acquisition of Westinghouse 

is expected to close in the third 

quarter of 2018, subject to Bankruptcy 

Court approval and customary closing 

conditions including, among others, 

regulatory approvals. Throughout the 

process, Westinghouse will continue 

to operate in the ordinary course of 

business under its existing senior 

management. 

PJT Partners is the financial advisor 

to Westinghouse, Weil, Gotshal & 

Manges LLP is Westinghouse’s legal 

counsel, and AlixPartners LLP is Westinghouse’s 

turnaround consultant. 

| | www.westinghousenuclear.com, 

www.brookfield.com 

People 

Appointment of the 

Framatome Managing Board 

(framatome) The Supervisory Board 

of Framatome, meeting under the 

chairmanship of Jean-Bernard Lévy, 

Chairman and CEO of EDF, appointed 

Some Questions and Answers About Energy. 

Answers 

1b. False: All leading scenarios predict a rise of the global 

energy demand for the next decades (2015 to 2040: 

between 10 % to 40 %) mainly driven by the increase 

of the population and the growing demand in developing 

countries. 

2b. False: All leading scenarios predict an over proportional rise 

of the global electricity demand for the next decades (2015 

to 2040: between 60 % to 80 %) mainly driven by the 

increase of the population, the growing demand in 

developing countries and the today’s poor access to 

electricity for about one third of the world’s population. 

3b. False: In 2017 the global coal production increased by 2 % 

compared with the previous year 2016. 

4c. Since 2010 about 11 % of world’s electricity demand is 

produced in nuclear power plants. 

5b. About 5 % of world’s electricity demand was produced by 

wind (4 %) and solar (1 %) in 2017. 

6c. United States, with about 6,800 billions of tonnes, 

98 % thereof coal; EU about 530 billions of tonnes, 

95 % thereof coal 

7d. China. The carbon dioxide emission are always twice 

the emissions of the USA and three times the emissions 

of all 28 EU countries. 

8d. Hydropower, 4 to 13 g CO 2 per kWh. 

Wind and nuclear: about 8 to 20 g CO 2 per kWh. 

Photovoltaics: 35 to 160 g CO 2 per kWh. 

9a. Nuclear power, especially small modular reactors 

with advanced fuel usage. 

10d. Nuclear power. The number of lost lifetime-days per 

kilowatt-hour produced from nuclear power is in the range 

of wind power and about 5- to 100-times lower than of 

every other primary energy source. 

11d. The natural radiation caused by Thorium and its decay 

products in Guarapari (Monazit area) is up to 

10,000-times higher than the radiation from nuclear 

reactors in normal operation. 

12b. False: There are 448 nuclear power plants in operation 

worldwide and 59 under construction. About 120 additional 

power plants are planned. Only some plants will be shutdown 

in the upcoming year, mainly in the „old“ countries. 

Further expansion programmes are under the way e.g. in 

China with more than 100 plants to be in operation in the 

period 2030 to 2040 and the „Newcomer“ countries in Asia. 

Bernard Fontana Chairman of the 

Managing Board and Chief Executive 

Officer. 

It also appointed Philippe Braidy 

Managing Director, member of the 

Managing Board. 

Bernard Fontana holds a degree in 

engineering from the École Polytechnique 

and the École Nationale 

Supérieure des Techniques Avancées 

in Paris. He has 30 years’ experience 

in the chemical, steel and building 

materials industries (SNPE, Arcelor- 

Mittal, APERAM and Holcim). 

From February 2012 to September 

2015, he served as CEO of Holcim Ltd. 

Since September 1, 2015, Bernard 

Fontana had been Chief Executive 

Officer of AREVA NP. 

Philippe Braidy, former Head of 

regional and local Development and 

network in French Caisse des Dépôts, 

has 30 years’ experience as Technical 

and Financial Director in public 

administrations (French Ministry 

of Budget, Prime minister’s office, 

CEA…). Up to now he has been 

managing the Finance, Strategy/Innovation/Communications, 

Legal/Compliance, 

Risks/Audit, and Information 

Systems Functions of AREVA NP. 

| | www.framatome.com 

Einige Fragen und Antworten zum Thema Energie. 

Die Antworten 

1b. Falsch: Alle führenden Szenarien prognostizieren einen Anstieg 

des globalen Energiebedarfs für die nächsten Jahrzehnte (2015 

bis 2040: zwischen 10 % und 40 %), der vor allem durch das 

Bevölkerungswachstum und die wachsende Nachfrage an 

Energie in den sich entwickelnden Ländern getrieben wird. 

2b. Falsch: Alle führenden Szenarien prognostizieren für die nächsten 

Jahrzehnte einen überpropor tio nalen Anstieg des weltweiten 

Strombedarfs (2015 bis 2040: zwischen 60 % und 80 %), der 

vor allem durch das Bevölkerungswachstum, die wachsende 

Nachfrage in den sich entwickelnden Ländern und dem heute 

fehlenden Zugang zu Elektrizität für etwa ein Drittel der Weltbevölkerung 

bedingt ist. 

3b. Falsch: Im Jahr 2017 stieg die weltweite Kohle förderung 

im Vergleich zum Vorjahr 2016 um 2 %. 

4c. Seit 2010 werden rund 11 % des weltweiten Strombedarfs 

in Kernkraftwerken erzeugt. 

5b. Rund 5 % des weltweiten Strombedarfs wurden 2017 

durch Wind (4 %) und Solarenergie (1 %) erzeugt. 

6c. USA mit rund 6.800 Mrd. t, davon 98 % Kohle; 

EU mit rund 530 Mrd. t, davon 95 % Kohle 

7d. China. Die Kohlendioxid-Emissionen sind doppelt so hoch wie 

die der USA und dreimal so hoch wie die aller 28 EU-Länder. 

8d. Wasserkraft, 4 bis 13 g CO 2 pro kWh. 

Wind und Atomkraft: ca. 8 bis 20 g CO 2 pro kWh. 

Photovoltaik: 35 bis 160 g CO 2 pro kWh. 

9a. Kernkraft, insbesondere kleine modulare Reaktoren 

mit fortschrittlichem Brennstoff. 

10d. Kernenergie. Die Anzahl der Ausfalltage pro Kilowatt stunde 

aus Kernenergie liegt im Bereich der Windkraft und 

etwa 5- bis 100-mal niedriger als bei jeder anderen 

Primärenergiequelle. 

11d. Die natürliche Strahlung, die Thorium und seine Zerfalls produkte 

in Guarapari (Monazit-Gebiet) verursachen, ist bis zu 

10.000-mal höher als die Strahlung aus Kernkraftwerken 

im Normalbetrieb. 

12b. Falsch: Weltweit sind 448 Kernkraftwerke in Betrieb und 

59 in Bau; rund 120 weitere Kraftwerke sind geplant. 

Nur einige Anlagen werden in den kommenden Jahren 

stillgelegt werden, vor allem in den “alten” Ländern. 

Weitere Ausbauprogramme werden verfolgt und umgesetzt, 

z.B. in China mit mehr als 100 Anlagen, die im Zeitraum 

2030 bis 2040 in Betrieb sein werden, sowie in den 

“Newcomer”-Ländern Asiens. 

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 2013 

• Uranium: 34.00–43.50 

• Conversion: 9.25–11.50 

• Separative work: 98.00–127.00 

January to December 2014 

• Uranium: 28.10–42.00 



January to June 2015 

• Uranium: 35.00–39.75 



June to December 2015 

• Uranium: 35.00–37.45 



2016 

January to June 2016 

• Uranium: 26.50–35.25 



July to December 2016 

• Uranium: 18.75–27.80 



2017 

January 2017 

• Uranium: 20.25–25.50 



February 2017 

• Uranium: 23.50–26.50 



March 2017 

• Uranium: 24.00–26.00 



April 2017 

• Uranium: 22.50–23.50 



May 2017 

• Uranium: 19.25–22.75 



June 2017 

• Uranium: 19.25–20.50 



News


July 2017 

• Uranium: 19.75–20.50 



August 2017 

• Uranium: 19.50–21.00 



September 2017 

• Uranium: 19.75–20.75 



October 2017 

• Uranium: 19.90–20.50 



November 2017 

• Uranium: 20.00–26.00 



December 2017 

• Uranium: 23.50–25.50 



| | 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 

I. quarter: 56.87; 8,627,347 

II. quarter: 56.12; 5,970,240 

III. quarter: 65.03, 7.257.041 

IV. quarter: 88.28; 7,932,550 

2017: 

I. quarter: 95.75; 8,385,071 

II. quarter: 86.40; 5,094,233 

III. quarter: 88.07; 5,504,908 

| | Source: BAFA, some data provisional 

www.bafa.de 

EEX Trading Results 

December 2017 

(eex) In December 2017, the European 

Energy Exchange (EEX) achieved a 

total volume of 234.5 TWh on its 

power derivatives markets (December 

2016: 287.4 TWh). The December 

volume comprised 160.8 TWh traded 

at EEX via Trade Registration with 

subsequent clearing. Clearing and 

settlement of all exchange transactions 

was executed by European 

Commodity Clearing (ECC). 

On the German power derivatives 

market, trading volumes in Phelix- 

DE Futures (72.9 TWh) exceeded 

| | Uranium spot market prices from 1980 to 2017 and from 2007 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 2007 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. 

Phelix- DE/AT Futures (66.0 TWh) for 

the first time. On the markets for Italy 

(50.0 TWh) and Spain (7.4 TWh), 

EEX recorded the highest monthly 

volume of the year 2017. Compared to 

the previous year, volumes in these 

markets increased by 43% (Italy) and 

10% (Spain). On the Dutch power 

derivatives market, trading volumes 

almost doubled to 1.8 TWh (December 

2016: 0.9 TWh). 

The Settlement Price for base 

load contract (Phelix Futures) with 

delivery in 2018 amounted to 

37.67 €/MWh. The Settlement Price 

for peak load contract (Phelix Futures) 

with delivery in 2018 amounted to 

46.80 €/MWh. 

On the EEX markets for emission 

allowances, 65.6 million tonnes of 

CO 2 were traded in December 

( December 2016: 117.6 million tonnes 

of CO 2 ). Primary market auctions 

contributed 45.0 million tonnes of 

CO 2 to the total volume. 

The EUA price with delivery in 

December 2017 amounted to 

7.10/8.21 €/ EUA (min./max.). 

| | www.eex.com 

MWV Crude Oil/Product Prices 

November 2017 

(mwv) According to information and 

calculations by the Association of the 

German Petroleum Industry MWV e.V. 

in November 2017 the prices for super 

fuel, fuel oil and heating oil noted 

slightly higher compared with the 

pre vious month October 2017. The 

average gas station prices for Euro 

super consisted of 138.54 €Cent 

( October 2017: 134.72 €Cent, approx. 

+2.84 % in brackets: each information 

for pre vious month or rather previous 

month comparison), for diesel fuel of 

118.52 €Cent (116.196; +2.01 %) and 

for heating oil (HEL) of 60.06 €Cent 

(57.07 €Cent, +5.24 %). 

The tax share for super with 

a consumer price of 138.54 €Cent 

(134.72 €Cent) consisted of 

65.45 €Cent (47.24 %, 65.45 €Cent) 

for the current constant mineral oil 

tax share and 22.12 €Cent (current 

rate: 19.0 % = const., 21.51 €Cent) for 

the value added tax. The product 

price (notation Rotterdam) consisted 

of 39.06 €Cent (28.19 %, 36.20 €Cent) 

and the gross margin consisted of 

11.91 €Cent (8.60 %; 11.74 €Cent). 

Thus the overall tax share for super 

results of 66.24 % (67.58 %). 

Worldwide crude oil prices 

(monthly average price OPEC/Brent/ 

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

higher, approx. +9.43 % (+3.27 %) 

in November compared to October 

2017. 

The market showed a stable 

development with higher prices; each 

in US-$/bbl: OPEC basket: 60.74 

(53.44); UK-Brent: 62.70 (57.51); 

West Texas Inter mediate (WTI): 

56.64 (51.58). 

| | www.mwv.de 

133 

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134 

NUCLEAR TODAY 

Links to reference 

sources: 

President Macron 

interview: http:// 

reut.rs/2EIkEgM 

Trump on Iran: http:// 

nyti.ms/2mF1Ecp 

UK statement on 

Euratom: http://bit.ly/ 

2mGhrbf 

Author 

John Shepherd 

nuclear 24 

41a Beoley Road West 

St George’s 

Redditch B98 8LR, 

United Kingdom 

Playing Politics with Nuclear 

is All Part of the Game 

John Shepherd 

If a week is a long time in politics – a statement attributed to former British prime minister Harold Wilson – then what 

about a month, or several months? Just eight months ago, Emmanuel Macron was elected president of France. Among 

his portfolio of political pledges was one to respect reductions in the country’s nuclear park set out by his predecessor, 

Francois Hollande. 

Hollande’s administration had established an energy 

transition law which set a target of reducing the share of 

nuclear in France’s electricity mix to 50 % by 2025 from 

around 75 %. 

Fast forward to November 2017 and Macron’s environment 

minister, Nicolas Hulot, admitted that this could not 

be done – at least in the timeframe envisaged – without 

pushing up CO2 emissions, endangering security of power 

supply and the not-so-insignificant matter of risking 

thousands of jobs. Instead, Hulot said the government 

would come up with a more “realistic” target. 

Now move forward into early 2018 and France has 

signed a deal for closer cooperation in the development of 

civil nuclear with the China National Nuclear Corporation 

(CNNC). The agreement, signed by Framatome and CNNC 

during Macron’s visit to Beijing in January, also renewed a 

contract under which Framatome will supply nuclear fuel 

components to CNNC. 

As Macron’s visit came to a close, he issued a joint statement 

with his Chinese counterpart, Xi Jinping, to express 

“their high appreciation of the active cooperation between 

the two countries in the field of civilian nuclear energy and 

support a deepening of cooperation in the entire nuclear 

cycle”. 

Now this was indeed good news. France has had more 

than its fair share of ups and downs in the state-backed 

nuclear sector in recent years. But it begs the question, why 

would Macron want to expand civil nuclear activities in 

cooperation with an overseas partner if, back home, the 

goal is to reduce the reliance on nuclear? 

The answer is politics. As Macron was quoted telling 

France 2 television in an interview last December: “I don’t 

idolise nuclear energy at all. But I think you have to pick 

your battle. My priority in France, Europe and internationally 

is CO 2 emissions and (global) warming.” 

A leader who certainly does not shy away from battles is 

US president Donald Trump, who has also had nuclear 

power in his sights – but he too gives mixed messages on 

nuclear. 

On the domestic front, President Trump has been 

outspoken in his support for the use of civil nuclear energy 

as indeed he has for rejuvenating his country’s coal 

industry. However, proposals that paved the way for the US 

to offer incentives to power plants such as coal and nuclear 

in a bid to improve the resilience of the nation's power grid, 

were recently rejected by federal energy regulators. 

But Trump’s reason for backing nuclear does not appear 

to be linked to a desire to help the climate – or maybe it 

does – depending it seems on his temperament from one 

day to the next. You will recall that he pulled the US out of 

the Paris climate accord reached on his predecessor’s 

watch. 

But then a few weeks ago Trump said the US could 

go “go back” into the Paris deal. “We could conceivably go 

back in... I feel very strongly about the environment,” the 

president said during a joint news conference with 

Norwegian prime minister Erna Solberg. 

In a related move, Trump has demanded that European 

allies agree to rewriting a deal struck with Iran in 2015 – 

which lifted economic sanctions in exchange for Tehran 

limiting its nuclear ambitions beyond power generation – 

otherwise he said the US would pull out of the deal in the 

coming months, effectively “killing it”. 

The UK is also attempting a balancing act on matters 

nuclear. The government has confirmed Britain will exit 

Euratom at the same time as it withdraws from membership 

of the European Union on 29 March 2019. 

Greg Clark, secretary of state for business, energy and 

industrial strategy, told parliament the government’s 

“No.1 priority is continuity for the nuclear sector”. Clark 

said: “It is vitally important that our departure from the EU 

does not jeopardise this success, and it is in the interests of 

both the EU and the UK that our relationship should 

continue to be as close as possible.” 

Tom Greatrex, chief executive officer of the UK's Nuclear 

Industry Association, warned that even with a suitable 

transition being negotiated for Britain’s exit from the EU 

there “remains much work for the government to do 

to prevent the significant disruption that industry is 

concerned about.” 

Greatrex is of course correct. The UK has barely limped 

through the first phase of talks relating to Brexit and time 

is not on the side of either party. So for a minister to be 

talking about leaving Euratom – while at the same time 

continuing to enjoy the benefits that Euratom brings the 

UK – is surprising to say the least. 

Of course all these political machinations could be 

applied to any sector or policy and in any country. But the 

nuclear industry has long accepted that it can be used as a 

political football, to be kicked into goal or off the pitch 

completely depending on the situation at hand. 

I am reminded of a quotation from Otto von Bismarck, 

the ‘Iron Chancellor’, who said: “Politics is the art of the 

possible, the attainable – the art of the next best.” 

No political leader wants the lights going off and 

hurting homes, hospitals and businesses while they are in 

charge. They also don’t want to be seen as responsible for 

driving up unemployment. 

In terms of nuclear, whether cheerleaders for the 

technology or not, as the French president said: “You have 

to pick your battle.” The nuclear industry is all too familiar 

with fighting battles – defending itself from attack while 

quietly going about its task of safely supplying clean 

electricity to power-hungry grids around the world. 

Our industry therefore has power in the political sense 

too, but with power comes responsibility – nuclear leaders 

know that only too well and now is as good as time as ever 

to lead by example. 

Nuclear Today 

Playing Politics with Nuclear is All Part of the Game ı John Shepherd

Kommunikation und 

Training für Kerntechnik 

International sicher agieren 

Seminar: 

Advancing Your Nuclear English (Aufbaukurs) 

Im internationalen Dialog ist Englisch die universelle Sprache. Dies gilt für Geschäfts beziehungen 

im Allgemeinen ebenso wie für die Branche der Kerntechnik im Speziellen. In Deutschland gewinnen 

der internationale Austausch und damit das Englische zudem durch die auf das Jahr 2022 politisch 

begrenzte Stromerzeugung aus Kernenergie eine noch größere Bedeutung. 

Seminarinhalte 

ı Participating in an international conference for nuclear experts on “New products and processes” 

ı Before and during the conference 

ı Holding a town hall meeting in an international setting on “Safety issues at nuclear power facilities” 

ı Planning and conducting a town hall meeting 

ı After a town hall meeting 

Den Teilnehmerinnen und Teilnehmern wird über eine praxisorientierte Didaktik und unter der 

Verwendung „kerntechnischen Vokabulars“ das notwendige Know-how für den beruflichen Alltag 

vermittelt. Dabei gilt es sprachlich bedingte Kommunikationsbarrieren mit internationalem Kollegium 

und Kunden zu überwinden. 

Zielgruppe 

Diese 2-tägige Schulung richtet sich an Führungskräfte, Projektverantwortliche sowie Mitarbeiterinnen 

und Mitarbeiter aus allen Fachbereichen, bei denen Englisch für die organisationsinterne und/oder 

externe Kommunikation von Bedeutung ist. 

Maximale Teilnehmerzahl: 12 Personen 

Voraussetzungen 

Teilnehmerinnen und Teilnehmer sollten grundsätzliche Englischkenntnisse, in Form der Fähigkeit 

der allgemeinen Konversation in Wort und Schrift, mitbringen. Hierbei kann es sich um Kenntnisse 

handeln, die entweder während der Schulzeit bzw. während der Ausbildung/des Studiums oder 

aber berufs begleitend erworben wurden. (CEFR: etwa Niveau B1/B2). 

Referentin 

Devika Kataja 

Konferenzdolmetscherin, Fachübersetzerin und Sprachtrainerin (English Native Speaker) 

Wir freuen uns auf Ihre Teilnahme! 

Termin 

2 Tage 

11. bis 12. April 2018 

Tag 1: 10:30 bis 17:30 Uhr 

Tag 2: 09:00 bis 16:30 Uhr 

Veranstaltungsort 

Geschäftsstelle der INFORUM 


10115 Berlin 

Teilnahmegebühr 

898,– € ı zzgl. 19 % USt. 

Im Preis inbegriffen sind: 

ı Seminarunterlagen 

ı Teilnahmebescheinigung 

ı Pausenverpflegung 

inkl. Mittagessen 

Kontakt 

INFORUM 

Verlags- und Verwaltungsgesellschaft 

mbH 


10115 Berlin 

Petra Dinter-Tumtzak 

Fon +49 30 498555-30 

Fax +49 30 498555-18 

seminare@kernenergie.de 

Bei Fragen zur Anmeldung 

rufen Sie uns bitte an oder 

senden uns eine E-Mail.

The International Expert Conference on Nuclear Technology 

Outstanding Know-How 

and Innovations 

Insights 

on AMNT 

Watch the video: 

http://youtu.be/ 

DDC3L3XhnoA 

The AMNT 2018 offers a great variety of high level sessions in the fields 

of know-how, innovations and regulation. International speakers 

will discuss current issues and relevant developments. Expand your 

professional network in meetings with experts and decision-makers 

working in industry, utilities, research and development as well as 

politics and administration. 

Sessions 

3 International Regulation – Radiation Protection: 

The Implementation of the EU Basic Safety Standards Directive 2013/59 

and the Release of Radioactive Material from Regulatory Control 

3 Safety of Advanced Nuclear Power Plants 

3 Know-How, New Build and Innovations 

3 Reactor Physics, Thermo and Fluid Dynamics 

3 Young Scientists’ Workshop 

3 Nuclear Energy Campus 

Outstanding 

Know-How & 

Sustainable 

Innovations 

Enhanced 

Safety & 

Operation 

Excellence 

Decommissioning 

Experience & 

Waste Management 

Solutions 

Don’t miss this key event of the global nuclear energy community. 

29 – 30 May 2018 

Estrel Convention Center Berlin 

Germany 

www.nucleartech-meeting.com

atw 2018-02

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