atw 2018-02

inforum

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

2. True or false:

The global electricity demand will

decrease in the next decades!

a. True

b. False

3. True or 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

12. True or false:

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


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

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

2. Richtig oder falsch:

Der weltweite Strombedarf wird

in den nächsten Jahrzehnten sinken!

a. Wahr

b. Falsch

3. Richtig oder 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

12. Richtig oder falsch:

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

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


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

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


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

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.

Corrosion Processes of Alloyed Steels

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


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

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


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

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


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

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


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

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

01.03.2018

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,

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

22.08.-31.08.2018

Frédéric Joliot/Otto Hahn (FJOH) Summer School

FJOH-2018 – Maximizing the Benefits of

Experiments for the Simulation, Design and

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

Nuclear Energy Division of Commissariat à l’énergie

atomique et aux énergies alternatives (CEA) and

Karlsruher Institut für Technologie (KIT),

www.fjohss.eu

28.08.-31.08.2018

TINCE 2018 – Technological Innovations in

Nuclear Civil Engineering. Paris Saclay, France,

Société Française d’Energie Nucléaire,

www.sfen.org, www.sfen-tince2018.org

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,

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

Calendar


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

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

Energy Policy, Economy and Law

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


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

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

Energy Policy, Economy and Law

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


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

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

Energy Policy, Economy and Law

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


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

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


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

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

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


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

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.

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


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

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.

ENVIRONMENT AND SAFETY 87

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.-%.

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ENVIRONMENT AND SAFETY 88

2.3 Variant calculation with a

10 times larger containment

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

10 times larger containment

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

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

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

ENVIRONMENT AND SAFETY 89

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

ENVIRONMENT AND SAFETY 90

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

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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

ENVIRONMENT AND SAFETY 91

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.

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ENVIRONMENT AND SAFETY 92

| | Fig. 5.

MIDAS Nodalization Scheme with 6 Channels.

| | Fig. 6.

MIDAS Nodalization Scheme with 12 Channels.

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

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| | Fig. 8.

Liquid Flow Pattern of KM100 Test Calculation

with 4 Channels Nodalization.

| | Fig. 9.

Liquid Flow Pattern of KM100 Test Calculation

with 6 Channels Nodalization.

| | Fig. 10.

Liquid Flow Pattern of KM100 Test Calculation

with 12 Channels Nodalization.

ENVIRONMENT AND SAFETY 93

| | Fig. 11.

Comparison of the Measured and Calculated ECC Bypass Fraction for Low

Steam Flow Cases.

| | Fig. 12.

Liquid Flow Pattern of KM114 Test Calculation

with 6 Channels Nodalization.

| | Fig. 13.

Vapor Flow Pattern of KM114 Test Calculation

with 6 Channels Nodalization.

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.

Environment and Safety

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


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

ENVIRONMENT AND SAFETY 94

| | Fig. 16.

Liquid Flow Pattern of KM114 Test Calculation

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

Environment and Safety

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


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

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

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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.

Operation and New Build

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

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

OPERATION AND NEW BUILD 97

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

OPERATION AND NEW BUILD 98

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

safe shutdown.

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

safe shutdown.

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.

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


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

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

safe shutdown.

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

OPERATION AND NEW BUILD 99

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

OPERATION AND NEW BUILD 100

| | 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

19 B1EEFHCC9SA H2 C27 B-EF-HV222 HSD-P2 HSD-S B1EZJF4TPBA SE

20 B1EEFHCC9SB H2 C27 B-EF-HV222 HSD-P2 HSD-S B1EZJF4TPBA SE

21 B1EEFHCC9SC H2 C27 B-EF-HV222 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.

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


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

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

OPERATION AND NEW BUILD 101

1 B-EF-HV203 ON

2 B-EF-HV206 ON

3 B-EF-HV221 ON

4 B-EF-HV222 OFF

5 B-EF-HV230 OFF

| | Tab. 6.

Valve states in hot shutdown mode.

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

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

1 B-EF-HV203 ON

2 B-EF-HV206 ON

3 B-EF-HV221 ON

4 B-EF-HV222 OFF

5 B-EF-HV230 OFF

| | 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

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

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

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

8 B1EEFHCC9SA H2 C27 B-EF-HV222 HSD-P2 HSD-S B1EZJF4TPBA SE

9 B1EEFHCC9SB H2 C27 B-EF-HV222 HSD-P2 HSD-S B1EZJF4TPBA SE

10 B1EEFHCC9SC H2 C27 B-EF-HV222 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

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

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

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Water Reactor Electric Generating

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| | 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

OPERATION AND NEW BUILD 103

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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


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

104

DECOMMISSIONING AND WASTE MANAGEMENT

Corrosion Processes of Alloyed Steels

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

Decommissioning and Waste Management

Corrosion Processes of Alloyed Steels in Salt Solutions ı Bernhard Kienzler


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

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

Decommissioning and Waste Management

Corrosion Processes of Alloyed Steels in Salt Solutions ı Bernhard Kienzler


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

DECOMMISSIONING AND WASTE MANAGEMENT 106

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

b) mass loss rate

| | 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.

Decommissioning and Waste Management

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DECOMMISSIONING AND WASTE MANAGEMENT 108

a) mass loss

b) mass loss rate

| | 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

Decommissioning and Waste Management

Corrosion Processes of Alloyed Steels in Salt Solutions ı Bernhard Kienzler


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

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,

Forschungszentrum Karlsruhe, KfK-

4552, 1989.

[3] Kommission Lagerung hoch radioaktiver

Abfallstoffe, ABSCHLUSSBERICHT:

Verantwortung für die Zukunft: Ein faires

und transparentes Verfahren für die

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Geschäftsstelle der

Kommission Lagerung hoch radioaktiver

Abfallstoffe, K-Drs 268, 2016.

[4] Gesetz zur Fortentwicklung des

Gesetzes zur Suche und Auswahl eines

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.

DECOMMISSIONING AND WASTE MANAGEMENT 109

Decommissioning and Waste Management

Corrosion Processes of Alloyed Steels in Salt Solutions ı Bernhard Kienzler


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

DECOMMISSIONING AND WASTE MANAGEMENT 110

[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,

J. Electrochem. Soc. , Vol. 123,

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.

Res. Rept. - Nucl. Sci. Technol., Vol. 5,

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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Decommissioning and Waste Management

Corrosion Processes of Alloyed Steels in Salt Solutions ı Bernhard Kienzler


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

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.

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

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


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

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.

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

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


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

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

RESEARCH AND INNOVATION 113

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

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


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

RESEARCH AND INNOVATION 114

| | 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

Computer Code.

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

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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

Computer Code.

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

KIANA Advance Computational Computer

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

Advance Computational Computer

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 )

RESEARCH AND INNOVATION 115

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| | 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

Computer Code.

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

KIANA Advance Computational Computer

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

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RESEARCH AND INNOVATION 117

| | Fig. 5.

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

Computer Code.

| | 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

Research and Innovation

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RESEARCH AND INNOVATION 118

| | Fig. 7.

Code Algorithms of Concentration of activity in products is calculated from

the raw product that used in construction of KIANA Advance Computational

Computer Code.

| | 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

Advance Computational Computer

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

KIANA Advance Computational Computer

Code is developed as dynamic

software (such as DoseCAL), only

dynamic codes' results are presented

for comparison. KIANA Advance

Computational Computer Code has a

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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

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RESEARCH AND INNOVATION 119

| | 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

Advance Computational Computer

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

KIANA Advance Computational Computer

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

Advance Computational Computer

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

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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

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RESEARCH AND INNOVATION 120

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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Exposure to Radionuclides in Air, Water

and Soil, 1993.

| | Environmental Modelling for Radiation

Safety (EMRAS) Programme, The

Chernobyl I-131 Release: Model

Validation and Assessment of the

Countermeasure Effectiveness: Report

of the Chernobyl 131-I Release Working

Group of EMRAS Theme 1.

| | EUR-18825, FZKA-6311, ISBN 92-894-

2085-5, European Communities 2001

Probabilistic Accident Consequence

Uncertainty Assessment Using COSYMA:

Uncertainty from the Dose Module.

| | EUR-18826, FZKA-6312, ISBN- 92-894-

2088-X, European Communities 2001

Probabilistic Accident Consequence

Uncertainty Assessment Using COSYMA:

Overall Uncertainty Analysis.

| | Eyüpoğlu, F., Türkiye topraklarının

verimlilik durumları, 1999.

| | Gardner, R.H.: Huff, D.D., O'Neill, R.V.,

Mankin, J.B., Carney, J. and Jones, J.:

1980, Application of Error Analysis to a

Marsh Hydrology Model, Water

Resources Res. 16, 659-664.

| | Gardner, R.H., O'Neill, R.V., Mankin, J.B.

and Carney, J.H.: 1981, A Comparison

of Sensitivity Analysis and Error Analysis

Based on a Stream Ecosystem Model,

Ecol. Modelling. 12, 173- 190.

| | Garger, E.K., Hoffman, F.O., Thiessen,

K.M., Uncertainty of the long-term

resuspension factor, Atmos. Environ. 31

(1997) 1647–1656.

| | Health Canada, Recommendations on

Dose Coefficients for Assessing Doses

from Accidental Radionuclide Releases

to the Environment, 1999.

| | Health Protection Agency, Application

of the 2007 Recommendations of the

ICRP to the UK, 2009.

| | Helton, J.C., Garner, J.W., Marietta,

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

Swift, EN.: 1993. Uncertainty and

Sensitivity Analysis Results Obtained in

a Preliminary Performance assessment

for the Waste Isolation Pilot Plant, Nuc.

Sci. and Eng. 114, 286-331.

| | Helton, J.C., Garner, J.W., McCurley, R.D.

and Rudeen, D.K. Sensitivity analysis

techniques and results for performance

assessment at the waste isolation pilot

plant. Albuquerque, NM: Sandia

National Laboratory; Report No.

SAND90-7103, 1991.

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

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


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

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.

Department of Nuclear Eng.

Science and Research Branch

of Islamic Azad University

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

Department of Nuclear Eng.

Deputy Manager of execution and

Research in Nuclear Eng. Faculty

Tehran, Iran

R. Khodadadi

Consultant

Science and Research Branch

of Islamic Azad University

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,

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

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

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

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

• Wann? Montag, 28. Mai 2018, 18.00 Uhr

• Wo? Estrel Convention Center, Raum IV

(2. OG), Sonnenallee 225, 12057 Berlin

33

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

• Wann? Montag, 28. Mai 2018, 19.00 Uhr

• 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


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

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


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

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


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

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


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

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)

Robert-Koch-Platz 4

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


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

128

KTG INSIDE

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Herzlichen

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Februar 2018

90 Jahre wird

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Erlangen

89 Jahre wird

20. Dr. Helmut Hübel, Bensberg

88 Jahre wird

5. Dr. Eberhard Teuchert, Leverkusen

87 Jahre wird

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85 Jahre wird

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84 Jahre werden

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23. Prof. Dr. Dr.-Ing. E.h. Adolf Birkhofer,

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6. Dr. Ashu-T. Bhattacharyya, Erkelenz

17. Dr. Helfrid Lahr, Wedemark

81 Jahre werden

5. Prof. Dr. Arnulf Hübner, Berlin

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21. Dipl.-Ing. Hubert Andrae, Rösrath

80 Jahre wird

15. Dr. Hans-Heinrich Krug, Saarbrücken

79 Jahre werden

3. Dr. Roland Bieselt, Kürten

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8. Dr. Herbert Spierling, Dietzenbach

22. Dr. Manfred Schwarz, Dresden

78 Jahre werden

9. Dr. Gerhard Preusche, Herzogenaurach

13. Dr. Hans-Ulrich Fabian, Gehrden

14. Dipl.-Ing. Kurt Ebbinghaus,

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Die KTG gratuliert ihren Mitgliedern sehr herzlich zum

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75 Jahre werden

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28. Dr. Klaus Tägder, Sankt Augustin

70 Jahre werden

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14. Reinhold Rothenbücher, Erlangen

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65 Jahre werden

3. Dr. Reinhard Knappik, Dresen

20. Dipl.-Ing. Berthold Racky, Nidderau

60 Jahre werden

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88 Jahre werden

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86 Jahre wird

14. Dr. Peter Engelmann,

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85 Jahre werden

26. Dipl.-Ing. Gerhard Frei, Uttenreuth

30. Dipl.-Phys. Dieter Pleuger, Kiedrich

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8. Dr. Frank Steinbrunn, Fröndenberg

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76 Jahre wird

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75 Jahre werden

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20. Dipl.-Ing. Jörg Brauns, Hanau

26. Dr. Jürgen P. Lempert, Hannover

26. Graeme William Catto, Buch a. Erlbach

70 Jahre werden

5. Dipl.-Wirtsch.-Ing. Bernd Pontani,

Alzenau

13. Dipl.-Kfm. Jochen Bläsing, Mörlenbach

22. Dr. Volker Mirschinka, Essen

65 Jahre wird

21. Dr. Ulrich Rohde, Dresden

60 Jahre wird

26. Dr. Sheikh Shahee, Leinburg

50 Jahre werden

20. Thomas Wiese, Ebermannstadt

30. Dipl.-Ing. Heiko Ringel, Offingen

KTG Inside


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

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


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

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


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

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

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NEWS

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

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

• Conversion: 7.25–11.00

• Separative work: 86.00–98.00

January to June 2015

• Uranium: 35.00–39.75

• Conversion: 7.00–9.50

• Separative work: 70.00–92.00

June to December 2015

• Uranium: 35.00–37.45

• Conversion: 6.25–8.00

• Separative work: 58.00–76.00

2016

January to June 2016

• Uranium: 26.50–35.25

• Conversion: 6.25–6.75

• Separative work: 58.00–62.00

July to December 2016

• Uranium: 18.75–27.80

• Conversion: 5.50–6.50

• Separative work: 47.00–62.00

2017

January 2017

• Uranium: 20.25–25.50

• Conversion: 5.50–6.75

• Separative work: 47.00–50.00

February 2017

• Uranium: 23.50–26.50

• Conversion: 5.50–6.75

• Separative work: 48.00–50.00

March 2017

• Uranium: 24.00–26.00

• Conversion: 5.50–6.75

• Separative work: 47.00–50.00

April 2017

• Uranium: 22.50–23.50

• Conversion: 5.00–5.50

• Separative work: 45.50–48.50

May 2017

• Uranium: 19.25–22.75

• Conversion: 5.00–5.50

• Separative work: 42.00–45.00

June 2017

• Uranium: 19.25–20.50

• Conversion: 5.55–5.50

• Separative work: 42.00–43.00

News


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

July 2017

• Uranium: 19.75–20.50

• Conversion: 4.75–5.25

• Separative work: 42.00–43.00

August 2017

• Uranium: 19.50–21.00

• Conversion: 4.75–5.25

• Separative work: 41.00–43.00

September 2017

• Uranium: 19.75–20.75

• Conversion: 4.60–5.10

• Separative work: 40.50–42.00

October 2017

• Uranium: 19.90–20.50

• Conversion: 4.50–5.25

• Separative work: 40.00–43.00

November 2017

• Uranium: 20.00–26.00

• Conversion: 4.75–5.25

• Separative work: 40.00–43.00

December 2017

• Uranium: 23.50–25.50

• Conversion: 5.00–6.00

• Separative work: 39.00–42.00

| | Source: Energy Intelligence

www.energyintel.com

Cross-border Price

for Hard Coal

Cross-border price for hard coal in

[€/t TCE] and orders in [t TCE] for

use in power plants (TCE: tonnes of

coal equivalent, German border):

2012: 93.02; 27,453,635

2013: 79.12, 31,637,166

2014: 72.94, 30,591,663

2015: 67.90; 28,919,230

2016: 67.07; 29,787,178

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

NEWS

News


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

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

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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

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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

Robert-Koch-Platz 4

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Teilnahmegebühr

898,– € ı zzgl. 19 % USt.

Im Preis inbegriffen sind:

ı Seminarunterlagen

ı Teilnahmebescheinigung

ı Pausenverpflegung

inkl. Mittagessen

Kontakt

INFORUM

Verlags- und Verwaltungsgesellschaft

mbH

Robert-Koch-Platz 4

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

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