atw - International Journal for Nuclear Power

nucmag.com

2020

ISSN · 1431-5254

24.– €

Energy Supply

Without Nuclear:

Winter 2022/23

is Coming

Dual-Use Act in Trialog

Nuclear Power Plants:

2019 atw Compact Statistics

Programme Overview Inside!

Kommunikation und

Training für Kerntechnik

Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik?

Wählen Sie aus folgenden Themen: Dozent/in Termin/e Ort

3 Atom-, Vertrags- und Exportrecht

Atomrecht – Ihr Weg durch Genehmigungs- und

Aufsichtsverfahren

RA Dr. Christian Raetzke 18.02.2020 Berlin

Atomrecht – Was Sie wissen müssen

RA Dr. Christian Raetzke

Akos Frank LL. M.

11.11.2020 Berlin

Atomrecht – Das Recht der radioaktiven Abfälle RA Dr. Christian Raetzke 10.03.2020 Berlin

Export kerntechnischer Produkte und Dienstleistungen –

Chanchen und Regularien

3 Kommunikation und Politik

RA Kay Höft M.A. (BWL) 17.06.2020 Berlin

Public Hearing Workshop –

Öffentliche Anhörungen erfolgreich meistern

Dr. Nikolai A. Behr 10.11. - 11.11.2020 Berlin

3 Rückbau und Strahlenschutz

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

Das neue Strahlenschutzgesetz –

Folgen für Recht und Praxis

Stilllegung und Rückbau in Recht und Praxis

Dr. Maria Poetsch

RA Dr. Christian Raetzke

Dr. Stefan Kirsch

RA Dr. Christian Raetzke

28.01. - 29.01.2020

16.06. - 17.06.2020

23.09. - 24.09.2020

Berlin

17.03. - 18.03.2020 Berlin

3 Nuclear English

English for Nuclear Business Angela Lloyd 01.04. - 02.04.2020 Berlin

3 Wissenstransfer und Veränderungsmanagement

Veränderungsprozesse gestalten – Heraus forderungen

meistern, Beteiligte gewinnen

Erfolgreicher Wissenstransfer in der Kerntechnik –

Methoden und praktische Anwendung

Dr. Tanja-Vera Herking

Dr. Christien Zedler

Dr. Tanja-Vera Herking

Dr. Christien Zedler

21.01. - 22.01.2020 Berlin

24.03. - 25.03.2020 Berlin

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atw Vol. 65 (2020) | Issue 1 ı January

USA: 80 Years Actually

When nuclear energy was at its beginning in the 1950s and 1960s and the first nuclear power plants for energy supply

were built commercially, the question of a suitable regulatory framework also arose. Among other things, it was

necessary to establish licensing and supervisory procedures that would guarantee safety and thus also responsibility

and acceptance at an optimum level from the first to the last day of operation and beyond. On the other hand, the

potential and future operators also had to be able to plan with a sufficiently safe operating period in order to decide on

the technical and financial investment.

In the USA, then President Dwight D. Eisenhower set an

important political signal for the national and international

expansion of nuclear energy with his “Atoms for Peace”

speech on 8 December 1953. The U.S. Atomic Energy Act of

1956 opened up a reliable, long-term perspective for

nuclear in the U.S. Since then, a staggered licensing concept

based on this and other regulations has ensured the reliable

operation of nuclear power plants in the United States from

a regulatory and technical/safety point of view. The regulatory

basis stipulates that the first operating licence is issued

for a period of 40 years. In addition, the regulations provide

for the possibility of a license extension for a further

20 years. There is no restriction on the number of such

subsequent licenses. These extensions are based on

corresponding evidence of plant safety, which has to be

demonstrated and guaranteed for the entire intended

licensing period.

Some of the other countries that use nuclear energy

have similar regulations, others deviate, for example, with

regard to licensing periods – e.g. follow-up licenses are

granted for 10 years – and others have no time limit at all.

It must be clearly pointed out at this point that the individual

safety of nuclear power plants must be guaranteed

at all times, irrespective of operating time regulations.

Safety depends on the situation.

In the USA, the Nuclear Regulatory Commission (NRC)

started in the early 1980s to systematically record and

investigate ageing processes and thus important aspects of

long-term nuclear safety. At the beginning of the 1990s,

the result was that the previously announced plant

extensions – the Atomic Energy Act dates from 1956, see

above – could also be technically implemented. From a

timing point of view, this statement was appropriate.

Early-to-mid-1990s, nuclear energy in the USA was at a

crossroads between the continued operation of existing

plants and short-term final shut-down. Only moderate

availability of nuclear power plants of average 55 to 70 %

(period: 1970 to 1990) and correspondingly significant

high generation costs stood in contrast to increasingly

favour able generation costs, especially for coal and

gas-fired power plants. The U.S. nuclear power plant

operators made the right decision both retrospectively and

with a view to the future: Measures to improve operational

reliability and availability were developed and implemented

in a coordinated and joint manner in the USA, all

with a view to extending operating lifetime. One result is

164 individual measures in U.S. nuclear power plants with

power increases in the plants totalling 7,921 MWe (net) –

roughly equivalent to the addition of 7 powerful nuclear

power plants. A further visible result is the increase in

availability to 92.3 % today (2018) and thus the top result

worldwide for a country's nuclear power plant park. In

other words, nuclear power plants in the USA today produce

50 to 80 % more electricity than 30 years ago.

As far as the long-term prospects for the operation of

nuclear power plants are concerned, the year 2000 set a

first mark: After two years of evaluation, five nuclear

power plant units were the first plants in the USA to receive

an initial renewal license in the spring of the year to

operate for a period of 20 additional years to a total

operating life of 60 years. To date, a further 94 permits

have been issued. Four further applications for lifetime

extensions have been announced to NRC until 2022. This

means that all nuclear power plants in the USA for which

longer operating times are planned by the operator now

have the required 20-year initial renewal licence or the

process has been initiated.

But that is not all: As mentioned above, the number of

further lifetime extensions is not limited to the first

approval according to U.S. regulations. One result of the

early evaluation of the long-term safety of the U.S. nuclear

power plants by NRC was also that beyond the operating

time of 60 years, essential components of the nuclear

power plants which determine the long-term safety can

easily be extended beyond that. In July 2017, the NRC

therefore published a guideline for the evaluation of

“ subsequent license renewal applications”: the guideline

includes, among other things, details of the 45-day policy

review for the application documents as well as the

subsequent review of the safety-related aspects and the

environmental impact assessment.

In January 2018, the operator Florida Power & Light of

the Turkey Point nuclear power plant – two pressurised

water reactors with a gross capacity of about 885 MWe

each are operated at the site located about 30 km south

of Miami in the U.S. state of Florida – submitted the

application for the second, subsequent 20-year lifetime

extension. On 5 December 2019, NRC announced that it

had approved the application for the extension of the

operating licence for both nuclear power plant units. For

the first time, this is an 80 year licence for a nuclear power

plant in the USA. The Turkey Point 3 unit can thus supply

the customers with electricity until 19 July 2052 and the

Turkey Point 4 unit until 10 April 2053. Two further

applications for the reactors Peach Bottom 2 & 3 and Surry

1 & 2 are in the approval process. Decisions on these are

expected in 2020.

Both the NRC's decision and the transparent approval

process – interested parties can expect around 5,000 pages

of publicly accessible technical information on the NRC

website alone – have one thing in common: it impressively

demonstrates, that it is not age but proven safety that is

decisive for nuclear energy.

Christopher Weßelmann

– Editor in Chief –

EDITORIAL

Editorial

USA: 80 Years Actually

atw Vol. 65 (2020) | Issue 1 ı January

EDITORIAL 4

USA: Tatsächlich 80 Jahre

Als die Kernenergie in den 1950er- und 1960er-Jahren aus den Kinderschuhen entwuchs und erste Kernkraftwerke für

die Energieversorgung kommerziell errichtet wurden, stellte sich auch die Frage nach einem geeigneten Regelwerk. Unter

anderem galt es, Genehmigungs- und Aufsichtsverfahren zu etablieren, die in Bezug auf die Sicherheit und damit auch

Verantwortung und Akzeptanz vom ersten bis zum letzten Betriebstag und darüber hinaus diese stetig auf optimalem

Niveau gewährleisten. Andererseits mussten die potenziellen und späteren Betreiber auch mit einer ausreichend sicheren

Perspektive planen können, um sich für die technische und finanzielle Investition zu entscheiden.

In den USA hatte der damalige Präsident Dwight D. Eisenhower

mit seiner „Atoms for Peace“-Rede am 8. Dezember

1953 ein wichtiges politisches Signal für den nationalen und

internationalen Ausbau der Kernenergie gesetzt. Die verlässliche,

langfristige Perspektive in den USA eröffnete dann

das U.S.-Atomgesetz von 1956. Ein darauf und weiteren

Regularien basierendes zeitlich gestaffeltes Geneh mi gungskonzept

gewährleistet seitdem regulatorisch und sicherheitstechnisch

verlässlich den Betrieb der Kernkraftwerke. Die

regulatorische Grundlage sieht vor, dass die erste Betriebsgenehmigung

auf eine Laufzeit von 40 Jahren festgelegt ist.

Zudem sehen die Regelungen die Möglichkeit einer

Genehmi gungsverlängerung für weitere 20 Jahre vor. Eine

Einschränkung für die Zahl solcher Verlängerungen besteht

nicht. Eine Vorraussetzung für diese Betriebszeitverlängerungen

sind entsprechende Nachweise für die Anlagensicherheit,

die für den gesamten angestrebten Genehmigungszeitraum

nachzuweisen und zu gewährleisten ist.

In den weiteren Kernenergie nutzenden Staaten

existieren teils ähnliche Regularien, teils weichen diese

zum Beispiel hinsichtlich der Genehmigungszeiträume ab

– Folgegenehmigungen werden z. B. für 10 Jahre ausgesprochen

oder aber sind auch gänzlich unbefristet.

Deutlich muss an dieser Stelle darauf hingewiesen

werden, dass die individuelle Sicherheit der Kernkraftwerke

unabhängig von Laufzeitregularien jederzeit zu

gewährleisten ist. Sicherheit ist abhängig von der Sachlage.

In den USA hatte die Aufsichtsbehörde Nuclear Regulatory

Commission (NRC) in den frühen 1980er-Jahren

begonnen, Alterungsprozesse und damit wichtige Aspekte

der Langzeitsicherheit systematisch zu erfassen und zu

untersuchen. Anfang der 1990er-Jahre war das Ergebnis,

dass sich die zuvor avisierten Betriebsverlängerungen – der

Atomic Energy Act stammt aus dem Jahr 1956, s. o. – auch

technisch realisieren lassen. Vom Zeitpunkt her passte diese

Feststellung. Anfang, Mitte der 1990er Jahre stand die

Kernenergie in den USA auf dem Scheideweg zwischen

Weiterbetrieb der bestehenden Anlagen und kurzfristiger

endgültiger Stilllegung. Nur mäßigen Arbeitsverfügbarkeiten

der Kernkraftwerke von im Mittel 55 bis 70 % (Zeitraum:

1970 bis 1990) und entsprechend signifikant hohen

Erzeugungskosten standen günstiger werdende Erzeugungs

kosten vor allem von Kohle- und Gaskraftwerken

gegenüber. Die U.S.-Kernkraftwerksbetreiber entschieden

sich sowohl rück- als auch in die Zukunft blickend damals

richtig: Koordiniert und gemeinsam wurden in den USA

Maßnahmen entwickelt und umgesetzt, um die betriebliche

Zuverlässigkeit sowie die Verfügbarkeit zu verbessern, alles

auch mit Blick auf Verlängerungen der Laufzeiten. Ein

Ergebnis sind 164 Einzelmaßnahmen in U.S.-Kernkraftwerken

mit Leistungserhöhungen in den Anlagen von in

Summe 7.921 MWe (netto) – dies entspricht in etwa dem

Zubau von sieben leistungsstarken Kernkraftwerken. Ein

weiteres sichtbares Ergebnis ist die Steigerung der Verfügbarkeiten

auf heute (2018) 92,3 % und damit das weltweite

Spitzen ergebnis für den Kernkraftwerkspark eines Landes.

Anders ausgedrückt, produzieren Kernkraftwerke in den

USA heute 50 bis 80 % mehr Strom als vor 30 Jahren.

Was die Langfristperspektiven des Kernkraftwerksbetriebs

betrifft, setzte das Jahr 2000 eine erste Marke:

Nach zwei Jahren Prüfung erhielten im Frühjahr des

Jahres gleich fünf Kernkraftwerksblöcke als erste Anlagen

in den USA überhaupt die Genehmigung für einen um

20 Jahre längeren Betrieb auf dann 60 Jahre Gesamtbetriebszeit.

Bis heute wurden weitere 94 Genehmigungen

erteilt. Vier weitere Anträge auf Laufzeitverlängerung sind

bis zum Jahr 2022 bei der NRC avisiert. Damit besitzen alle

Kernkraftwerke in den USA, für die längere Betriebszeiten

vom Betreiber geplant sind, die erforderliche Genehmigung

bzw. der Verfahrensprozess ist initiiert.

Damit nicht genug: Wie eingangs erwähnt, ist die

Anzahl von weiteren Laufzeitverlängerungen auf die Erstgemnehmigungen

gemäß U.S.-Regularien nicht begrenzt.

Ein Ergebnis der frühen Evaluierung der langfristigen

Sicherheit der U.S.-Kernkraftwerke durch die NRC war

auch, dass jenseits der Betriebszeit von 60 Jahren wesentliche,

die Langzeitsicherheit bestimmende Komponenten

der Kernkraftwerke ohne Weiteres auch darüber hinaus

gehende Laufzeiten gestatten. Im Juli 2017 veröffentlichte

die NRC daher einen Leitfaden für die Evaluierung von

„Folgeanträgen auf Laufzeitverlängerung“: der Leitfaden

umfasst unter anderem Details zum 45 Tage dauernden

Grundsatzreview für die Antragsunterlagen sowie die

folgende Prüfung der sicherheitstechnischen Aspekte und

der Umweltverträglichkeitsprüfung.

Im Januar 2018 übermittelte der Betreiber Florida

Power & Light des Kernkraftwerks Turkey Point – zwei

Druckwasserreaktoren mit jeweils rund 885 MWe Bruttoleistung

werden am rund 30 km südlich von Miami im

U.S.-Bundesstaat Florida gelegenen Standort betrieben –

den Antrag auf die zweite 20-Jahres-Laufzeit verlängerung.

Am 5. Dezember 2019 teilte die NRC mit, dass sie den

Antrag auf Verlängerung der Betriebsgenehmigung für

beide Kernkraftwerksblöcke genehmigt habe. Dies ist

erstmalig eine Genehmigung für 80 Jahre Laufzeit für ein

Kernkraftwerk in den USA. Der Block Turkey Point 3 kann

damit bis zum 19. Juli 2052 Strom produzieren, die Anlage

Turkey Point 4 bis zum 10. April 2053. Zwei weitere

Anträge für die Reaktoren Peach Bottom 2 & 3 sowie

Surry 1 & 2 befinden sich im Genehmigungsprozess.

Ent scheidungen zu diesen werden in 2020 erwartet.

Eines haben sowohl die Entscheidung der NRC als auch

der transparente Genehmigungsprozess – den Interessierten

erwarten allein rund 5.000 Seiten öffentlich

zugäng licher technischer Informationen auf den Webseiten

der NRC – gemeinsam: eindrucksvoll wird für die

Kernenergie demonstriert, dass nicht das Alter, sondern

die nach ge wiesene Sicherheit entscheidend ist.

Christopher Weßelmann

– Chefredakteur –

Editorial

USA: 80 Years Actually

atw Vol. 65 (2020) | Issue 1 ı January

Forward-looking Balance of the Supply and Demand Equilibrium

for Electricity in France by RTE

2019 Edition

The French transmisson system operator RTE (Réseau de transport

d'électricité) presented the annual Forward-looking balance

of the supply and demand equilibrium for electricity in France

(Bilan prévisionnel de l’équilibre offer-demande d’électricité en

France) in November. The 2019 edition includes a modeling of

the electrical systems of other European countries next to the

analysis of the French situation. The latter one is characterized for

the upcoming years by the phase-out of the remaining coal fired

power plants till 2022 (-3 GW) and the shut-down of the two

units of the NPP Fessenheim in February and June 2020

(-1,8 GW). This takes place in the context of an increased number

of NPP 10-year refurbishments in France, the postponed start of

operation of unit 3 of the NPP Flamanville only in 2023/24

(+1,6 GW) and of major coal and nuclear phase-out policies of

neighboring countries of France. Altogether these factors will

lead to a period of high alertness concerning the security of

electricity supply in France from the end of 2022 to 2025. In this

period the national criterion (the duration in which the balance

between supply and demand of electricity cannot be guaranteed

by the electricity market has to be inferior to three hours in all

analyzed scenarios) cannot be guaranteed. In this time a cold

spell such as in 2012 will lead to the necessity of load shedding in

most scenarios and the electrical system will be vulnerable to

weather situations in which low wind prevails in many parts of

Europe and will make the import of electricity to France difficult or

impossible. Below you find an overview of major planned or

announced reductions to disposable generation in countries

neighboring France.

5Did you know...?

DID YOU EDITORIAL KNOW...? 5

Main objectives for the phase-out of thermal power plants in Europe

Gradual phase-out

of coal power till 2025

-4 GW

Shut-down of the

last reactor in 2025

-6 GW

Gradual phase-out

of coal power till 2030

-9 GW

Shut-down of the

last reactor in 2022

-9.5 GW

Gradual phase-out

of coal power till 2038

-15 GW till 2030

Source: RTE, Bilan prévisionnel de l’équilibre offre- demande d’électricité en France, Édition 2019

Gradual phase-out

of coal power till 2025

-6 GW

For further details

please contact:

Nicolas Wendler

KernD

Robert-Koch-Platz 4

10115 Berlin

Germany

E-mail: presse@

KernD.de

www.KernD.de

Did you know...?

atw Vol. 65 (2020) | Issue 1 ı January

Issue 1 | 2020

January

CONTENTS

Contents

Editorial

USA: 80 Years Actually E/G . . . . . . . . . . . . . . . . . . . . . . . . . . 3

Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

Inside Nuclear with NucNet

Nuclear Fusion / Revived € 20 Billion Iter Project

‘Entering a Critical Phase’ . . . . . . . . . . . . . . . . . . . . . . . . . . .8

Feature | Energy Policy, Economy and Law

Energy Supply Without Nuclear:

Winter 2022/23 is Coming . . . . . . . . . . . . . . . . . . . . . . . . . . 9

Spotlight on Nuclear Law

Dual-Use-Verordnung im Trilog G . . . . . . . . . . . . . . . . . . . . . 11

Calendar . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

Environment and Safety

Analysis of Ultimate Response Guidelines

for Chinshan Nuclear Power Plant in Taiwan to Cope

with Postulated Compound Accident . . . . . . . . . . . . . . . . . . . 13

Decommissioning and Waste Management

Decommissioning & Dismantling of the

Rossendorf Research Reactor RFR | Part 2 G . . . . . . . . . . . . . . 17

Research and Innovation

Thermal-Hydraulic Analysis for Total Loss

of Feedwater Event in PWR using SPACE Code . . . . . . . . . . . . . 25

CFD Simulation of Flow Characteristics and

Thermal Performance in Circular Plate and Shell Oil Coolers . . . . 29

Research on Neutron Diffusion and Thermal Hydraulics

Coupling Calculation based on FLUENT and

its Application Analysis on Fast Reactors . . . . . . . . . . . . . . . . . 35

Kerntechnik 2020

Programme Overview E/G . . . . . . . . . . . . . . . . . . . . . . . . . 45

KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47

Statistics

Nuclear Power Plants: 2019 atw Compact Statistics . . . . . . . . . . 48

Obituary

Prof. Dr. Dr. Adolf Birkhofer G . . . . . . . . . . . . . . . . . . . . . . . 53

Cover:

Requirements and challenges

for a secure electricity supply.

E/G

= German

= English/German

News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54

Nuclear Today

New Year Brings a Fresh Political Challenge

for a Champion of Climate Change . . . . . . . . . . . . . . . . . . . . 58

Imprint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44

Contents

atw Vol. 65 (2020) | Issue 1 ı January

Feature

Energy Policy, Economy and Law

9 Energy Supply Without Nuclear:

Winter 2022/23 is Coming

CONTENTS

Roman Martinek

Spotlight on Nuclear Law

11 Dual-Use Act in Trialog

Dual-Use-Verordnung im Trilog

Ulrike Feldmann

Decommissioning and Waste Management

17 Decommissioning & Dismantling of the

Rossendorf Research Reactor RFR | Part 2

Stilllegung und Rückbau

des Rossendorfer Forschungsreaktors RFR | Teil 2

Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

Research and Innovation

25 Thermal-Hydraulic Analysis for Total Loss of Feedwater Event in PWR

using SPACE Code

MinJeong Kim, Minhee Kim, Junkyu Song, Bongsik Chu, Jae-Seung Suh and Hyunjin Lee

29 CFD Simulation of Flow Characteristics and Thermal Performance

in Circular Plate and Shell Oil Coolers

Shen Ya-jie, Gao Yong-heng and Zhan Yong-jie

Statistics

48 Nuclear Power Plants: 2019 atw Compact Statistics

Editorial

Contents

atw Vol. 65 (2020) | Issue 1 ı January

Nuclear Fusion / Revived € 20 Billion Iter

Project ‘Entering a Critical Phase’

INSIDE NUCLEAR WITH NUCNET

Success will represent a breakthrough

that could secure clean, safe energy for millions of years

The International Thermonuclear Experimental Reactor (Iter) under construction at Cadarache in southern France is

more than 65 % complete and entering a critical phase as it aims to meet a first plasma deadline of 2025, project head

Bernard Bigot told NucNet.

First plasma means that the reactor is able to successfully

generate a molten mass, 840 m 3 to be exact, of electricallycharged

gas, or plasma, inside its core.

For the next three years the focus is getting all main

components for the fusion reactor in place, Mr Bigot said,

adding that “there is a lot of pressure”. Some components

weigh up to 500 tonnes and making sure they are delivered

on time and fit as they should is a huge challenge.

Mr Bigot, who earlier this year was appointed to a

second five-year term as director-general of the Iter Organisation,

extending his tenure to March 2025, confirmed

that the budget for the project, at 2016 prices, is € 20 bn.

“It is my deep belief this project is needed and will

work,” he said. “The world needs it. Fossil fuels will be

depleted over the coming century and we need to find a

replacement.”

“As a scientist I have been looking for this technology

for several decades. If we succeed it will be a real breakthrough

for the world’s energy, not only in this century but

for millions of years.”

Fusion is the fundamental energy of the universe,

perpetually powering the sun and stars. The desire to

recreate and control this atomic energy on earth is the

driving force behind Iter.

Iter – meaning “the way” in Latin – will be the world’s

largest fusion experiment. The steel and concrete superstructures

nestled in the hills of southern France will house

a 23,000-tonne machine, known as a tokamak, capable of

creating what is essentially an earthbound star. The

tokomak building, into which the tokomak itself will be

placed, will be available from March 2020, Mr Bigot said.

Scientists will heat a ring-shaped vacuum chamber to

150 million (10 6 ) °C, 10 times hotter than the sun’s core.

Inside this chamber two types of hydrogen atoms will collide

with enough force to fuse in a superheated plasma at

the highest temperatures in the universe.

This “atomic soup” will be kept suspended away from the

reactor walls using the force field of a magnet cage created

by a coil of the world’s most powerful magnets. To withstand

the heat these will be supercooled to the temperature

of deep space, near absolute zero or minus 273 °C.

Building a structure to contain mankind’s most

advanced scientific experiment requires the combined

efforts of more than 30 countries and many thousands of

scientists from Iter’s core members: the European Union,

China, India, Japan, South Korea, Russia and the US.

Mr Bigot took to the helm of Iter four years ago, tasked

with rescuing the project as it became beset by delays and

spiralling costs.

“We have seen a lot of change,” Mr Bigot said. “ Everyone

is now complying with new best standards for project management.

The atmosphere of the project has completely

changed too. People believe that fusion is on track. Before

it was almost a dream.”

By 2025 they expect to start the first milestone

experiments to prove that fusion technology can produce

10 times more energy than it uses. The challenge ahead

lies in keeping the contributions of 35 countries, and

500 companies, carefully aligned to its schedule. This must

be followed by painstaking assembly of the component

parts on site in France.

What is Fusion?

Fusion is the same process involved in powering the sun

and other stars in our universe. Energy is produced by

fusing together light atoms, such as hydrogen, at the

extremely high pressures and temperatures. These

particular conditions are present in the sun’s core, delivering

temperatures of up to 15 million °C.

The extremely high temperatures can transfer a gas into

a state of plasma, which is essentially an electricallycharged

gas. Although plasma is rarely found on Earth, it is

thought that more than 99 % of the universe exists as

plasma.

To replicate this process on Earth, gases need to be heated

to extremely high temperatures of about 150 million °C

at which point atoms become completely ionised.

The easiest method for this type of fusion reaction is

with two hydrogen isotopes: deuterium, extracted from

water, and tritium, produced during the fusion reaction

through contact with lithium.

When deuterium and tritium nuclei fuse, they form a

helium nucleus, a neutron and a lot of energy.

The Tokamak

The Iter Tokamak will weigh 23,000 tonnes and be 60 m in

height. In a Tokamak the plasma is held in the looping

structure. Using coils, a magnetic field is created that

causes the plasma particles to oribit in spirals, without

making contact with the chamber walls.

The neutron has no electrical charge and is unaffected

by the magnetic fields, allowing them to move away from

the bond of the plasma.

The neutrons are then absorbed by the surrounding

walls transferring their energy into heat and generating

steam from pools of water.

Author

NucNet

The Independent Global Nuclear News Agency

Editor responsible for this story: Kamen Kraev

Secretary General, NucNet

Avenue des Arts 56 2/C

1000 Bruxelles

www.nucnet.org

Inside Nuclear with NucNet

Nuclear Fusion / Revived € 20 Billion Iter Project ‘Entering a Critical Phase’

atw Vol. 65 (2020) | Issue 1 ı January

Feature | Energy Policy, Economy and Law

Energy Supply Without Nuclear:

Winter 2022/23 is Coming

Roman Martinek

Only three and a few years are left in Germany before the scheduled shutdown of the country’s last nuclear power

plants: by December 31, 2022, Isar, Neckarwestheim and Emsland NPPs (one reactor at each) will be disconnected from

the grid.

Of course, it cannot actually be argued that this event will

mark the end of the atomic age in Germany – as research

reactors and supporting industry enterprises (for example,

uranium enrichment plants) continue to run smoothly and

the issue of final disposal of nuclear waste still remains to

be solved. And yet, giving up nuclear energy for electricity

generation in Europe’s leading economy will be a milestone

in itself.

As this date draws closer, it is curious to observe the

melting confidence that the decision taken in the spring of

2011 to accelerate the nuclear phase-out was strategically

reasonable. Without a doubt, considered per se, this step is

quite feasible from a technical point of view – there could

hardly arise any problem with shutting down several

nuclear reactors (many reactors are routinely dis connected

from the grid from time to time for scheduled maintenance).

How calibrated this decision is with regard to the future

energy supply in the country, is a different and increasingly

resonating question asked by politicians who can no longer

be branded as solely right-radical adepts of the Alternative

for Germany.

For example, this summer, the so-called “Union of

Values” within the Christian Democratic Union (CDU)

called for an extension of the service life of existing NPPs,

while at the same time blaming the party’s leadership for

an insufficiently determined climate policy. Politicians

argued their position with the threat that is seemingly

becoming ever more real threat that Germany will not

achieve its climate targets under the Paris Agreement, as

well as with rising electricity prices. By postponing the

nuclear phase-out, the German government could give up

coal in a more visible time, representatives of the Union

said.

Further, in September, Peter Hauk, Minister of Agriculture

of the Federal State of Baden-Württemberg, took

the floor. He expressed a similar idea: in his opinion, a

discussion is needed on how feasible it could be to quit coal

ten years ahead of schedule. A measure that could ensure

the implementation of this idea into reality, according to

Hauk, could be extending the service life of the reactors at

Neckarwestheim and Philippsburg for the same ten years.

Meanwhile, the politician did not fail to point out the lack

of political will, noting that while nuclear energy is

generally off the agenda, some compromise was still

reached at the political level regarding coal. “This is a

mistake, because the climate goals of the federal and state

government will not be achieved,” Hauk lamented.

Alarmed signals are also emanating from the industry:

back in early 2019, Alfred Gaffal, President of the

Association of the Bavarian Economy (VBW) drew

attention to the threatening deficit of the region’s own

electricity capacities that would hit the Bavarian economy,

and said that “if there are no other options left, the issue of

service life extension of nuclear reactors in Bavaria cannot

be taken off the table”.

In this light, recent reports indicating that Bavaria’s

own electricity generation will certainly not be sufficient to

meet the region’s needs imply that Bavaria will be forced to

import electricity after 2022. The question is where this

electricity will be supplied from – the pace of electricity

grid expansion, which, as planned, should ensure the

uninterrupted transmission of excess electricity from the

north to the south of the country, is noticeably stalled. The

threat seems quite real that Bavaria will be greeting the

winter season of 2022/23 without an answer needed this

badly.

Meanwhile, in 2011, the state’s Prime Minister Horst

Seehofer promised that even after the shutdown of

Bavarian NPPs the region would still be capable of

providing itself with electricity on its own. Now it seems

that Bavaria is very, very far from this goal. In addition to

the hampered development of power grid infrastructure,

there are noticeable protests against windmills construction.

Besides, gas power plants that regional

politicians doubled down on a little more than eight years

ago have proved to be overly expensive under the current

market conditions. The “Energiewende” (i.e. German

energy transition) seems to be close to a dead end – at least

in Bavaria.

FEATURE | ENERGY POLICY, ECONOMY AND LAW 9

Feature

Energy Supply Without Nuclear: Winter 2022/23 is Coming ı Roman Martinek

atw Vol. 65 (2020) | Issue 1 ı January

FEATURE | ENERGY POLICY, ECONOMY AND LAW 10

Peter Hauk is certainly not the only high-profile

politician who has questioned the advisability of nuclear

phase-out until 2022. The European Commissioner for

Budget, CDU member Gunter Oettinger believes that after

2022, Germany will have to import electricity from foreign

NPPs for rather a long time: “Thus, an automobile in

Karlsruhe will drive eco-friendly on nuclear electricity

from France”. In this connection, a logical question arises:

does it make sense at all to shut down NPPs for some

( apparently ideologically-motivated) dislike of nuclear

energy, if this step forces imports of electricity from NPPs

that are different only in that they are not in Germany?

Finally, in late November, Prime Minister of North

Rhine-Westphalia Armin Laschet said: “If we assume that

CO 2 emissions and climate change present the most serious

issues we should take care of, the order of phasing out

nuclear and coal energy was chosen incorrectly“. According

to Laschet, Germany should have first quit coal, instead

of giving up nuclear.

As is evident from the above positions, the question of

which energy sources will be used in Germany for future

national electricity supply is closely intertwined with the

issue of compliance with the Paris Agreement goals and

achievement of targets that the German government

committed itself to. On the one hand, it is quite obvious

that the current state and volumes of renewable energy

generation will not allow to replace the outgoing nuclear

power capacities with solar and wind energy immediately

after 2022. In other words, it is highly probable that one

should expect an increase in the use of coal – the source of

energy that could be used until 2038, which is almost

20 years from now.

On the other hand, most environmental organizations

and their activists, who invariably point this out, are still

not ready to admit that nuclear power could be very helpful

in combating climate change. In fact this would mean

giving up one of the central postulates of the modern green

ideology. It is thus unsurprising that the sensational March

statement of Greta Thunberg, the “icon” of the Fridays for

Future movement, that nuclear power could become an

element in the greater CO 2 -free energy balance of the

future, is today presented almost as a slip of the tongue.

In the meantime, the solution to climate problems can

be found in modern technologies, including nuclear

energy – which is indicated by many experts, including

Tristan Horx from the Institute for the Future (Zukunftsinstitut),

a non-profit analytical center.

“Although I support the Fridays for Future activities and

watch the current environmental agenda with interest,

I do not really welcome the statements that technological

development is harmful to the environment and that our

planet is doomed if we do not return to living in wooden

huts and riding exclusively a bicycle”, the expert argues.

“I believe in the innovative potential of humanity and the

ability to find a solution to existing problems”.

As a transition technology that is able to offer a solution

to energy sector issues, including the coal use issue,

nuclear power is an excellent option, says Horx. “It

contributes to the reduction of total CO 2 emissions, and

this is what many experts confirm. However, it is impossible

to want the world to remain green and at the same time

frankly demonize a perfectly functioning energy source – it

does not work like this. Coal energy, in my opinion, carries

a lot more problems in comparison with nuclear. However,

if you talk about it with the Greens, most of them will be

telling you that nuclear power is the worst thing we have at

all. But this is simply not true today”.

Instead of the traditional “green” ecology concept,

which calls for abstinence, reduction and in every possible

way positions human as the planet’s parasite, Horx favors

a bit different approach: “An approach which does not

imply that we have to sacrifice technological development

– along with innovative technologies, we will be using the

technologies that we already have – including, for example,

nuclear energy”.

Authors

Roman Martinek

Expert for Communication

Czech Republic

Feature

Energy Supply Without Nuclear: Winter 2022/23 is Coming ı Roman Martinek

atw Vol. 65 (2020) | Issue 1 ı January

Dual-Use-Verordnung im Trilog

Ulrike Feldmann

Vor zwei Jahren wurde an dieser Stelle über die unendlich erscheinende Geschichte der Revision der Verordnung (EG)

Nr. 428/2009 über die Kontrolle der Ausfuhr, der Verbringung, der Vermittlung und der Durchfuhr von Gütern mit

doppeltem Verwendungszweck (im folgenden: Dual-Use-Verordnung) berichtet (atw 1 (2018) S. 19). Nunmehr ist das

Revisionsverfahren in ein neues Stadium, das „Trilog“-Verfahren, eingetreten.

Hintergrund

Hintergrund für die erneute Revision der Verordnung ist

ein verändertes technologisches und sicherheitspolitisches

Umfeld (Moderne Überwachungs- und Hacking-Technologien,

die zu Menschenrechtsverletzungen eingesetzt

werden können, sowie gesteigerte Terrorgefahr), dem die

EU-Kommission mit ihrem Vorschlag Rechnung tragen

will. Gleichzeitig soll für die europäische Industrie ein

handelspolitisches Umfeld geschaffen werden, in dem die

EU-Industrie unter Wettbewerbsbedingungen antreten

kann, die mit in Drittstaaten geltenden Wettbewerbsbedingungen

vergleichbar sind („level playing field“).

Lange Zeit dümpelten die Beratungen im Rat der EU dahin.

Eine Einigung der Mitgliedstaaten auf eine gemeinsame

Position zu dem Vorschlag der EU-Kommission schien

nicht in Sicht.

Am 05.06.2019 fand der Rat aber dann schließlich doch

zu einer gemeinsamen Position und konnte sein Mandat für

die Verhandlung mit der EU-Kommission und dem Europäischen

Parlament (EP) annehmen. Der Text der Position

des Rates ist mit dem Verhandlungsmandat auf der

Homepage des Rates, https://www.consilium.europa.eu,

veröffentlicht.

Zum Vergleich: Die strittigsten Regelungen im

Kommissionsentwurf und in der Position des EP

Wie erinnerlich betreffen die strittigsten Regelungen im

Vorschlag der EU-Kommission Cyber-Überwachungstechnologien

und den Schutz von Menschenrechten. Die

EU-Kommission möchte den Export von Technologien

stärker kontrollieren, wenn das Risiko besteht, dass diese

Technologien zur Überwachung von Menschen genutzt

werden können (Stichwort: Arabischer Frühling). Die

Cyber-Überwachungstechnologien sollen nach dem

Vorschlag der EU-Kommission als eigener, neuer Typus

eines Dual-Use-Gutes in die revidierte Fassung der Dual-

Use-Verordnung aufgenommen werden. Eine „Catch-All“

Klausel zum Schutz der Menschenrechte soll für alle nicht

bereits gelisteten Güter eingeführt werden, die möglicherweise

einen negativen Einfluss auf Versammlungs- und

Vereinigungsfreiheit, Recht auf freie Meinungsäußerung

sowie das Recht auf Privatsphäre haben können.

Diese Vorschläge wurden und werden vom EP prinzipiell

unterstützt. Allerdings lehnte das EP im Plenum

in seinen zahlreichen Änderungsvorschlägen eine Erweiterung

der Exportkontrolle auf Terrorabwehr ab und

machte Vorschläge zur Präzisierung der Vorschriften

zum Menschenrechtsschutz. Der EP-Ausschuss für internationalen

Handel (INTA) unter ihrem Berichterstatter

Prof. Dr. Klaus Buchner wollte dagegen sogar die Exportkontrollen

noch stärker als im Kommissionsvorschlag

ausweiten und die Menschenrechte zum zentralen

Anliegen der Exportkontrolle machen. Die Argumente,

die gegen eine solche Ausweitung der Exportkontrolle

bestehen, sind im SoNL-Beitrag in Heft 1 der atw 2018

nachzulesen (u. a. Überforderung der Unternehmen, den

Stand von Menschenrechtsstandards in den verschiedenen

Ländern nachzuprüfen und zu validieren).

Die Position des Rates

Diesen Bestrebungen hat nun im Sommer der Rat mit

seinem Verhandlungsmandat eine Absage erteilt. Eine

„Catch-all“-Klausel geht den EU-Mitgliedstaaten zu weit.

Der Rat lehnt es ab, den Menschenrechtsschutz und die

Terrorabwehr auf die Unternehmen zu verlagern, sondern

sieht weiterhin darin eine originäre Staatsaufgabe.

Nach dem einstimmig beschlossenen Ratsmandat sollen

jedoch die Mitgliedstaaten ähnlich wie im Kom missionsvorschlag

die Möglichkeit erhalten, auf nationaler Ebene

eine Selbstkontrolle der Unternehmen einzuführen. Haben

die Unternehmen berechtigte Gründe für die Annahme

(Verdachtsmomente), dass das Exportgut militärisch genutzt

werden könnte, sollen sie verpflichtet werden können,

eine Genehmigung zu beantragen.

Darüber hinaus sollen Unternehmen auf EU-Ebene

verpflichtet werden, gegenüber der Behörde eine Endverbleibserklärung

abzugeben, wobei allerdings den

Mitgliedstaaten zugestanden wird, Aus nahmen von dieser

Pflicht zu machen. (Der Kommissionsentwurf lehnt diese

und andere nationalen Öffnungs klauseln im Sinne einer

europäischen Harmonisierung ab).

Zu der vielfach diskutierten Einführung interner Kontroll

programme („Internal Compliance Programmes“/

ICPs), die Kommission und EP befürwortet hatten, hat sich

der Rat in seinem Mandat so positioniert, dass er eine entsprechende

Regelung auf EU-Ebene ablehnt, es jedoch den

Mitgliedstaaten überlässt, derartige ICPs vorzuschreiben.

Ferner befürwortet der Rat die Einführung neuer EU-

Allgemeingenehmigungen, wobei es auch im Rat keine

Mehrheit gab, eine EU-Allgemeingenehmigung für nicht

sensitive Nukleargüter einzuführen, was die Nuklearbranche

nachdrücklich angeregt hatte.

Das Trilog-Verfahren

Mit dem Vorliegen der gemeinsamen Ratsposition konnte

inzwischen das Trilog-Verfahren eröffnet werden, in dem Rat,

EP und EU-Kommission versuchen müssen, sich auf einen

endgültigen Revisionsentwurf zu einigen. Als ziemlich sicher

gilt bereits, dass mit für die Nuklearbranche positiven Änderungen

des Annex IV zur Dual-Use-Verordnung nicht mehr zu

rechnen ist. Als positiv bei den Trilog- Verhandlungen darf

gewertet werden, dass sie auf der Grundlage der Rats position

geführt werden und nicht etwa auf der Grundlage der EP-

Position. Die amtierende finnische Präsidentschaft hatte den

Willen bekundet, bis Ende des Jahres das Revisions verfahren

abgeschlossen zu haben, was angesichts der teilwei se doch

sehr weit auseinandergehenden Positionen von Rat, Kommission

und EP und dem zeitlich eher aufwendigen Trilog-Verfahren

von vorneherein recht ambitioniert erschien. Nach der

jüngsten Trilog-Sitzung am 28.11.2019 zeichnet sich nunmehr

deutlich ab, dass das Thema mindestens noch unter der

kroatischen Ratspräsidentschaft wird fortgeführt werden

müssen. Zu der Frage neuer EU-Allgemeingenehmigungen

gibt es beispielsweise bislang noch keinen Konsens. An der

„unendlichen“ Geschichte der Revision der EG Dual-Use-

Verordnung 428/2009 wird also noch weiter geschrieben.

SPOTLIGHT ON NUCLEAR LAW

Spotlight on Nuclear Law

Dual-Use Act in Trialog ı Ulrike Feldmann

atw Vol. 65 (2020) | Issue 1 ı January

Calendar

2020

CALENDAR

10.02. – 14.02.2020

37 th Short Courses on Multiphase Flow. Zurich,

Switzerland, Swiss Federal Institute of Technology

ETH, www.lke.mavt.ethz.ch

10.02. – 14.02.2020

ICONS2020: International Conference on Nuclear

Security. Vienna, Austria, The International Atomic

Energy Agency (IAEA), www.iaea.org

12.02. – 13.02.2020

7 th Nuclear Decommissioning & Waste

Management Summit 2020. London, UK, ACI,

www.wplgroup.com

18.02. – 20.02.2020

GEN IV International Forum. Boulogne-Billancourt,

France, www.snetp.eu

02.03. – 03.03.2020

Forum Kerntechnik. Berlin, Germany, VdTÜV & GRS,

www.tuev-nord.de

02.03. – 06.03.2020

International Workshop on Developing a

National Framework for Managing the Response

to Nuclear Security Events. Madrid, Spain, IAEA,

www.iaea.org

08.03. – 12.03.2020

WM Symposia – WM2019. Phoenix, AZ, USA,

www.wmsym.org

08.03. – 13.03.2020

IYNC2020 – The International Youth Nuclear

Congress. Sydney, Australia, IYNC, www.iync2020.org

15.03. – 19.03.2020

ICAPP2020 – International Congress on Advances

in Nuclear Power Plants. Abu-Dhabi, UAE, Khalifa

University, www.icapp2020.org

18.03. – 20.03.2020

12. Expertentreffen Strahlenschutz. Bayreuth,

Germany, TÜV SÜD, www.tuev-sued.de

22.03. – 26.03.2020

RRFM – European Research Reactor Conference.

Helsinki, Finland, European Nuclear Society,

www.euronuclear.org

25.03. – 27.03.2020

H2020 McSAFE Training Course. Eggenstein-

Leopoldshafen, Germany, Karlsruhe Institute of

Technology (KIT), www.mcsafe-h2020.eu

29.03. – 02.04.2020

PHYSOR2020 — International Conference on

Physics of Reactors 2020. Cambridge, United

Kingdom, Nuclear Energy Group,

www.physor2020.com

31.03. – 02.04.2020

4 th CORDEL Regional Workshop on

Harmonization to support the Operation and

New Build fo NPPs including SMRs. Lyon, France,

NUGENIA, www.nugenia.org

30.03. – 01.04.2020

INDEX International Nuclear Digital Experience.

Paris, France, SFEN Société Française d’Energie

Nucléaire, www.sfen-index2020.org

31.03. – 03.04.2020

ATH'2020 – International Topical Meeting on

Advances in Thermal Hydraulics. Paris, France,

Société Francaise d’Energie Nucléaire (SFEN),

www.sfen-ath2020.org

19.04. – 24.04.2020

International Conference on Individual

Monitoring. Budapest, Hungary, EUROSAFE,

www.eurosafe-forum.org

20.04. – 22.04.2020

World Nuclear Fuel Cycle 2020. Stockholm,

Sweden, WNA World Nuclear Association,

www.world-nuclear.org

05.05. – 06.05.2020

KERNTECHNIK 2020.

Berlin, Germany, KernD and KTG,

www.kerntechnik.com

10.05. – 15.05.2020

ICG-EAC Annual Meeting 2020. Helsinki, Finland,

ICG-EAC, www.icg-eac.org

11.05. – 15.05.2020

International Conference on Operational Safety

of Nuclear Power Plants. Beijing, China, IAEA,

www.iaea.org

12.05. – 13.05.2020

INSC — International Nuclear Supply Chain

Symposium. Munich, Germany, TÜV SÜD,

www.tuev-sued.de

17.05. – 22.05.2020

BEPU2020– Best Estimate Plus Uncertainty International

Conference, Giardini Naxos. Sicily, Italy,

NINE, www.nineeng.com

18.05. – 22.05.2020

SNA+MC2020 – Joint International Conference on

Supercomputing in Nuclear Applications + Monte

Carlo 2020, Makuhari Messe. Chiba, Japan, Atomic

Energy Society of Japan, www.snamc2020.jpn.org

20.05. – 22.05.2020

Nuclear Energy Assembly. Washington, D.C., USA,

NEI, www.nei.org

31.05. – 03.06.2020

13 th International Conference of the Croatian

Nuclear Society. Zadar, Croatia, Croatian Nuclear

Society, www.nuclear-option.org

06.06. – 12.06.2020

ATALANTE 2020. Montpellier, France, CEA,

www.atalante2020.org

07.06. – 12.06.2020

Plutonium Futures. Montpellier, France, CEA,

www.pufutures2020.org

08.06. – 12.06.2020

20 th WCNDT – World Conference on

Non-Destructive Testing. Seoul, Korea, EPRI,

www.wcndt2020.com

15.06. – 19.06.2020

International Conference on Nuclear Knowledge

Management and Human Resources Development:

Challenges and Opportunities. Moscow,

Russian Federation, IAEA, www.iaea.org

15.06. – 20.07.2020

WNU Summer Institute 2020. Japan, World Nuclear

University, www.world-nuclear-university.org

02.08. – 06.08.2020

ICONE 28 – 28 th International Conference on

Nuclear Engineering. Disneyland Hotel, Anaheim,

CA, ASME, www.event.asme.org

01.09. – 04.09.2020

IGORR – Standard Cooperation Event in the International

Group on Research Reactors Conference.

Kazan, Russian Federation, IAEA, www.iaea.org

09.09. – 10.09.2020

VGB Congress 2020 – 100 Years VGB. Essen,

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

09.09. – 11.09.2020

World Nuclear Association Symposium 2020.

London, United Kingdom, WNA World Nuclear

Association, www.world-nuclear.org

16.09. – 18.09.2020

3 rd International Conference on Concrete

Sustainability. Prague, Czech Republic, fib,

www.fibiccs.org

16.09. – 18.09.2020

International Nuclear Reactor Materials

Reliability Conference and Exhibition.

New Orleans, Louisiana, USA, EPRI, www.snetp.eu

28.09. – 01.10.2020

NPC 2020 International Conference on Nuclear

Plant Chemistry. Antibes, France, SFEN Société

Française d’Energie Nucléaire,

www.sfen-npc2020.org

28.09. – 02.10.2020

Jahrestagung 2020 – Fachverband Strahlenschutz

und Entsorgung. Aachen, Germany, Fachverband

für Strahlenschutz, www.fs-ev.org

12.10. – 17.10.2020

FEC 2020 – 28 th IAEA Fusion Energy Conference.

Nice, France, IAEA, www.iaea.org

26.10. – 30.10.2020

NuMat 2020 – 6 th Nuclear Materials Conference.

Gent, Belgium, IAEA, www.iaea.org

09.11. – 13.11.2020

International Conference on Radiation Safety:

Improving Radiation Protection in Practice.

Vienna, Austria, IAEA, www.iaea.org

24.11. – 26.11.2020

ICOND 2020 – 9 th International Conference on

Nuclear Decommissioning. Aachen, Germany,

AiNT, www.icond.de

07.12. – 10.12.2020

SAMMI 2020 – Specialist Workshop on Advanced

Measurement Method and Instrumentation

for enhancing Severe Accident Management in

an NPP addressing Emergency, Stabilization and

Long-term Recovery Phases. Fukushima, Japan,

NEA, www.sammi-2020.org

17.12. – 18.12.2020

ICNESPP 2020 – 14. International Conference on

Nuclear Engineering Systems and Power Plants.

Kuala Lumpur, Malaysia, WASET, www.waset.org

This is not a full list and may be subject to change.

Calendar

atw Vol. 65 (2020) | Issue 1 ı January

Analysis of Ultimate Response Guidelines

for Chinshan Nuclear Power Plant

in Taiwan to Cope with Postulated

Compound Accident

Jieqing Zheng

Taiwan Power Company (TPC) together with its engineering consultation, research company and institute have

been working on the development of guidelines for the compound accident which was caused by the nature disaster of

a combination of seismic and tsunami events occurred in Fukushima, Japan. As a result, Ultimate Response Guidelines

(URGs) for Chinshan Nuclear Power Plant (NPP) in northern Taiwan have been developed. This paper provides highlight

of the features for URGs developed by TPC and successfully demonstrated that at least 127 gpm cooling water is needed

using MAAP5 if the peak cladding temperature (PCT) is maintained below 1088.6 K (1500 °F). On the other hand, when

the injecting timing is delayed, the fuel rods in the core will overheat and generate substantial amount of hydrogen, and

the plant has a high risk that rising levels of hydrogen inside the containment could cause a blast.

1 Introduction

After the Fukushima nuclear accident

in Japan, concerns have been raised

to examine the previously existed

emergency operating procedures

(EOPs) and severe accident management

guidelines (SAMGs). It’s found

that they may not be adequate to deal

with the compound accident [1-3].

The MAAP5 code has been used as

a tool to evaluate the execution of

URGs in compound accident [4]. The

development of URGs for compound

accident beyond that of design basis is

necessary to ensure the health and

safety of people at and surround

the plant site [5-8]. The Chinshan

NPP, which possesses a boiling water

reactor (BWR) the same as Fukushima

NPP and includes many safety features

in its design, was chosen in this

study. The objective of this paper is

to simulate the station blackout

accident caused by compound accident

( CASBO) and investigate how

the execution of URGs could mitigate

the accident process.

2 Chinshan ultimate

response guidelines

Chinshan URGs was first developed by

TPC in 2011 to supplement EOPs and

SAMGs for plant under compound

accident conditions [8]. It has high

possibility that all the emergency core

cooling system (ECCS) will be out

of work when compound accident

happens, so plant-specific bases shall

be used for initiation of URGs and

for taking subsequent actions. TPC

especially puts emphases on possessing

manoeuvrability and shortenning

the response time to cope with all

possible situations.

Timing for initiation of URGs was

estalished according to one of the

following three conditions. The first

condition is loss of makeup water to

the reactor vessel to maintain the

covering of the fuel rods by water. The

second condition is that on-site and

off-site AC powers have already lost.

The third condition is scram of reactor

due to severe seismic event con current

with the announcement of oncoming

tsunami by the Central Weather

Bureau. As shown in Figure 1, the

URGs will be site-specificly used for

Chinshan NPP.

When entrying the Chinshan URGs,

there are 3 stages to be gradually

initiated, as shown in Table 1. Under

normal circumstances, the plant status

will recover in time throughout these

strategies. The strategy should be

performed synchronously in the same

phase and must be done as soon as

possible. If phase1 has been successfully

executed, then the operators

perform phase2 and phase3. As a

result, long-term cooling will be

established to prevent reactor core

from being damaged. Once the worst

situation that has the same complexity

as Fukushima event happens, the

target of phase1 strategy can not be

reached. When RCIC turbine pump is

tripped off and the electrical power

cannot be recovered, any water available

should be injected into the

Chinshan RPV as soon as possible.

To strengthen memorization of the

actions for the plant operators to be

taken to implement URGs, DIVING,

such as the term used in submarine

under attack, was adopted as an

abbreviation. DIV means depressurization,

water injection, vent, respectively,

and ING means acting simultaneously.

The decision-making mechanism

of the plant to decide the timing to

inject raw water or sea water into the

core or spent fuel pool is the most

important part of the URGs. Once the

relatively non-purified water is used

for coolant injection to prevent overheating

of fuel rods to ensure the

safety and health of people, it’s unlikely

to use the CSNPP again for

| Fig. 1.

URGs flowchart.

ENVIRONMENT AND SAFETY

Environment and Safety

Analysis of Ultimate Response Guidelines for Chinshan Nuclear Power Plant in Taiwan to Cope with Postulated Compound Accident ı Jieqing Zheng

atw Vol. 65 (2020) | Issue 1 ı January

ENVIRONMENT AND SAFETY 14

Phase Target Timing Strategy

Phase1

Phase2

Phase3

mitagate and

control the event

recover

the power

establish

long-term cooling

| Tab. 1.

CSNPP action strategies.

within

1 hour

within

8 hours

within

36 hours

subsequent power generation without

tedious clean-up work. The levels of

authorization should be done as the

following. The plant manager informs

the chairperson of the Emergency

Plan Execution Committee and

executes the plan after obtaining

consent from the chairperson. If

communication to the chairperson is

not available, then the plant manager

is authorized to implement the

URGs. If communication to the plant

manager is not available, then the

supervisor on duty is authorized to

implement the URGs.

Except for scheduled (twice per

year) drills of the operators on duty,

the minimum water injection rate will

be calculated by MAAP5 code .

3 Assumptions used

for the analysis

To perform the analysis of the effectiveness

of URGs, there are some

initial assumptions adopted in this

study:

(1) At time zero, a strong seismic

event takes place and the reactor is

scrammed.

(2) The Chinshan NPP loses all the

on-site and off-site AC power.

(3) RCIC comes on when the reactor

water level reaches L2.

(4) RCIC becomes unavailable at 20

minutes from the start due to the

fact that the tsunami hits the plant

then.

(5) URGs are initiated due to the fact

that the plant loses all injection

water to the core.

(6) Raw water or firewater becomes

available to inject water into the

core in one hour after the initiation

of the compound accident.

1. inject raw-water or fire-water or water from the nearby

creek/sea into the reactor vessel

2. depressurize the reactor vessel (SBO)

3. vent the containment (SBO)

4. connect pipes to fire engine to inject water

(raw water; fire water; creek water)

5. activate RCIC manually

6. power supply to the two reactors

by the fifth diesel generator

7. power supply to the two reactors

8. by turbine-driven diesel generators

1. movable air compressor/nitrogen bottles

to provide gas to SRV/ADS

2. connect to 480 V manoeuvrable diesel-generator

3. connect to 4.16 kV power cart

4. extend the duration of DC power supply

5. add water to the spent fuel pool

6. draining operation of the submerged pump

7. inject water into CST by manoeuvrable water source

1. remove trash at the emergency water inlet

2. replace emergency service water (ESW) motor

3. provide alternate long-term cooling

Whenever the on-site and off-site

power is unavailable, emergency

depressurization has to be gradually

performed by operating safety/relief

valves. In the same time, raw water

injection line also has been prepared.

These measures are all completed

within one hour after the initiation of

the event. If the plant status cannot

recover in time, any water available

will be injected into the reactor vessel.

For the sensitivity studies on the water

injection rates, 125 gpm, 150 gpm,

and 250 gpm are used.

4 Results and discussion

4.1 Simulation without URGs

being implemented

The accident is initiated by a strong

seismic event followed by loss of all AC

power, including the onsite and offsite

power. As a result, the high-pressure

injection system (HPFL) and the low

pressure flooder (LPFL) fail, and RCIC

is the only system which is available to

mitigate the consequences of compound

accidents. The assumption has

been adopted with some extremes.

Taking Fukushima as an example,

the emergency generator had been

working for an hour when the off-site

power failed and RHR had performed

to cool the suppression pool. Comparing

to the Fukushima accident, the

result simulated in this study with

MAAP5 is conservative.

The details of simulation sequences

are illustrated in Table 2. Initially,

AC power is lost, followed by MSIV

closure, CRD and feedwater being not

available. After the reactor scrammed

at 4.2 s, the power of the RPV rapidly

drops to decay power which is 2.9

percent of the rated power, as shown

in Figure 2. When the core water level

reaches L2 that is a signal to initiate

the RCIC, the turbine and pump of

RCIC activate to suct cooling water

from the condensate storage tank

(CST). The water level has been

maintained between L2 to L8 . All of

ECCS system fails when RCIC cannot

inject water into the RPV 20 minutes

later, and the level of water decreases

rapidly because of boiling off, as

shown in Figure 3. With MSIV closing,

the pressure in the RPV quickly rises

up to SRVs setpoint so that the initial

trend of PPS (that is, the pressure

in the primary system) is cycling

(Figure 4). Because the loss of all

ECCS, cladding of the fuel rods begins

to heat up. Its temperature reaches

1088.6 K at 5500 s, which is an

im portant temperature to decide

whether the reactor is safe or not. If

Number Time(s) Events Remark

1 0

Loss of all AC power

HPCS locked off

LPCI loop locked off

2 4.2 Reactor scramed L3

3 50 RCIC on L2

4 1,200 RCIC turbine pump tripped Power unavailable

5 4,500 Core uncovered

6 5,406 Hydrogen generated

7 5,500 Tcl max reached a critical point 1088.6 K

8 6,606 Core melted down

| Tab. 2.

Time sequences for the simulation case without using URGs.

Water level at 8.89 m

above the bottom of

the vessel

9 12,611 Core relocated Core support plate fail

10 21,416 RPV failed

11 132,621 COPS activated

12 172,800 Simulation ended

Environment and Safety

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atw Vol. 65 (2020) | Issue 1 ı January

| Fig. 2.

Core power response.

| Fig. 3.

Core water level response.

| Fig. 4.

RPV pressure response.

ENVIRONMENT AND SAFETY 15

| Fig. 5.

Debris mass response.

| Fig. 6.

Wetwell pressure response.

| Fig. 7.

Core water level response.

the cladding temperature is above

1088.6 K, the zirconium-water reaction

will become very intense and

emit a lot of heat and hydrogen to

increase the potential of explosion for

the secondary containment. The core

support plate fails at 12,611 s, and,

subsequently, molten corium relocates

to the lower plenum region of

the reactor pressure vessel (RPV).

As shown in Figure 5, the mass of

the molten fuel bundles and channel

boxes totally has a weight of

155,475 kg; in fact, these two

materials actually have 107,000 kg.

The main reason is that the melt

contains the other reactor components

falling into the lower plenum,

such as the fuel rods support plates

and the core shroud. After the bottom

of the vessel fails at 21,416 s, debris

drops to the lower cavity, which means

that the boundary of RPV has been

breached. The falling debris contacts

with the bottom of the container and

causes further chemical reaction,

releasing large amount of energy,

steam, and non-condensable gas,

which gradually increases the temperature

and pressure of the drywell.

Because that the RHR (residual heat

removal) systems are not available,

COPS (containment overpressure

protection system) finally activates at

132,621 s due to the fact that the wet

well is over-pressurized, as shown in

Figure 6. Radioactive material CsI and

CsOH will begin to increase to be

released to the environment after

COPS activates, and then decreases.

The radioactive material is generally

on the magnitude of 10-5 because of

the scrubbing effect of the suppression

pool. By the end of the simulation,

a total of 50 kg of hydrogen is

produced.

4.2 Simulation with URGs

addition

Three different water injection rates

are assumed in this simulation, but

the behaviors of the progression of

events are similar: (1) After the

reactor scrams at 4.2 s, the power of

the RPV rapidly drops to decay power

which is 2.9 percent of the rated

power. (2) The water level is maintained

between L2 and L8 (that is,

12.065 m to 14.622 m above the top of

the fuel rods) within 20 minutes. (3)

The water injection rates of high

pressure core injection and low

pressure core injection remain

unavail able from time zero.

According to the exercises of TPC,

the fastest time between informing

operators for taking the ultimate actions

and implementing pipe hookups

for injecting water into the reactor is 1

hour. After all pre parations are done,

the emergency depressurization is

performed to make raw water/firewater

operable by opening 5 SRVs. Response

of the reactor core water level

is illustrated in Figure 7. Three different

injection rates used in this study

can all result in the core water level to

be at safe position, which fluctuates

between L2 and L8. It is obvious that

the higher the firewater injection rate,

the faster the core water level will get

to the safe position. Comparing the

time it takes to reach that safe state,

the time for the case with water injection

rate of 125 gpm is calculated to be

26,414 s later than that for the case

with 250 gpm. Therefore greater

amount of raw water or firewater will

be required to restore the core water

evel.

As shown in Figure 8, the peak

cladding temperature for the case

with water injection rate of 125 gpm

Environment and Safety

Analysis of Ultimate Response Guidelines for Chinshan Nuclear Power Plant in Taiwan to Cope with Postulated Compound Accident ı Jieqing Zheng

atw Vol. 65 (2020) | Issue 1 ı January

ENVIRONMENT AND SAFETY 16

| Fig. 8.

Peak cladding temperature response.

has already reached 1213.7 K which is

considered to have the zirconiumwater

reaction becoming drastic. As a

result, it is a case which is deemed to

be not acceptable for implementing

the URGs. From numerous sensitivity

studies on the injection flow rates, the

critical point to maintain the peak

temperature below 1088 K is around

127 gpm.

Timing for injecting water is also

an important factor in the compound

accident. Considering the complexity

of the accident, sometimes injection of

the water may not be made available

right at 1 hour. It’s very important to

investigate the latest timing for

injection. Taking the injection rate of

200 gpm as an example, the peak

cladding temperature will be below

1088 K. As shown in Figure 9, if water

is made available at 75 minutes after

the initiation of the event, the peak

cladding temperature will reach

1080 K which is nearly the same as the

critical temperature of 1088 K. The

total amount of time that includes

personnel getting ready and pipes for

water injection gotten hooked up

should be less than 75 minutes for the

case with the water injection rate of

200 gpm. The cumulative hydrogen

generation in core for this case is only

0.85 kg which is considered to be

minimal. While the amount of total

hydrogen generated will go up to

118 kg if the injection is further

delayed (from 75 minutes after initiation

of the event to 90 minutes). Thus,

by further delaying the timing for

injection of water for 15 minutes,

the amount of hydrogen generated

increases by more than 100 times. The

response of the amount of hydrogen

generated is shown in Figure 10.

5 Conclusions

This paper illustrates the idea of

Ultimate Response Guidelines for NPP

| Fig. 9.

Peak cladding temperature response.

together with the simulations of the

compound accident cases with and

without URGs using the MAAP5

evaluation methodology. Based on the

results obtained from these simulations,

the following conclusions can

be summarized for the Chinshan NPP.

1) For the compound accident, if

there is no water available after

RCIC pump trips off, the accident

will result in melting of the core

and breaching of the reactor vessel.

2) The timing to enter the URGs must

conform to one initial condition.

The sooner the operator injects

water into the core, the less danger

the plant becomes. According to

the calculated results obtained

from the MAAP5 code, the flow

rate of 127 gpm is the minimum

necessary to maintain the PCT

below 1088.6 K.

3) Implementation of URGs can effectively

mitigate the consequences of

a postulated compound accident.

In this study, with the water

injection rate of 127 gpm being

injected to the reactor at 1 hr from

the initiation of the event, the

Chinshan NPP has demonstrated to

enter a safe state where its reactor

core overheating is prevented.

Acknowledgements

The authors’ heartfelt gratitude to the

supports of Taiwan Power Company,

Institute of Nuclear Energy Research,

Chung Yuan Christian University, and

Science&Technology Department of

Fujian Province, P.R.C (JK2016023)

for this project.

References

[1] Kim YH, Kim MK, Kim WJ. Effect of the Fukushima nuclear

disaster on global public acceptance of nuclear energy:

Energy Policy 2013; 61:822–828.

[2] Funabashi H. Why the Fukushima Nuclear Disaster is a

Man-made Calamity: International Journal of Japanese

Sociology 2012; 21:65-75.

| Fig. 10.

Hydrogen generation response.

[3] Ozdemir OE, George TL, Marshall MD. Fukushima Daiichi Unit

1 power plant containment analysis using GOTHIC: Annals of

Nuclear Energy 2015; 85:621–632.

[4] MAAP5-Modular Accident Analysis Program User’s Manual,

Fauke & Associates Inc., 2008.

[5] Huh CW, Suh ND, Park GC. Optimum RCS depressurization

strategy for effective severe accident management of station

blackout accident: Nucl Eng Des 2009; 239:2521–2529.

[6] Liu KH, Hwang SL. Human performance evaluation: the

procedures of ultimate response guideline for nuclear power

plants: Nucl Eng Des 2012; 253: 259–268.

[7] Vo TH, Song JH, Kim TW, Kim DH. An analysis on the severe

accident progression with operator recovery actions: Nucl Eng

Des 2014; 280: 429–439.

[8] Wang TC, Wang JR, Lin HT, et al. The ultimate response

guideline simulation and analysis using TRACE, MAAP5, and

FRAPTRAN for the Chinshan Nuclear Power Plant: Annals of

Nuclear Energy 2017; 103:402–411.

Authors

Jieqing Zheng

Cleaning Combustion and Energy

Utilization Research Center

of Fujian Province

Jimei University

9 Shigu Road, Xiamen, China

Environment and Safety

Analysis of Ultimate Response Guidelines for Chinshan Nuclear Power Plant in Taiwan to Cope with Postulated Compound Accident ı Jieqing Zheng

atw Vol. 65 (2020) | Issue 1 ı January

Stilllegung und Rückbau des

Rossendorfer Forschungsreaktors RFR

Teil 2: Ausgewählte Aspekte der Durchführung von Stilllegung und Rückbau

Reinhard Knappik, Klaus Geyer, Sven Jansen und Cornelia Graetz

Im Teil 1 der Veröffentlichung (atw 11/12 2019) erfolgten nach einer Einführung die Objekt beschreibung, die Darstellung

der Ausgangssituation (radiologisch, konventionell), die Erläuterung der Genehmigungsverfahren, das realisierte

Planungskonzept sowie die Aufzählung von Meilensteinen der Stilllegung und des Rückbaus. Im zweiten Teil wird von

ausgewählten Aspekten der Stilllegung- und Rückbaudurchführung berichtet.

7 Technische/

technologische Aspekte

Basierend auf den erteilten Genehmigungen

erfolgten die Stilllegung

und der Rückbau in den in Tabelle 1

dargestellten Zeiträumen, aus denen

in diesem Kapitel einige wichtige

Aspekte aus technisch/technologischer

Sicht dargestellt werden.

7.1 Betriebsführung der

abgeschalteten Anlage

gemäß Aufsichtlicher

Anordnungen

Die Betriebsführung der abgeschalteten

Anlage erfolgte bis zum Erhalt

der Ersten Stilllegungsgenehmigung

am 30. Januar 1998 u. a. auf der

Grundlage der Aufsichtlichen Anordnung

VKTA 40-42 des SMU vom

30. Dezember 1991 [13]. Im Zeitraum

bis Oktober 1998 wurden technische,

sicherheitstechnische und strahlenschutztechnische

Maßnahmen zur

Anpassung an den bundesdeutschen

Standard durchgeführt, Genehmigungsanträge

erarbeitet, das Betriebsund

Prüfhandbuch sowie der Sicherheitsbericht

RFR erstellt, gutachterlich

und behördlich geprüft, Stilllegungsarbeiten

und insbesondere die

Umlagerung der Brennelemente technisch

und genehmigungsmäßig vorbereitet.

Von besonderer Bedeutung

war dabei die Umstellung der Entsorgungskonzeption

von CASTOR-THTR

auf CASTOR® MTR 2-Behälter. Mit

einer eigens entwickelten Mobilen

Umladestation sollte die Möglichkeit

geschaffen werden, dass auch andere

deutsche Forschungsreaktoren Brennelemente

auf diese Art entsorgen können.

Die Entwicklung, der Bau und

die Kalterprobung der Mobilen Umladestation

erfolgten in sehr guter

Zusammenarbeit mit verschiedenen

Partnern von 1993 bis 1999. Der

atomrechtliche Genehmigungsantrag

für die Überführung der Brennelemente

in die Transport- und Lagerbehälter

CASTOR® MTR 2 wurde im

Zeitraum Tätigkeiten Genehmigung

06/1991 - 02/1998 sichere Betriebsführung der abgeschalteten Anlage gemäß

Aufsichtlicher Anordnungen

02/1998 - 07/2019 Innehaben, Betriebsführung gemäß

Erster RFR-Stilllegungsgenehmigung

11/1998 - 04/1999 Abbau 2. Kühlkreislauf 2.

04/2001 - 04/2005 Rückbau von Systemen und Komponenten des RFR 3.

04/2005 - 12/2007 Vorbereitende Maßnahmen zum Rückbau Reaktorbaukörper 4.

07/2007 - 04/2014 Abbau des Reaktorbaukörpers und Gebäude-Entkernung 4.

08/2013 - 11/2018 Abbruch der Objekte und Herstellen „Grüne Wiese“ 4.

| Tab. 1.

Zeiträume von Stilllegung und Rückbau sowie deren Genehmigungsbezug.

Januar 1994 gestellt und im Dezember

1998 die Genehmigung [5] erteilt.

Zu den Vorarbeiten gehörten die

Umlagerung von bestrahlten Brennelementen

(ca. 400 Stück) vom

Brennelemente-Lagerbecken AB 1

in das Brennelemente-Lagerbecken

AB 2, welches im Jahr 1997 durch Einsatz

diversitärer Messtechnik, erhöhtem

Leckageschutz und Verbesserung

der Wasseraufbereitung ertüchtigt

wurde. Ab diesem Zeitpunkt befanden

sich 889 Brennelemente im Lagenbecken

AB 2 und 62 im Reaktorkern.

Im September 1997 wurde das Wasser

des Lagerbeckens AB 1 abgegeben

und danach das Becken für die Nutzung

als Reststofflagerbecken saniert.

Im Jahr 1994 erfolgte die Über gabe

eines mobilen Betriebssystems sowie

von unbestrahlten Brenn elementen

zur Nachnutzung an einem ungarischen

Forschungsreaktor und 1995 die

Ausfuhr eines weiteren Betriebssystems

an ein Forschungszentrum in der

Tschechischen Republik.

7.2 Betriebsführung gemäß

1. AtG-Genehmigung,

Teilabbau 2. Kühlkreislauf,

CASTOR-Beladung

Nach Erhalt der Ersten Stilllegungs-

Genehmigung [3] erfolgten ab

Februar 1998 die Betriebsführung

sowie der Ablauf der Stilllegungsarbeiten

auf der Grundlage dieser

| Abb. 7.

Entladung der Brennelemente aus dem Reaktorkern.

Genehmigung. Im April 1998 wurden

die im Reaktorkern befindlichen

Brennelemente in das Lagerbecken

AB 2 umgelagert (Abbildung 7) und

mit dem Ausbau der kernbrennstoffhaltigen

Neutronendetektoren (Spaltkammer)

die Kernmaterialfreiheit des

Reaktorbehälters hergestellt.

Nach Erhalt der Genehmigung [5]

konnte mit der Überführung der 951

bestrahlten Brennelemente mit einer

Gesamtaktivität von 8,91E+15 Bq

und einer U-235-Masse von rund

54,6 kg in CASTOR® MTR 2-Behälter

begonnen werden. Dies erfolgte in

zwei Etappen, da Teile der Mobilen

Umladestation von April bis Juli 1999

für den Einsatz am Reaktor der Medizinischen

Hochschule Hannover genutzt

wurden. Die Abbildung 8 zeigt

Aufsichtliche

Anordnungen

1. ; [5 , 6, 7]

DECOMMISSIONING AND WASTE MANAGEMENT

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atw Vol. 65 (2020) | Issue 1 ı January

DECOMMISSIONING AND WASTE MANAGEMENT 18

| Abb. 8.

Beladung eines CASTOR-Behälters.

das Aufsetzen des mit bestrahlten

Brennelementen gefüllten Umlagebehälters

Cäsar auf den CASTOR-

Behälter. Mit dem Transport der restlichen

CASTORen in die Transportbereitstellungshalle

des VKTA bis

November 2000 wurde das Vorhaben

gemäß einer § 9 AtG-Genehmigung

[6] abgeschlossen. Die kollektive

Strahlenexposition bei den CASTOR-

Beladearbeiten betrug 1,8 mSv

und die maximale Individualdosis

0,24 mSv und war damit wesentlich

niedriger als im Geneh migungsantrag

ausgewiesen. Die CASTORen verblieben

bis zum Abtransport in das

Zwischenlager Ahaus im Mai/Juni

2005 in der Transport bereit stellungshalle

am Forschungsstandort Rossendorf.

Die weiteren kernbrennstoffhaltigen

Abfälle wurden nach Erteilung

einer Genehmigung nach § 9

AtG [7] im Februar 2001 verpackt und

der radioaktive Abfall ins Zwischenlager

Rossendorf überführt, so dass

nach Herstellung der Kernmaterialfreiheit

der RFR- Anlage am 26. Februar

2001 die Aufhebung der Sicherungsbereiche

der Anlage erfolgen konnte.

Nach Erhalt der Zweiten Still legungs-

Genehmigung am 30. Oktober

1998 [4] wurden alle Systeme und

| Abb. 9.

Ziehen des Reaktorbehälters.

Komponenten des 2. Kühlkreislaufes

(KKL) außer Betrieb genommen und

von den Medienversorgungen getrennt.

Vor Abgabe von rund 130 m 3

deionisiertem Wasser aus dem 2. KKL

an die entsprechende Fachabteilung

des VKTA wurde auf der Basis von

Probennahmen und Analysen die Freigabe

zur Ableitung erteilt. Im Verlauf

des Jahres 1999 erfolgten der Rückbau

der Komponenten sowie im

August 1999 die Entlassung des

Systems und der Gebäude (Pumpenund

Armaturenhaus, Trockenkühltürme)

des 2. KKL aus dem Geltungsbereich

des AtG. Nach Erhalt der baurechtlichen

Genehmigungen wurden

die Ent kernung und der Abbruch dieser

Gebäude sowie die Rekultivierung

des Geländes vorgenommen.

7.3 Rückbau von Systemen

und Komponenten des RFR

Von April 2001 an erfolgte in einem

Zeitraum von vier Jahren neben der

Entsorgung der Betriebsmedien, die

Außerbetriebnahme und der Rückbau

aller nicht mehr benötigten Systeme

und Komponenten des RFR in 14 Teilschritten.

Ein Teilschritt, der Abbau

des Deaerators, konnte aus technologischen

Gründen erst im Rahmen des

Vierten Stilllegungsschrittes erfolgen.

Die Leistungen wurden bis auf einen

Teilschritt durch das ehemalige Reaktorpersonal

bewältigt. Wichtige Teilschritte

waren die Demontage der

Einbauten und der Ausbau des Reaktorbehälters

(Abbildung 9), das Freiräumen,

die Dekontamination und

Demontage der in Rossendorf als

Heiße Kammern (HK, Abbildung 10)

bezeichneten Heißen Zellen, der

Abbau der Thermischen Säule (Abbildung

11), die Außerbetriebnahme

und der teilweise Rückbau der Lagerbecken

AB 1 und AB 2 sowie der Rückbau

des 1. Kühlkreislaufes. So wurden

beispielsweise beim Abbau des 1. KKL

im Pumpenraum 95 % der kontaminierten

Edelstahlteile (rund 40 Mg)

nach Zerlegung zur Behandlung in

die VKTA-Einrichtung transportiert,

während rund 30 Mg anderer Stahl,

das Abschirmmaterial aus Beton und

Blei und der restliche Edelstahl

uneingeschränkt freigegeben werden

konnten.

Der Reaktorbehälter aus Aluminium

wurde nach dem Ziehen gesäubert,

beschichtet, verpackt und zur

Konditionierung zu einem Dienstleister

überführt. Letztendlich erhielt

der VKTA 14 Abfallgebinde mit einer

Nettomasse von rund 2,9 Mg zur Einstellung

in das Zwischenlager Rossendorf

zurück.

Die vier Heißen Kammern waren

mit Stahlblech ausgekleidet, mit einer

Schwerbeton-Abschirmung ummantelt

und untereinander mit einem

Transportkanal verbunden. Sie verfügten

über einen Fußbodenablauf.

Die Bedienung jeder Heißen Kammer

erfolgte über einen Manipulatorraum

mit zwei Manipulatoren. Genutzt

wurden die Heißen Kammern, um die

mittels Rohrpostanlage vom Reaktor

kommenden bestrahlten Isotopenkassetten

für die Weiterverarbeitung in

der Isotopenproduktion vorzu bereiten.

Die Demontage, Dekontamination und

Verpackung des Inventars der Heißen

| Abb. 10.

Blick in eine Heiße Kammer während der Demontage der Inneneinrichtung.

| Abb. 11.

Thermische Säule vor der Demontage.

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atw Vol. 65 (2020) | Issue 1 ı January

Kammern, der Transportwagenanlage

sowie zugehöriger Antriebssysteme in

den Manipulator räumen mussten aufgrund

des Kon taminationszustandes

z. T. mittels fremdbelüfteter Vollschutzanzüge

erfolgen.

Bei der Demontage der Thermischen

Säule, die aus sechs aluminiumummantelten

Graphit-Segmenten

und einem Fahrwagen bestand, traten

Dosisleistungen bis 3 mSv/h auf, zu

deren Minimierung am Arbeitsort das

aktivierte Vorderteil des Fahrwagens

mit Bleiblechen ab geschirmt wurde.

Nach einer Abklinglagerung konnten

die Segmente 4 bis 6 nach Freimessung

2019 freigegeben werden. Der erreichte

Zustand des Reaktorbaukörpers

wird in Abbildung 12 gezeigt.

7.4 Vorbereitende

Maßnahmen zum Rückbau

der Baustrukturen

Die vorbereitenden Maßnahmen, vor

allem zum Rückbau der Baustrukturen

sowie der bisher noch benötigten

Systeme, begannen wegen fehlender

finanzieller Mittel zeitlich um

18 Monate versetzt Ende September

2006, mit vorbereitenden Arbeiten,

Umbauten an den Zugängen zum

Kontrollbereich und Freischaltarbeiten,

wie beispielsweise

p die Anpassung der Personen- und

Materialwege,

p die Bereitstellung von Ausrüstungen

und Transportmitteln,

p der Abbau der äußeren Anbauten

am Reaktorbaukörper,

p die Anpassung der Medienver- und

-entsorgung,

p die Errichtung einer Einhausung

um den Reaktorbaukörper,

p umfangreiche lüftungstechnische

Änderungen,

p der schrittweise Aufbau einer

Baustromversorgung und

p statische Maßnahmen zur Erhöhung

von Tragfähigkeiten.

Außerdem wurden die äußeren Reaktoranbauten,

wie z. B. Kabeltrassen,

demontiert, um Baufreiheit für die zu

errichtende Einhausung zu schaffen.

Die Einhausung wurde als Stahlbau

errichtet, mit schwer entflammbaren,

leicht dekontaminierbaren Folienwänden

verkleidet, zur Be- und Entlüftung

Filteranlagen im Umluftbetrieb

eingesetzt sowie mit einem

5 t-Brückenkran ausgestattet.

7.5 Abbau des Reaktorbaukörpers

und

Gebäude-Entkernung

Diese Etappe begann, wiederum in

Teilschritten gegliedert, im September

2007 im Kellergeschoss mit

Abbrucharbeiten im Bereich der

Abluftkanäle. Hier wurde der beim

Abbruch des Reaktorbaukörpers zum

Einsatz kommende funkferngesteuerte

Abbruchbagger Top Tec 1850E

getestet. In einem Durchführungszeitraum

von ca. sieben Jahren

wurden die in der angegebenen Folge

aufgeführten Arbeiten erledigt:

p Abbau der Auskleidungen und

Einbauten im Lagerbecken und in

den Heißen Kammern

(08/2007 bis 02/2011)

p Abbau des RFR-Baukörpers

(04/2008 bis 08/2009)

p Abbau der in Beton verlegten

Abluftkanäle und Rohrleitungen

(01/2008 bis 12/2010)

p Abbruch der Heißen Kammern

p Abbau der Einhausung in der

Reaktorhalle

p Entkernung und Dekontamination

des Kontrollbereiches

(09/2011 bis 09/2012)

p Entkernung des Labortraktes und

der Warte (03/2014 bis 04/2014)

p Demontage der lüftungstechnischen

Anlagen im Ventilationsund

Filtergebäude (02/2013 bis

08/2013) einschließlich des

Ausbaus der im Erdreich verlegten

Abluftkanäle von der Reaktorhalle

(06/2008 bis 08/2009)

p Entkernung des Ventilationsund

Filtergebäudes

(02/2013 bis 08/2013)

p Abbau und Entsorgung

des Fortluftschornsteins

(06/2013 bis 02/2014)

Der Abbruch des RFR-Baukörpers

wurde mit dem erwähnten Bagger,

der auf eine Plattform aufgesetzt

wurde, durchgeführt. Die Abbildung

13 zeigt eine schematische Darstellung

des Abbruchs. Dabei erfolgte

die Befestigung der Plattform auf dem

obersten innenliegenden Gusseisenring

des Reaktorkörpers. Entsprechend

des Abbruchfortschrittes und

der Reichweite des Baggers wurden

dann die Plattform mit dem Bagger

abgenommen, einige Gusseisenringe

entfernt und nach erneutem Aufsetzen

auf den nächsten Gusseisenring

die Abbrucharbeiten weitergeführt.

Reaktormittig war anstatt der

Gusseisenringe zur Fixierung der

Strahlrohre eine ca. 2,50 m hohe

zylindrische Stahlzarge mit oberem

und unterem Flansch eingebaut.

Das Abbruchmaterial wurde innerhalb

der Reaktor-Einhausung mittels

Brecher zerkleinert und anschließend

in 500-l-Boxen verpackt zum Freimesszentrum

transportiert. Der

Abbruch der im Kellergeschoss befindlichen

vier Heißen Kammern war

| Abb. 12.

Reaktorbaukörper nach Abbau der Komponenten.

| Abb. 13.

Schematische Darstellung des Abbaus des RFR-Baukörpers.

zunächst nicht geplant. Nach dem

Ausbau der Einbauten und Auskleidungen

der Heißen Kammern musste

aber festgestellt werden, dass eine

Freigabe des Betonkörpers der Heißen

Kammern mit den eingebauten Fenstern

und Plugs an der stehenden

Struktur nicht erfolgen konnte. Somit

wurden die Heißen Kammern nach

Änderung der Materialwege im

Gebäude ebenfalls abgebrochen.

Dabei gestalteten sich die Arbeiten an

der Transportwagen-Anlage mit den

dazugehörigen Antriebssystemen in

den Manipulatoren-Räumen aufgrund

des Kontaminationszustandes

ebenfalls als schwierig und musste

teilweise in fremdbelüfteten Vollschutzanzügen

durchgeführt werden.

Des Weiteren erschwerte das unerwartete

Auffinden einer komplizierten

Stahlrahmenkonstruktion der

Heißen Kammern die Arbeiten. Der

Abbruch des Reaktorbaukörpers einschließlich

der Heißen Kammern und

Nebenanlagen endete im Juni 2011.

Ein interessanter Meilenstein war

der Abbau des Fortluftschornsteines

(Abbildung 14), der im Juli 2013

mittels zweier Mobilkräne (90 Mg und

250 Mg) vom Dach des Ventilationsgebäudes

gehoben und im Hof des

RFR zur Dekontamination und Zerlegung

abgelegt wurde. Nach Verschluss

aller Öffnungen und Anbringen einer

partiellen Einhausung erfolgten

Dekontamination, Freimessung nebst

uneingeschränkter Freigabe und

zeitnah die Reststoffentsorgung nach

entsprechender Zerlegung vor Ort.

DECOMMISSIONING AND WASTE MANAGEMENT 19

Decommissioning and Waste Management

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

atw Vol. 65 (2020) | Issue 1 ı January

DECOMMISSIONING AND WASTE MANAGEMENT 20

| Abb. 14.

Abbau des Fortluftschornsteines

vom Ventilations- und Filtergebäude.

Um die zwischen dem Reaktorgebäude

sowie dem Ventilations- und

Filtergebäude teilweise unter dem als

Pavillon bezeichnetem Nebengebäude

verlaufenden Abluftkanäle und Rohrleitungen

ausbauen zu können, wurde

der Pavillon im Zeitraum von September

bis Oktober 2013 abgebrochen.

Mit den weiteren Demontage-, Entkernungs-

und Grobdekontaminationsarbeiten

wurden die Voraussetzungen

für die schrittweise Freimessung

des Labortraktes mit Reaktorwarte,

Reaktorhalle und Ventilations-

und Filtergebäude geschaffen.

Zudem erfolgten Vorbereitungsarbeiten

wie beispielsweise Umbauten

am Hallenkran zum Freimessen der

Hallendecke.

7.6 Abbruch der Objekte und

Herstellen „Grüne Wiese“

Mit Erteilung der 2. Änderungsgenehmigung

zum Vierten Stilllegungsschritt

wurde der Überwachungsbereich

erweitert. Die erteilten SMUL-

Zustimmungen zum Teilabbruch des

Ventilations- und Filtergebäudes

( Oktober 2014) sowie zum Abriss von

Gebäudestrukturen des Labortraktes

inklusive Warte und Reaktorhalle

( Juli 2015) bildeten eine der Voraussetzung

für die letzte Rückbau-Etappe.

Dabei waren aufgrund vorausgehender

konventioneller Untersuchungen

insbesondere teerhaltige

Beschichtungen und künstliche Mineralfasern

als Schadstoffe zu beachten,

entsprechend zu separieren und zu

entsorgen. Außerdem ist zu erwähnen,

dass bedingt durch lokale Kontaminationen

mit dem Alphastrahler

Am-241 die Abluftanlagen im Ventilations-

und Filtergebäude unter

erhöhten Strahlenschutzmaßnahmen,

vor allem zum Inkorporationsschutz

durchgeführt werden mussten. Zu

diesem Arbeitsabschnitt gehörten

weiterhin:

p der Ausbau von Rohrleitungen,

Kabeln, Kanälen und Schächten

im Hofbereich

(Abschluss Dezember 2016)

p das Verfüllen der zuvor freigegebenen

Baugruben und Gräben

nach jeweiliger Zustimmung

durch das SMUL

p die Abdeckung der Hofflächen

p die Profilierung des Geländes bis

zur Herstellung der „Grünen

Wiese“ (November 2018)

Die Abbildung 15 zeigt den Abbruch

des Ventilations- und Filtergebäudes,

der im Zeitraum von Dezember 2014

bis Ende April 2015 durchgeführt

wurde und die entstandene Baugrube

nach deren Teilverfüllung, damit die

im hinteren Geländebereich bis zum

Zaun des RFR-Geländes befindlichen

Rohrleitungen in einem weiteren Teilschritt

ausgebaut werden konnten.

Der oberirdische Abbruch des freigegebenen

Labortraktes inklusive

Reaktorhalle erfolgte von August bis

November 2015 unter Einsatz eines

50 t Baggers mit sogenannter Longfront

(Abbildung 16). Anschließend

wurden die unterirdischen Baustrukturen

mittels Abbruchbagger bis

August 2016 abgebrochen (Abbildung

17), zerkleinert und bis auf die

Massen aus den „Freigabeinseln“ [15]

(siehe Abschnitt 8) entsorgt. An den

zwei tiefsten Seiten wurde die Baugrube

vor Abbruch der Kellerstrukturen

zur Minimierung von Erdstoffbewegungen

und zur Sicherung von

Fahrwegen mit einer Spundwand gesichert,

die im Zuge der Verfüllung

wieder gezogen wurde. Der sukzessive

Ausbau von Rohrleitungen und

Schächten, beispielhaft gezeigt in

Abbildung 18 vor dem Abbruch des

Labortraktes, war eine logistisch und

entsorgungstechnisch anspruchsvolle

Aufgabe, da erhebliche Erdstoffmassen

bewegt bzw. zwischengelagert

werden mussten, ohne die

weiteren Rückbau- bzw. Messaufgaben

zu behindern. Nach dem

Ausbau erfolgten die Vorbereitungen

| Abb. 15.

Abbruch des Ventilations- und Filtergebäudes (links) und teilverfüllte Baugrube (rechts).

| Abb. 16.

Abbruch des Labortraktes und der Reaktorhalle.

| Abb. 18.

Abbau von Abluftleitungen

an der Reaktorhalle.

| Abb. 17.

Abriss der Kellerstrukturen.

| Abb. 19.

Teilansicht der Baugrube RFR mit Rastermarkierung

und zusätzlich ausgehobenen

Rasterflächen.

Decommissioning and Waste Management

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

atw Vol. 65 (2020) | Issue 1 ı January

zur Freimessung der Bodenflächen,

die Messungen u. a. mittels In-situ-

Gamma spektrometrie sowie die Entnahme

von Erdreich-Proben, die hinsichtlich

Radionuklide und konventioneller

Schadstoffe untersucht und

bewertet wurden. Bei der Bewertung

der Baugrube RFR stellte sich heraus,

dass drei Rasterflächen (Abbildung

19) noch eine PAK-Kontamination

aufwiesen, so dass ein weiterer Erdaushub

und erneute Analysen erforderlich

wurden, um letztendlich die

Schadstofffreiheit des Baufeldes RFR

festzustellen.

Bis Ende 2016 waren alle Objekte

im RFR-Gelände ausgebaut und bis

auf eine Restmenge alle Stoffe freigemessen,

freigegeben und entsorgt.

Die Verfüllung der Baugruben mit

Kontrolle der bodenmechanischen

Kennwerte und des eingebauten Erdreiches

sowie die Profilierung des

Geländes erfolgten bis Ende 2018. Im

Juni 2018 wurde der Antrag auf Entlassung

des RFR aus dem Geltungsbereich

des AtG beim SMUL gestellt

und mit weiteren Unterlagen bis Juli

2019 ergänzt.

7.7 Arbeits- und

Brandschutzaspekte

Wie bei allen Rückbauprojekten besaß

der Arbeits- und Brandschutz eine

hohe Priorität. Alle Arbeiten wurden

stets unter Beachtung der gesetzlichen

Bestimmungen sowie entsprechend

den Vorschriften der Unfallversicherungsträger

durchgeführt.

In Vorbereitung eines jeden Rückbauvorhabens

wurde ein Rückbauerlaubnisverfahren

durchgeführt. Dabei

betrachtete man mittels einer Checkliste

„Voraussetzungsprüfung für

Rückbauphase“ u. a. vorliegende

sicherheitstechnische Anlagen des

RFR wie Brandbekämpfungseinrichtungen.

Vor Beginn eines jeden Loses

eines Vorhabens wurden im Rahmen

einer Arbeitsplatz-Gefährdungsbeurteilung

die möglichen Gefährdungen,

wie gesundheitsgefährdende

Stäube etc., ermittelt sowie die technischen

und organisatorischen Maßnahmen

zur Abwendung der Gefährdungen

und zur Gewährleistung der

Sicherheit festgelegt. Bei den Gefährdungsbeurteilungen

waren die verantwortlichen

Mitarbeiter der jeweiligen

Dienstleister involviert. Jeder

Dienstleister erstellte dazu noch die

für seine Tätigkeiten erforderlichen

Unterlagen wie z. B. Abbruchanweisungen

oder spezielle Gefährdungsbeurteilungen.

Mit einem Arbeitserlaubnis-Schein

überprüften Gesamtverantwortlicher,

Einsatzleiter,

Durchführender und Strahlenschutzbeauftragter

die festgelegten notwendigen

Arbeitssicherheits-, Brandschutz-

und Strahlenschutzmaßnahmen

mit dem Ziel der Freigabe des

jeweiligen Arbeitsvorhabens. Diese

Arbeitserlaubnis beinhaltete auch die

Überprüfung der Notwendigkeit eines

Erlaubnisscheins für Erdarbeiten, eines

Erlaubnisscheins für Arbeiten in Behältern

und engen Räumen, einer Arbeitserlaubnis

für feuergefährliche Arbeiten

sowie einer Freischalter laubnis. Im

Vorfeld wurden den Dienst leistern notwendige

Unterlagen zum Verhalten auf

dem Betriebsge lände übergeben. Vor

Arbeitsauf nahme fand eine Unterweisung

der Dienstleister statt. Hier erhielten

sie Informationen zum Strahlenschutz,

zur Gewährleistung der

Ersten Hilfe (Standorte der Verbandskästen),

zu Notrufen, Fluchtwegen

und Sammelplätzen. In der gesamten

Rückbauzeit gab es keine bedeutsamen

Arbeits unfälle und keine Brände.

8 Strahlenschutzaspekte

Die Aufgaben des Strahlenschutzes

im Rückbau gliederten sich in drei

Schwerpunkte:

p Emissionsüberwachung

p Dosimetrische Überwachung

p Betrieblicher Strahlenschutz/Anlagenüberwachung

Die Emissionsüberwachung und die

dosimetrische Überwachung erfolgten

durch den zentralen Strahlenschutz

des VKTA. Die betriebliche Strahlenschutzüberwachung

/ Anlagen über wachung

erfolgte durch Mitarbeiter des

Rückbau-Strahlenschutzpersonals, untergeordnet

durch Mitarbeiter des

zentralen Strahlenschutzes.

8.1 Emissions- und

Immissionsüberwachung

Die Überwachung erfolgte im Rückbauzeitraum

in Anlehnung an die

Richtlinie zur Emissions- und Immissionsüberwachung

kerntechnischer

Anlagen.

Abwasser

Im Zeitraum 1998 bis inkl. 2015 erfolgte

die Überwachung kontaminationsverdächtiger

Abwässer durch

Probennahmen an insgesamt vier

Sammelstellen. Später wurde die

Sammlung in entsprechenden Kleinbehältern

realisiert, wobei es sich dabei

hauptsächlich um Waschwässer handelte.

Von insgesamt ca. 700 m³ angefallenen

Wässern konnte für ca. 430 m³

ein Entscheid zur Ableitung erteilt werden.

Diese erfolgte überwiegend über

die Laborabwasserreinigungsanlage

des Forschungsstandortes, die seit

2001 in Betrieb ist. Eine Ausnahme

bildete hier die Ableitung von ausschließlich

H-3-haltigem Deionat aus

dem Abklingbehälter 2 des RFR im

März 2005. Diese Ableitung wurde

nach Zustimmung der zuständigen Behörde

dosiert direkt in den Vorfluter

Kalter Bach vorgenommen und stellte

mit 3,8E+10 Bq H-3 zugleich die bedeutendste

Ableitung aus dem RFR

dar. Die Ausschöpfung der festgelegten

Obergrenze betrug maximal 5 %.

Die restlichen Abwässer mit erhöhtem

Radionuklidinventar wurden vor

Ableitung dekontaminiert. Insgesamt

betrug das Abwasseraufkommen aus

dem RFR und den dazugehörigen

Behältern 5 % der Gesamtwassermenge

des Forschungsstandortes im

oben genannten Zeitraum.

Fortluft

Die gefilterte Abluft wurde über den

41,8 m hohen Fortluftschornstein des

Ventilations- und Filterhauses abgeleitet.

Die Überwachung der Fortluft

erfolgte kontinuierlich durch Messungen

am isokinetisch aus dem Fortluftschornstein

entnommenen Teilvolumenstrom.

Dieser wurde mittels

Aerosol- und H3/C14-Sammlers hinsichtlich

Alpha-, Beta,- und Gamma-

Aerosolen sowie hinsichtlich gasförmigem

H3 und C-14 gemessen und

bilanziert. Dabei betrug im Rückbauzeitraum

die maximale Ausschöpfung

der Obergrenzen bei Gasen 31 % (H-3

im Jahr 1994) und bei Schwebstoffen

36 % (Alphastrahler im Jahr 2004),

11 % (Betastrahler im Jahr 2004)

sowie 0,2 % (Gammastrahler im Jahr

2007).

Im Zuge des fortschreitenden Rückbaus

wurde mit der Inbetriebnahme

mobiler Abluftanlagen sowie dem

Rückbau des Fortluftschornsteins ab

August 2012 die Fortluftüberwachung

der Restanlage auf grund der vernachlässigbaren

Emissionen eingestellt.

Für eine Übergangszeit erfolgte allerdings

bis 2013 noch eine Überwachung

auf Aerosole mit Ableitung

der Fortluft über einen eigens am

Ventilations- und Filter gebäude errichteten

10-m-Kamin. Die im Fortluft-

Emissionsplan für den Emittenten

„RFR“ festgelegten Obergrenzen

wurden im gesamten Zeitraum für alle

Nuklidgruppen weit unterschritten.

Immissionsüberwachung

Die Immissionsüberwachung des

Forschungsstandorts Rossendorf umfasste

neben der Überwachung der

weiteren Umgebung (Beprobung

Sediment, Grasproben, Lebensmittelproben)

auch die Ortsdosimetrie mit

DECOMMISSIONING AND WASTE MANAGEMENT 21

Decommissioning and Waste Management

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

atw Vol. 65 (2020) | Issue 1 ı January

DECOMMISSIONING AND WASTE MANAGEMENT 22

Festkörperdosimetern an Grenzen

von Strahlenschutzbereichen, so auch

am RFR. Der maximale Wert durch

Direktstrahlung an Grenzen des

Strahlenschutzbereichs RFR zum

Betriebsgelände wurde dabei mit

0,7 mSv in Jahr 1994 gemessen.

8.2 Dosimetrische

Überwachung

Die Überwachung der Inkorporationen

erfolgte durch Messungen im

Ganzkörperzähler hinsichtlich gammastrahlender

Nuklide sowie Ausscheidungsanalysen

(Stuhl, bspw.

hinsichtlich U/Pu-Nuklide oder Am-

241, Urin, bspw. Sr-90+). Hierbei

wurden das Eigenpersonal sowie das

Fremdpersonal überwacht. Die maximale

effektive Folgedosis in einem

Kalenderjahr betrug 2,36 mSv.

Die amtliche Überwachung der

äußeren Exposition erfolgte durch

Ganzkörperdosimeter. Dabei kamen

Albedo- und Filmdosimeter zum Einsatz.

Es wurde nur das Eigenpersonal

überwacht. Das Fremdpersonal wurde

durch die jeweiligen Fremdfirmen

überwacht. Die maximale Körperdosis

in einem Kalenderjahr betrug

0,6 mSv. Die Kollektivdosis als Summe

der effektiven Folgedosis aus Inkorporation

und der äußeren Exposition

von 1998 bis 2018 lag bei etwa

18 mSv.

Die zusätzliche betriebliche Überwachung

erfolgte mit Hilfe von

elektronischen Dosimetern sowie

Festkörperdosimetern zur Teilkörperüberwachung.

8.3 Anlagenüberwachung

Die Anlagenüberwachung umfasste die

Überwachung der Oberflächenkontamination,

der Ortsdosisleistung und

der Raumluftaktivität. Dies geschah in

Form routinemäßig überwachter Messpunkte

im festen Turnus sowie projektoder

anlassbezogen. Dies bedeutete,

dass jeder Rückbauschritt durch ein abgestimmtes

Überwachungsprogramm

abgedeckt wurde.

Aus den Ergebnissen der Anlagenüberwachung

wurden Rückschlüsse

auf die einzusetzende Schutzkleidung

gezogen, um jederzeit einen ausreichenden

Schutz der Mitarbeiter zu

gewährleisten, ohne überzogene

Schutzmaßnahmen festzulegen. Da

bei vielen Rückbauschritten das Vorhandensein

von Alphastrahlern in

inkorporationsrelevanten Größenordnungen

nicht ausgeschlossen werden

konnte, musste eine entsprechend

feingliedrige Anlagenüberwachung

durchgeführt werden. Die Datenhaltung

und -auswertung erfolgte mit

Hilfe einer im VKTA entwickelten

Datenbank.

Überwachung

der Ortsdosisleistung

Zur Überwachung der Ortsdosisleistung

wurden an festgelegten

Messpunkten in den Strahlenschutzbereichen

und an deren Grenzen

jährlich ca. 600 bis 1000 Messungen

vor genommen, wobei nur an einzelnen

Stellen Werte >15 µSv/h

gemessen wurden. (Das in Arbeitsbereichen

ermittelte Maximum lag

bei 3 mSv/h.).

Überwachung

der Oberflächenkontamination

Zur Überwachung der Oberflächenkontamination

wurden an festgelegten

Messpunkten jährlich neben

ca. 400 bis 800 Routinemessungen

projektbegleitende Messungen an

bestimmten Rückbauorten mittels

Wischprobennahmen und α- bzw. β/γgesamtzählenden

Direktmessungen

vorgenommen. Untergeordnet kamen

Kratzprobennahmen zum Einsatz.

Während die Ergebnisse der Routineuntersuchungen

im Bereich der Nachweisgrenze

bzw. innerhalb der Grenzwerte

nach StrlSchV lagen, ergaben

sich an Arbeitsorten z. T. Messwerte,

die Größenordnungen darüber lagen.

Dazu gehörten auch α-Oberflächenkontaminationen,

die auf eine Am-

Freisetzung zurückzuführen waren.

8.4 Meldepflichtige Ereignisse

Im gesamten Stilllegungs- und Rückbauzeitraum

gab es zehn meldepflichtige

Ereignisse, z. B. einen Defekt des

Hallenhubtores oder die Beschädigung

der Laufkatze des Reaktorhallenkrans.

Sie besaßen alle keine

radiologische Relevanz.

9 Freimessung und

Freigaben

Um den hauptsächlichen Stoffanteil

des RFR-Rückbaus einer Verwertung

oder Entsorgung zuzuführen, bestand

die Zielstellung, möglichst zeitnah

umfassend Freimessungen vorzugsweise

aller Reststoffe durchzuführen,

auf deren Basis die Freigabe gemäß

§ 29 StrlSchV2001 erteilt werden

kann. Im VKTA wurden dazu in

Abstimmung mit dem SMUL zwei

Verfahrenswege genutzt:

1) Zum einen konnten auf der Grundlage

behördlich zur Freimessung

zugelassener Messverfahren Reststoffe

freigemessen und nach

Bewertung der Ergebnisse durch

den Freigabe-Strahlenschutzbeauftragten

bzw. ab 12/2005 dem

Freigabebeauftragten freigegeben

werden, sofern er die Übereinstimmung

mit den Festlegungen des

auf dem Freigabebescheid

fußenden innerbetrieblichen Regelwerkes

festgestellt hatte.

2) Zum anderen wurden für Gebäude/

Gebäudestrukturen zur Weiterverwendung

bzw. zum Abriss, Baugruben,

Gräben zur Verfüllung und

die Freigabe von Bodenflächen

sogenannte Frei messprogramme

vom VKTA erstellt, die nach behördlicher

Zustimmung umgesetzt

wurden. Auf der Basis erstellter

Unterlagen (Ergebnisbericht, Freigabeanträge)

erfolgten die betriebliche

Freigabe und die Dokumentenübermittlung

an das SMUL, das

in der Regel einen Sachverständigen

einschaltete. Mit den Ergebnissen

des Sachverständigen

erteilte das SMUL nach deren Prüfung

die Freigabe.

Die Vorbereitung und Durch führung

der Freimessung erfolgte schon mit

Beginn des jeweiligen Rückbauschrittes.

Sie durchläuft dabei prinzipiell

die Abfolge:

p Historische Erkundung

p Radiologische Erkundung

p bei Bedarf auch Dekontamination,

diese mit Ergebniskontrolle

p Vormessung

p Erstellung des Freimessprogrammes

nebst behördlichem Bestätigungsverfahren/Begutachtung

p ggf. Feindekontamination

p ggf. Messungen zur Überprüfung

des Dekontaminationserfolges

p Entscheidungsmessung

p Übergabe der Ergebnisse in Analogie

zum Freimessprogramm,

Auswertung an die Behörde

p Kontrollmessung/Begutachtung

p Prüfung durch die Behörde

p Freigabe

Zur radiologischen Bewertung kamen

vorrangig folgende Verfahren zum

Einsatz:

p Gesamtzählende Direktmessungen

der Ortsdosisleistung und der

Oberflächenkontamination

p Probenauswertung mit α-, β-, γ-gesamtzählenden

Messplätzen sowie

Flüssigszintillationszählern

p Laborgammaspektrometrische Untersuchungen

von Proben mittels

HP-Ge-Detektoren in abgeschirmten

Messplätzen sowie α- und β-

nuklidspezifische Analysen

p In-situ-gammaspektrometrische

Messungen

p Messungen von Gebinden im

Freimesszentrum des VKTA

Oftmals fanden vor Einsatz der

Messverfahren Probenaufbereitungen

Decommissioning and Waste Management

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

atw Vol. 65 (2020) | Issue 1 ı January

statt, teils in Form aufwändiger radiochemischer

Trennverfahren. Nähere

Einzelheiten zum konzeptionellen

Herangehen und zur messtechnischen

Umsetzung der Freimessung RFR

können aus [14] entnommen werden.

Freimessungen und Freigaben wurden

zeitlich gestrafft und erreichten in

der Rückbauphase einen großen Umfang;

erfolgten aber auch schon in der

Stilllegungsphase bis 2001. Beispielhaft

sei erwähnt, dass man für die Entscheidungsmessungen

und Freigaben

der Baustrukturen Labortrakt, Reaktorwarte

und Reaktorhalle rund 12

Monate (Januar 2013 bis Januar

2014) benötigte.

Nicht sofort freigabefähige Komponenten,

an denen aber nach Behandlung

eine vollständige oder teilweise

Freigabe erwartet werden konnte,

wurden der VKTA-Einrichtung zur Behandlung

schwachradioaktiver Abfälle

zugeführt. Diese Bearbeitung sowie

die Behandlung von Abklingabfällen

aus dem Zwischenlager Rossendorf

werden noch einige Zeit benötigen.

Während in der ursprünglichen

Planung für den Labortrakt mit Reaktorhalle

sowie für das Ventilationsund

Filtergebäude die komplette Freigabe

an der stehenden Struktur mit

anschließendem konventionellen Abriss

vorgesehen war, konnte dies nicht

realisiert werden. Gründe dafür lagen

vor allem in statischen Erfordernissen

– die Entfernung oder Dekontamination

kontaminierter Komponenten

bzw. die messtechnische Bewertung

einzelner Objekte war nicht möglich,

ohne die Statik des Gebäudes zu

gefährden. Diese Stellen wurden als

sogenannte „Freigabeinseln“ aus der

Gesamtheit der freizugebenden

Strukturen herausgenommen und vor

Ort nebst entsprechendem Sicherheitspuffer

gekennzeichnet. Durch

diese Freigabeinseln ergaben sich Umplanungen

bei der Durchführung des

Abbaus und Ergänzungen in Form von

weiteren Erläuterungsberichten. Nach

Abriss des restlichen (weit überwiegenden)

Teils des Gebäudes

wurden die Freigabeinseln ausgebaut

und entsprechend Verfahrensweg 1)

bewertet [15].

Für Stilllegung und Rückbau der

RFR-Anlagen wurden seit 2008 ca.

1400 Freigabevorgänge in Form von

Einzel- bzw. Gruppenanträgen erfolgreich

abgeschlossen, zwischen 1998

und 2007 ca. 450. Insgesamt erfolgten

Freigaben für eine Stoffmenge von

rund 20.000 Mg. Zudem wurde bedingt

durch Baugrubenböschungen

und Gräben sowie einige nicht vermeidbare

Mehrfachbewertungen eine

Gesamtfläche von rund 12.000 m 2

überwiegend nach StrlSchV 2001 Anlage

III Tabelle 1 Spalte 6 entsprechend

einer mit dem SMUL

abgestimmten Verfahrensweise freigegeben.

Folgende Freigabepfade

(Bewertung nach StrlSchV 2001 Anlage

III Tabelle 1) wurden im Zuge der

Freimessung und Freigabe beschritten:

p Spalten 4 und 5/9 (bzw. 9a, 9c) für

Einzelteile

p Spalten 5/9 (bzw. 9b, 9d) für

brenn bare Reststoffe

p Spalte 5/9 für entnommenes Erdreich

bzw. entnommenen Bauschutt

p Spalte 6 für tiefliegende Teile des

Erdreichs nach Zustimmung der

Behörde (mit anschließender Abdeckung

von 80 cm – im Randbereich

von 30 cm und entsprechendem

Überdeckungsnachweis

an die Behörde)

p Spalte 7 für oberflächennahe Teile

des Erdreichs, Bodenoberflächen

p Spalte 8 für nicht ohne weiteres

rückbaubare tiefliegende Strukturen

(Ausnahmefall)

p Spalte 10 für Gebäude und Gebäudeteile

zum Abriss

10 Freigegebene Reststoffe

und radioaktive Abfälle

Der erreichbare Rückbaufortschritt

wird maßgeblich von der zügigen Entfernung

der freigebbaren Reststoffe

und radioaktiven Abfälle von der Baustelle

bestimmt. Um kostenaufwendige

Zwischenlagerschritte zu minimieren,

wurde auf die zügige Entsorgung

freigabefähiger Reststoffe

großen Wert gelegt. Bereits durch die

Voruntersuchungen, die sowohl radiologisch

als auch schadstoffbezogen

durchgeführt wurden, konnten Art

und Umfang relevanter Schadstoffklassen

erkannt und Vorarbeiten für

die spätere Deponierung und Verbrennung

(z. B. Klärung des Deklarationsumfanges,

Vertragsbindung mit

Entsorgungsanlagen, Entwicklung von

Mess- und Bewertungsverfahren) gelegt

werden. Weiterhin wurden rückbaubegleitend

Überprüfungen veranlasst

bzw. detaillierte Untersuchungen

an den Stellen vorgenommen, die in

den Voruntersuchungen nicht oder

nur partiell erfasst werden konnten.

Durch die sorgfältige und kleinteilige

Trennung des radioaktiven Abfalls von

den Reststoffen wurde der Stoffanteil

zur Endlagerung minimiert.

Die Betrachtung der potentiellen

wie realen Schadstoffsituation aus

chemotoxischer Sicht nahm in

der Bearbeitung einen nicht zu

unterschätzenden Anteil ein, da sowohl

hinsichtlich der Entsorgung und

Verwertung der Reststoffe als auch

hinsichtlich der Endlagerung die stoffliche

(Schadstoff-) Charakterisierung

einen hohen Stellenwert besitzt.

Als dominierender konventioneller

Schad stoff für die Einstufung der

Abfälle nach Freigabe ergaben sich die

Polyzyklische Aromatischen Kohlenwasserstoffe

(PAK; Vorkommen z. B.

in teerhaltigen Dachpappen, Sperrschichten

außerhalb und innerhalb

von Gebäuden) sowie damit korrelierend

der Phenolindex. Für die

Bewertung in Hinblick auf die Entsorgung

wurden die Zuordnungskriterien

gemäß Deponieverordnung

[16] und LAGA [17] zu Grunde gelegt.

Bedingt durch die außenliegenden

Sperrschichten im Bereich der unteririschen

Baustrukturen war in gebäudenahen

Bereichen das Erdreich teilweise

PAK-kontaminiert. In [11] wird

detaillierter auf die Reststoffentsorgung

nach Freigabe gemäß

§ 29 StrlSchV 2001 beim RFR-Rückbau

eingegangen.

Insgesamt wurde beim RFR-

Rückbau eine Stoffmenge von rund

41.000 Mg erhalten, die sich zunächst

in radioaktiver Abfall (330 Mg), behandlungsfähiges

Material (200 Mg)

und freigabefähige Reststoffe

(40.470 Mg) aufteilte. Der Anteil an

radioaktivem RFR-Abfall konnte bedingt

durch die Behandlung einer Teilmenge

sowie der Abklinglagerung auf

rund 150 Mg reduziert werden, wobei

sich davon bereits rund 2 Mg im Endlager

für radioaktive Abfälle Morsleben

befinden. Die noch zu behandelnde

Reststoffmasse beträgt rund

40 Mg. Daraus ergeben sich derzeit

folgende prozentuale Anteile: 99,5 %

freigegebene Reststoffe davon rund

49 % zweckgerichtet freigegeben,

0,4 % radioaktiver Abfall und 0,1 %

noch zu behandelndes Material. In der

Abbildung 20 wird zu diesen Angaben

ein Materialbezug hergestellt.

| Abb. 20.

Prozentuale Materialangaben zu den Rückbaumassen.

DECOMMISSIONING AND WASTE MANAGEMENT 23

Decommissioning and Waste Management

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

atw Vol. 65 (2020) | Issue 1 ı January

DECOMMISSIONING AND WASTE MANAGEMENT 24

Die Konditionierung der radioaktiven

Abfälle zur Endlagerung, z. B.

durch Hochdruckverpressung, verlief

z. T. parallel zum Rückbau. Vor allem

aufgrund der Randbedingungen im

Rahmen der Produktkontrolle sind

bis zur Abgabe aller vorhandenen

radioaktiven Abfälle in das Endlager

Konrad noch eine ganze Reihe von

Aufgaben zu bewältigen.

11 Kosten

Die genaue Kostenaufschlüsselung im

VKTA und speziell die Betrachtungsweise

der Rückbau- und Sanierungsprojekte

wechselte seit 1992 mehrfach.

Somit ist eine genaue Ab schätzung

alleine für den RFR nicht möglich. Für

den Rückbau und die Sanierung der

Flächen des RFR, der Isotopenproduktion

und der Ent sorgungsanlagen wurden

ca. 59 Millionen € verwendet. In

dieser Summe sind aber die Kosten der

Neubauten (z. B. Entsorgungsanlage

für Kern material, Einrichtung zur

Behandlung von schwachradioaktiven

Abfällen, Zwischenlager) nicht enthalten.

Die Entsorgung des gesamten

Kernbrennstoffinventares einschließlich

der bestrahlten Brennelemente

fehlt ebenfalls bei dieser Summe.

12 Fazit

Das aus rückbautechnischer und technologischer

Sicht geplante Rückbaukonzept

zur Beseitigung des Rossendorfer

Forschungsreaktors mit all

seinen peripheren Einrichtungen

wurde erfolgreich umgesetzt, auch

wenn ursprünglich erst die Ent lassung

des Rossendorfer Forschungsreaktors

aus dem Geltungsbereich des AtG und

danach der konventionelle Abriss der

Gebäude vorgesehen waren. Aufgrund

sogenannter „Freigabeinseln“

[15], die erst im Zuge des Abbaus der

Gebäudestrukturen zur Freigabe

geführt werden konnten, musste der

Gebäudeabriss unter Strahlenschutzbedingungen

erfolgen. Zeitlich und

finanziell ergaben sich dadurch allerdings

keine größeren Probleme. Die

neu entstandenen Flächen wurden

nach der Sanierung rekultiviert

und sollen dem Helmholtz-Zentrum

Dresden- Rossendorf übergeben werden,

um eine zukünftige Nutzung zu

ermöglichen.

Beginnend mit dem Sächsischen

Kabinettsbeschluss zur Stilllegung

und zum Rückbau des RFR im Jahre

1993 bis zum Abschluss des Vorhabens

im Jahr 2019 gab es bezüglich

des Strahlen-, Arbeits-, Brandschutzes

keine nennenswerten Ereignisse.

Es hat sich bewährt, mit dem Rückbau

auch gleichzeitig die Entsorgung

gezielt voranzutreiben, so dass der

VKTA derzeit nur noch rund 40 Mg

(entspricht rund 0,1 %) der Rückbaumasse

in Bearbeitung hat, um Freigaben

zu erreichen. Neben den 951im

Brennelement-Zwischen lager Ahaus

lagernden Brennelementen befinden

sich aus dem RFR-Rückbau weiterhin

noch ca. 148 Mg radioaktive Abfälle

im Zwischen lager Rossendorf. Bewährt

haben sich der Einsatz des RFR-

Betriebspersonals insbesondere in

der Zeit von 1992 bis 2007 sowie die

interne Zusammenarbeit hinsichtlich

des Strahlenschutzes, der radiologischen

Messungen, der Analytik im

akkreditierten Labor des VKTA und

des Führens der atomrechtlichen

Genehmigungsverfahren.

Insgesamt wurden ca. 59 Millionen

€ für den Rückbau und die

Sanierung der Flächen des RFR, der

Isotopenproduktion und der Entsorgungsanlagen

verwendet.

Der VKTA dankt den eingesetzten

Fremdfirmen für ihre Unterstützung,

den Mitarbeitern der Genehmigungsbehörde

für die konstruktive Zusammenarbeit

und dem Freistaat Sachsen

für die Finanzierung sowie das erbrachte

Vertrauen gegenüber dem

VKTA hinsichtlich der Erfüllung von

Stilllegung und Rückbau des Rossendorfer

Forschungsreaktors bis hin zu

„Grünen Wiese“.

Referenzen

[1] Hieronymus, W. et al.: Beiträge zur Geschichte des

Rossendorfer Forschungsreaktors RFR,

ISBN: 978-3-941405-04-2, überarbeitete Fassung 2009

[2] Grahnert, T., Jansen, S., Boeßert, W., Kniest, S. Stilllegung und

Rückbau der Rossendorfer Isotopenproduktion, atw, Vol. 61

(2016) und Vol. 62 (2016)

[3] Erste Genehmigung 45-4653.18 VKTA 01 gemäß § 7 Absatz

3 AtG zur Stilllegung des Rossendorfer Forschungsreaktors

RFR (1. Stilllegungsgenehmigung RFR – Innehaben,

Betriebsführung, Überführung der Brennelemente aus der

Spaltzone in des Brennelementlagerbecken AB 2) des SMU,

erteilt am 30.01.1998, mit 1. Änderung vom 06. 11. 2000

[4] Zweite Genehmigung 45-4653.18 VKTA 02 gemäß § 7

Absatz 3 AtG zur Stilllegung des Rossendorfer Forschungsreaktors

RFR (2. Stilllegungsgenehmigung RFR – Rückbau

des 2. Kühlkreislaufes) des SMU, erteilt am 30.10.1998, mit

1. Änderung vom 11. 02.1999

[5] Genehmigung 74-4653.13 gemäß § 9 AtG zur sonstigen

Verwendung von Kernbrennstoffen außerhalb

genehmigungs pflichtiger Anlagen und zum Umgang mit

sonstigen radioaktiven Stoffen zur Überführung der

bestrahlten Brennelemente des Rossendorfer Forschungsreaktors

(RFR) in Transport- und Lagerbehälter vom Typ

CASTOR® MTR 2 des SMUL, erteilt am 17.12.1998

[6] Genehmigung 4653.15 gemäß § 9 AtG zur sonstigen Verwendung

von Kernbrennstoffen außerhalb genehmigungspflichtiger

Anlagen und zum Umgang mit sonstigen radioaktiven

Stoffen (Überführung von Kernbrennstoffen aus den

Verwahrorten der Reaktorhalle in Abfallgebinde) des SMUL,

erteilt am 06.02.2001

[7] Genehmigung 74-4653.93 gemäß § 9 AtG zur sonstigen

Verwendung von Kernbrennstoffen außerhalb

genehmigungs pflichtiger Anlagen und zum Umgang mit

sonstigen radioaktiven Stoffen (Transportbereitstellung der

CASTOREN in der Transportbereitstellungshalle (TBH) sowie

im Freigelände um die TBH) des SMUL, erteilt am

21.12.1998

[8] Dritte Genehmigung 4653.18 VKTA 03 gemäß § 7 Absatz 3

AtG zur Stilllegung und zum Abbau des Rossendorfer

Forschungsreaktors RFR (3. Stilllegungsgenehmigung RFR –

Abbau des Reaktorsystems und seiner Komponenten) des

SMUL, erteilt am 03.04.2001

[9] Vierte Genehmigung 4653.18 VKTA 04 gemäß § 7 Absatz 3

AtG zum Abbau der Restanlage des Rossendorfer

Forschungsreaktors RFR SMUL (4. Stilllegungsgenehmigung

RFR – Totalabbruch der RFR-Restanlage) des SMUL, erteilt

am 01.02.2005 mit 1. Änderungsbescheid vom 09. 11.2010

und mit 2. Änderung vom 09.01.2014

[10] Langer, R., Steinbach, F. Michael: Entsorgung freigegebener

Reststoffe nach Rückbau des RFR, KONTEC 2017

[11] Steinbach, P., Johne, B., Steinhardt, M., Knappik, R.

Kerntechnischer Rückbau unter Beachtung des Boden- und

Grundwasserschutzes, KONTEC 2019

[12] Große, H., Jähnichen, S., Michael, F., Steinbach, P.: Analytik

von Polyzyklischen aromatischen Kohlenwasserstoffen bei

Rückbau kerntechnischer Anlagen, KONTEC 2019

[13] Aufsichtliche Anordnung VKTA 40-42 des SMU

vom 30.12.1991

[14] Bothe, M., Knappik, R., Kahn, A., Emmrich, U.

Konzeptionelles Herangehen und messtechnische

Umsetzung zur Freimessung der Gebäude des Rossendorfer

Forschungsreaktor, KONTEC 2013

[15] Jansen, S., Michael, F., Johne, B.

Beseitigung der „Freigabeinseln“ beim Rückbau

des Rossendorfer Forschungsreaktors, KONTEC 2017

[16] Verordnung über Deponien und Langzeitlager (Deponieverordnung

– DepV) vom 27.04.2009 (BGBl. I S. 900),

die zuletzt durch Artikel 2 der Verordnung vom 27.09. 2017

(BGBl. I S. 3465) geändert worden ist

[17] LAGA Anforderungen an die stoffliche Verwertung von

mineralischen Abfällen: Teil I »Allgemeiner Teil« der LAGA M

20 (Stand 6.11.2003), Teil II: Technische Regeln für die

Verwertung, 1.2 Bodenmaterial (TR Boden), Stand:

05.11.2004, Teil III »Probenahme und Analytik« (Stand

5.11.2004), Vorläufige Hinweise zum Einsatz von Baustoffrecyclingmaterial

(länderspezifische Regelung Sachsen)

vom 11.01.2006, verlängert am 24.10.2014

Authors

Reinhard Knappik

Klaus Geyer

Sven Jansen

Cornelia Graetz

VKTA Rossendorf

Bautzner Landstraße 400

01328 Dresden

Decommissioning and Waste Management

Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

atw Vol. 65 (2020) | Issue 1 ı January

Thermal-Hydraulic Analysis for

Total Loss of Feedwater Event in PWR

using SPACE Code

MinJeong Kim, Minhee Kim, Junkyu Song, Bongsik Chu, Jae-Seung Suh and Hyunjin Lee

After the Fukushima nuclear accident, in Japan, comprehensive safety assessments for nuclear power plants are

performed by regulators around the world. As a part of the safety enhancement effort, additional failure of the safety

components are considered and to maintain safety margin, review and improve emergency procedures. In Korea, a new

regulatory requirement is introduced, which requires all nuclear power plants to submit Accident Management Plan

(AMP) that covers design basis accidents, multiple failure accidents and severe accidents.

Total Loss of Feedwater (TLOFW)

event is one of the main multiple

failure accident which assumes failure

of both main feedwater and auxiliary

feedwater system. Since there is no

feedwater supply to steam generators,

heat cannot be removed through

steam generators. In TLOFW event,

primary side feed and bleed operation

is manually performed to remove

heat. Feed and bleed operation continues

until reactor coolant system

(RCS) is cooled and depressurized to

the point where shutdown cooling

system can be used to remove heat

from RCS.

In this paper, thermal-hydraulic

analysis of TLOFW event for OPR1000

plants is performed to evaluate the

validity of RCS cool down strategy

using Safety and Performance Analysis

Code for Nuclear Power Plants

(SPACE). Hanul units 3&4 are selected

as the reference plants and analysis

results show that the RCS cool down

strategy through the feed and bleed

has sufficient core cooling capacity

which prevents core damage.

1 Introduction

After Fukushima nuclear accident, in

Japan, nuclear regulators of around

the world launched a comprehensive

check for their nuclear power plants.

They concluded that nuclear power

plants should consider accidents of

Design Extension Condition (DEC).

Considering beyond design basis

accidents has become very important

for developing cool down strategies

for the Reactor Coolant System (RCS)

and recovery actions. It is also necessary

to consider additional failure of

the safety components in terms of

sufficient safety margin with applying

of proper emergency operating procedures

[1].

The revision of the nuclear safety

act in June, 2015 required all nuclear

power plants in Korea to submit

Accident Management Plan (AMP)

that covers design basis accidents,

multiple failure accidents and severe

accidents.

The Total Loss of Feedwater

(TLOFW) event is one of the multiple

failure events. TLOFW assumes that

the feed water supply is completely

stopped by failure of both main feedwater

and auxiliary feedwater(AFW)

due to pump failure, pipe break or

other. Since there are two motor

driven AFW pumps and two turbine

driven AFW pumps, probability of

TLOFW occurring is very low. There

are several safety limits related to the

TLOFW event. Core damage from fuel

heat-up should not occur during the

event. The maximum allowable fuel

cladding temperature is 1,204 °C

(2,200 °F ). To maintain fuel cladding

temperature below the limit, it is

necessary for the primary system to

have sufficient core cooling capability.

When heat removal through the

secondary system is not available, the

decay heat of the core should be

removed by rapid depressurization of

the primary system and operation of

the Emergency Core Cooling System

(ECCS). According to recent studies

on the TLOFW event [2-6], the feed

and bleed operations has been one of

the useful strategies for removing the

decay heat. The OPR1000 plants have

been designed to manually operate

the feed and bleed strategy during the

TLOFW event by using the Safety

Depressurization System (SDS). The

SDS valves provides rapid depressurization,

which is connected to the

top of the pressurizer with two flow

paths. The feed and bleed operations

can be started after the pressure of

the primary system reaches safety

injection actuation point.

In this paper, we present thermalhydraulic

analysis for the TLOFW

event assuming the loss of the secondary

cooling function by the failure of

main feedwater and auxiliary feedwater

system. We use the Safety and

Performance Analysis Code for

Nuclear Power Plants (SPACE) code

which is an advanced thermal hydraulic

analysis code with two-fluid and

three-field governing equations [7].

The comparative study covers three

cases according to operations of the

SDS and safety injection system

during the TLOFW event. We also

examine the effectiveness of the RCS

cool down strategy through the feed

and bleed operations in accordance

with the emergency operating procedure

(EOP). The reference plants

for this study are the Hanul units 3&4.

2 Analysis information

2.1 SPACE code

The Korea Hydro & Nuclear Power Co.

through collaborative works with

other Korean nuclear industries and

research institutes has been developing

the thermal-hydraulic analysis

code for the safety analysis of the

Pressurized Water Reactors (PWRs),

which was named the SPACE. The

SPACE is the best-estimate two-fluid

and three-field analysis code for

analyzing the safety and performance

of the PWRs. The code has been

developed to improve the prediction

accuracy of the thermal hydrodynamic

behavior of the nuclear reactor

system in transient conditions. The

semi-implicit scheme has been used

for the time integration method. The

SPACE code consists of the package of

the input and output package, the

reactor kinetics model, the hydrodynamic

model, and the heat structure

model.

The hydrodynamic model package

is composed of hydraulic solver, constitutive

models, special process

models, and component models. The

hydraulic solver is based on two-fluid

and three-field governing equations

RESEARCH AND INNOVATION

Research and Innovation

Thermal-Hydraulic Analysis for Total Loss of Feedwater Event in PWR using SPACE Code ı MinJeong Kim, Minhee Kim, Junkyu Song, Bongsik Chu, Jae-Seung Suh and Hyunjin Lee

atw Vol. 65 (2020) | Issue 1 ı January

RESEARCH AND INNOVATION 26

| Fig. 1.

Nodalization diagram of OPR1000.

which are gas, continuous liquid, and

droplet fields. The SPACE code has

an advantage in solving a dispersed

liquid field as well as vapor and

continuous liquid fields. The constitutive

models involve the flow regime

map to simulate the mass, momentum,

and energy distributions such as

surface area, surface heat transfer,

surface-wall friction, droplet separation

and adhesion, and wall-fluid heat

transfer. The heat structure model

can solve transient heat conduction

problems in the rectangular or

cylindrical geometry with various

boundary conditions for convection

and radiation problems and user

defined variables such as the temperature,

heat flux, and heat transfer

coefficient. Nuclear fission heat from

nuclear fuel rods can be calculated by

using point kinetics approximation

and treated as a heat source in the

heat conduction equation. Reactivity

feedbacks are considered in terms of

the moderator density, moderator

temperature, fuel temperature, boron

concentration, reactor scram, and

power defect. The SPACE 3.0 version

is used in this investigation.

2.2 Steady state

Figure 1 shows the system nodalization

used in the SPACE code for analyzing

the TLOFW event. Before entering

transient conditions using the restart

function of the SPACE code, the

steady-state calculation is performed

to confirm the initial conditions.

The initial conditions of steady-state

were represented in Table 1. The

Parameter Design value Steady state value

Core power (MWt) 2815 2815

Cold-leg Temperature (°C) 295.8 298

Hot-leg Temperature (°C) 327.2 329

RCS flow 15,308.7 15,336

PZR pressure (MPa) 15.5132 15.5

PZR level (%) 52.6 52.6

Steam Generator Pressure (MPa) 7.38 7.389

Steam flow rate (kg/s) 802.9 801.5

Feedwater flow rate (kg/s) 802.6 798

| Tab. 1.

Initial conditions for the TLOFW event.

calculation for steady-state condition

is performed for 3,000 seconds.

The Pressurizer Safety Valve (PSV)

as modeled as a component C511

with opening pressure setpoint of

17.23 MPa. When the TLOFW event

occur, the SDS valves modeled as

components C551 and C552 is used

for bleed operations to remove the

decay heat of the core. The multiple

failure accident which includes the

TLOFW accident can be analyzed

using best estimate methods. In order

to obtain realistic steam pressure

response after turbine trip, Steam

Bypass Control System (SBCS) was

used. The SBCS was modeled into

eight separate valves, C811 ~ C818.

2.3 Sequence of events

Different simulation scenarios are

considered for the TLOFW event

based on design requirements of the

SDS described in the Final Safety

Analysis Report (FSAR) of Hanul

Units 3&4 [8]. OPR1000 plants carry

out the feed and bleed operations

with two Safety Injection Pumps

(SIPs) and two SDS trains, respectively.

The simulation scenarios with

consideration for design requirements

of the SDS described in Ref. [8] are as

follows.

p When one SIP is available, each

SDS train shall be designed to have

sufficient capacity to prevent the

reactor core exposure, assuming

that the SDS path is opened simultaneously

with the opening of the

PSV in the TLOFW event. (Case 1)

p When two SIPs are available, two

SDS trains shall be designed to

have sufficient capacity to prevent

the reactor core exposure, assuming

that the opening of the SDS

paths is delayed by 30 minutes

from with the opening of the PSV

in the TLOFW event. (Case 2)

Furthermore, Case 3 is considered to

create the additional situation in the

TLOFW event. In this case, one SIP

and one SDS are available, but assuming

that the opening of the SDS path is

delayed by 30 minutes from with the

opening of the PSV. In all cases, it is

assumed that the Safety Injection

Tanks (SITs) and the SBCS are fully

available during the event. All cases

are summarized in Table 2.

The TLOFW event starts with loss

of main feedwater. Water level of the

Steam Generators (SGs) continues to

decrease and reach the set point of the

reactor trip. Turbine trip occurs with

reactor trip. The decay heat from the

core is removed through SGs, with

steam flow controlled by SBCS. SG

level continues to drop and auxiliary

feedwater actuation setpoint is

reached. However, auxiliary feedwater

is assumed to fail. It is assumed

that the Reactor Coolant Pumps

(RCPs) is stopped at 10 minutes after

the reactor and turbine trip. As the

water inventory of SGs continues to

decrease and SGs become dry, the

Research and Innovation

Thermal-Hydraulic Analysis for Total Loss of Feedwater Event in PWR using SPACE Code ı MinJeong Kim, Minhee Kim, Junkyu Song, Bongsik Chu, Jae-Seung Suh and Hyunjin Lee

atw Vol. 65 (2020) | Issue 1 ı January

Case

Case 1

Case 2

Case 3

Remark

| Tab. 2.

Summary of all cases.

| Tab. 3.

Sequence of the events.

PSV open + SDS 1 train open (0sec) + 1 out of 2 SIP available

PSV open + SDS 2 train open (1800sec) + 2 out of 2 SIP available

PSV open + SDS 1 train open (1800sec) + 1 out of 2 SIP available

No. Event Remark

1 accident occur

2 Reactor trip (SG low level) 42.9 % WR

3 TBN trip

4 Auxiliary feedwater injection fail 23.4 % WR

5 RCP trip manual

6 SG dry out

7 PSV open PPZR > 17.23 MPa

8 SDS valve open (manual) Case 1: PSV open + 0 s

Case 2: PSV open + 1800 s

Case 3: PSV open + 1800 s

9 Low Pressurizer Pressure (LPP) signal

10 HPSI injection

11 SIT injection

heat removal through SGs are no

longer possible. Without heat removal

through SGs, the temperature and

pressure of the primary system increases

and reaches the set point of

the PSV opening. The feed and bleed

operations start with manual opening

of the SDS valve in accordance with

the EOP. The pressure of the primary

system decrease to the set point of the

High Pressure Safety Injection (HPSI).

The sequence of events is summarized

in the Table 3.

3 Simulation result

Figure 2 shows the pressures of the

primary and secondary systems in

Case 1. After initiating the TLOFW

event, the reactor and turbine trip

occur by the low SG level signal. After

the turbine trip, the pressure of the

secondary system increases and

reaches the SBCS actuation signal. As

the heat removal capacity of the SGs is

diminished by the loss of the feed

water supply, the pressure of the

primary system increases to the set

pressure of the PSV. The feed and

bleed operations by manual opening

of the SDS can reduce the pressure of

the primary system. As shown in

Figure 3, the PSV closes as soon as the

SDS valve opens.

The pressure of the primary system

decreases as the primary inventory is

discharged through the SDS. The SIP

is triggered by the low pressurizer

pressure signal. The RCS is sufficiently

depressurized and its pressure

reaches the injection pressure of SITs.

Figure 4 shows the mass flow rates of

the SIP and SIT.

Figure 5 shows the pressures of

the primary and secondary system in

Case 2. The pressure of primary

system increases until the PSV valves

open. The PSV repeats open and close

as shown in Figure 6. And then the

SDS valves are manually opened

30 minutes after the PSV is first

opened. Both pressures of the primary

and secondary systems decrease with

the bleed operation with the SDS.

The SIP start to operate by the low

pressurizer pressure signal. When

the primary pressure decreases to

the actuating pressure of the SITs

RESEARCH AND INNOVATION 27

| Fig. 2.

Pressures of the primary and secondary system (Case 1).

| Fig. 3.

Mass flow rates of PSV and SDS (Case 1).

| Fig. 4.

Mass flow rates of SIP and SIT (Case 1).

| Fig. 5.

Pressures of the primary and secondary system (Case 2).

| Fig. 6.

Mass flow rates of PSV and SDS (Case 2).

| Fig. 7.

Mass flow rates of SIP and SIT (Case 2).

Research and Innovation

Thermal-Hydraulic Analysis for Total Loss of Feedwater Event in PWR using SPACE Code ı MinJeong Kim, Minhee Kim, Junkyu Song, Bongsik Chu, Jae-Seung Suh and Hyunjin Lee

atw Vol. 65 (2020) | Issue 1 ı January

RESEARCH AND INNOVATION 28

| Fig. 8.

Pressures of the primary and secondary system (Case 3).

| Fig. 11.

Peak cladding temperatures.

pro viding the borated water to the

RCS. Figure 7 shows the mass flow

rates of SIP and SIT in Case 2.

Case 3 is the additional scenario in

this case study as described in previous

section. Figure 8 shows the pressures

of the primary and secondary systems

in Case 3. The pressure of the primary

system increases to the set point of the

PSV and then oscillates. Thermal

hydraulic behavior of Case 3 is similar

to that of Case 2 until SDS valve opens.

In Figure 9, one train of SDS opens

30 minutes after the PSV is first

opened. The pressures of the primary

and secondary systems decrease as

the primary inventory is discharged

through the SDS. The SIP is operated

by the low pressurizer pressure signal.

The primary pressure continues to

decrease and reaches the injection

pressure of SITs. Figure 10 shows the

mass flow rates of the SIP and SIT.

Peak cladding temperatures of all

cases are shown in Figure 11. Case 1

is 325 °C , while both Case 2 and Case

3 are 354 °C . Injection of the SIP cools

the core down immediately. The fuel

cladding temperatures of all cases

don’t exceed the maximum allowable

fuel cladding temperature, 1,204 °C

(2,200 °F ). Which means the core

cooling capabilities are sufficient

in all cases.

| Fig. 9.

Mass flow rates of PSV and SDS (Case 3).

4 Conclusions

In this study, we present thermalhydraulic

analysis for the TLOFW

event in OPR1000 using the SPACE

3.0 code. Three different cases

according to operations of the SDS

and safety injection system were

analyzed to examine the effectiveness

of the RCS cool down strategy through

the feed and bleed operations to

mitigate the TLOFW event. The feed

and bleed operations start with

manually opening of the SDS valve

after the PSV opening in accordance

with the EOP.

The simulation scenarios of Case 1

and Case 2 were based upon design

requirements of the SDS described in

the FSAR of Hanul units 3&4. Case 3

was the additional scenario in this

comparative study. The peak cladding

temperatures of all cases did not

exceed 1,204 °C (2,200 °F) which is

the maximum allowable fuel cladding

temperature. The RCS cool down

strategy through the feed and bleed

operations can guarantee the core

cooling capabilities during the TLOFW

event. The earlier feed and bleed

operation was more effective strategy

for removing the decay heat. We also

confirmed that the SPACE code is very

useful code for analyzing the multiple

failure accidents in the PWR.

Acknowledgments

This work was supported by the

Korea Institute of Energy Technology

Evaluation and Planning (KETEP)

grant funded by the Korea government

(MOTIE) (No. 20161510101840,

Development of Design Extension

Conditions Analysis and Management

Technology for Prevention of Severe

Accident).

References

1. Korea Hydro and Nuclear Power Co. Ltd., “Development of

Design Extension Conditions Analysis and Management

Technology for Prevention of Severe Accident Report”,

September, 2017.

| Fig. 10.

Mass flow rates of SIP and SIT (Case 3).

2. Kwon, Y.M. et al., “Comparative simulation of feed and bleed

operation during the total loss of feedwater event by

RELAP5:MOD3 and CEFLASH-4AS:REM computer codes,

Nuclear Technology, Vol. 112, pp. 181– 193, 1995.

3. Kwon, Y.M., Song, J.H., “Feasibility of long term feed and bleed

operation for total loss of feedwater event”, Journal of Korean

Nuclear Society, Vol. 28 (3), pp. 257–264, 1996.

4. Park, R.J. et al., “Detailed evaluation of coolant injection into

the reactor vessel with RCS depressurization for high pressure

sequences”, Nuclear Engineering and Design, Vol. 239,

pp. 2484–2490, 2009.

5. Pochard, R. et al., “Analysis of a feed and bleed procedure

sensitivity study performed with the SIPACT simulator on a

French 900 MWe NPP”, Nuclear Engineering, Des. 215,

pp. 1–14, 2002.

6. Reventós, F. et al., “Analysis of the feed & bleed procedure

for the Ascó NPP first approach study for operation support”,

Nuclear Engineering, Des. 237, pp. 2006–2013, 2007.

7. S. J. Ha et al., “Development of the SPACE Code for Nuclear

Power Plants,” Nuclear Engineering & Technology, Vol. 43,

No. 1, pp. 45, 2011.

8. Final Safety Analysis Report Hanul 3,4, KHNP.

Authors

MinJeong Kim

Minhee Kim

Junkyu Song

Bongsik Chu

Central Research Institute

Korea Hydro and Nuclear Power

Co., Ltd.

Deajeon, 34101

Rep. of Korea

Jae-Seung Suh

Hyunjin Lee

System Engineering and

Technology Co., Ltd.

Daejeon, 34324

Rep. of Korea

Research and Innovation

Thermal-Hydraulic Analysis for Total Loss of Feedwater Event in PWR using SPACE Code ı MinJeong Kim, Minhee Kim, Junkyu Song, Bongsik Chu, Jae-Seung Suh and Hyunjin Lee

atw Vol. 65 (2020) | Issue 1 ı January

CFD Simulation of Flow Characteristics

and Thermal Performance in Circular

Plate and Shell Oil Coolers

Shen Ya-jie, Gao Yong-heng and Zhan Yong-jie

Circular plate and shell heat exchangers were gradually applied as oil coolers. Hence, it was necessary to investigate

their performance at low Reynolds number with high viscous oil. This paper provided a CFD simulation of flow

characteristic and thermal performance in circular plate and shell oil cooler with different plate parameters, such as plate

angle β, ratio of plate pitch to height p/h and corrugation styles. The fiction factor f and Colburn factor j were investigated

for the various plate parameters. The numerical results showed that f increased with increasing β, and both increased as

p/h decreased. When β

atw Vol. 65 (2020) | Issue 1 ı January

RESEARCH AND INNOVATION 30

Re 5-50

domain, as shown in Figure 1 (c).

There are quite a lot of contact points

in the channel. It can enhance bearing

capacity. It is also found that the outlet

of the calculation domain is extended,

which can effectively eliminate backflow.

Table 1 lists the operating conditions

and geometrical parameters,

which are typical used in industrial

application. The range of Reynolds

number is selected in accordance with

experimental condition.

Analytical conditions

β 15°, 30°, 45°, 60°, 75°

l/h 5.0, 3.3, 2.5, 2.0

Corrugation shape

Inclination corrugation,

Chevron corrugation

| Tab. 1.

Operation condition and geometrical parameters.

2.2 Governing equations

In oil loop, Reynolds numbers is low

atw Vol. 65 (2020) | Issue 1 ı January

| Fig. 2.

Grid sensitivity analysis.

(a) XY coordinates

| Fig. 3.

A partial view of the final grid.

3 Results and analyses

| Fig. 4.

Comparison of numerical results and experimental data.

(b) ZX coordinates

(a) f with respect to Re

(13)

(b) j with respect to Re

The valid range of the Reynolds

number for Eqs.(14) and (15) is from

5 to 50.

Comparison of numerical results

and experimental data is shown as

Figure 4(a) and (b). It is found that

the RNG k-ε model is more suitable

than laminar model in CPSHE. The

related difference of numerical results

of RNG k-ε model and experimental

data is within 15%. It is verified that

the results of simulation is reliable

during Re range from 5 to 50.

RESEARCH AND INNOVATION 31

3.1 Evaluation factors

The friction factor ƒ and Colburn

factor j are respectively considered as

evaluation factors of flow resistance

and heat transfer. JF factor is used to

evaluate the comprehensive performance

[11-13]. And ƒ, j and JF factors

are respectively defined as:

(9)

(10)

(11)

Where L is the length of the channel,

Nu is the Nusselt number, Pr is the

Prandtl number, μ o is the viscosity of

oil at the average temperature of oil,

μ w is the viscosity of oil at the average

temperature of the wall.

The valid range of the Reynolds

number for Eqs.(12) and (13) is from

5 to 50.

Numerical results with varying

inlet flow rate are collected and

analyzed. The criterion equation for

heat transfer and characteristic

equation for flow resistance in the

form of fanning friction coefficient are

obtained for the circular corrugated

plate, given as:

(14)

(15)

3.3 Corrugation angle

3.3.1 Bulk flow patterns

Figure 5(a)-(c) display the bulk flow

patterns for water with 15°≤β≤75°,

u = 0.35 m/s and Re = 30. When

atw Vol. 65 (2020) | Issue 1 ı January

RESEARCH AND INNOVATION 32

The zig-zag flow is that flowing fluid is

still mainly in the groove, but turning

points no longer appear in the left and

right sides of the plate, but occur in

corrugation contacts. The bulk flow

pattern is continuous parallel corrugations.

For β=75°, the bulk flow

pattern becomes zig-zag flow. This

phenomenon is in accordance with

the existing literature researches.

Figure 5(d)-(f) shows the bulk

flow patterns for oil with 15°≤β≤75°,

u=0.35 m/s and Re=1400. The bulk

flow pattern always remains zig-zag

flow with increasing β. It is not consistent

with that of water.

The bulk flow pattern mainly

depends on driving force and friction

force. The driving force F d comes from

that two sets of working fluid moving

along the grooves on the opposite

plates, one set of working fluid is

effected by the driving force F d from

(a) β=15°

the other one. The friction force F f

depends on the viscosity of working

fluid. When the working fluid is low

viscous, it can be ignored.

The viscosity of oil is much bigger

than that of water. The friction force F f

of oil so big that prevents oil moving

along the groove. In this case, the

driving force F d can easily drive oil to

turn to the groove of the opposite

plate at corrugation points. As a result,

zig-zag flow comes into being and

remains unchanged with increasing β.

3.3.2 Flow maldistribution

From Figure 6(a)-(c), flow distribution

is displayed as wave shape, and

wave crests appear at x=-0.065 m and

x=0.065 m. Furthermore, the wave

crests become flat with increasing β.

This is because for β45°, the component of

F d , along the flow direction, becomes

contrary to the flow direction. It

further hinders oil from moving

along the groove of one plate, which

| Fig. 6.

The bulk flow pattern of water and oil with different β.

(d) β=60°

(e) β=75°

(a) ƒ and j factors with respect to β

| Fig. 7.

ƒ, j and JF with 15°≤β≤75° and p/h=5.0.

(b) JF factor with respect to β

Research and Innovation

CFD Simulation of Flow Characteristics and Thermal Performance in Circular Plate and Shell Oil Coolers ı Shen Ya-jie, Gao Yong-heng and Zhan Yong-jie

atw Vol. 65 (2020) | Issue 1 ı January

makes characteristics of zig-zag flow

more substantial. These make catkin

shape become sparse and flow maldistribution

serious.

From Figure 6(a)-(e), the flow

maldistribution is at a minimum with

β=45°. The corrugation points reach

the maximum value. In addition, the

corrugation shape is not very steep,

allowing partial oil moving to the side

of circular corrugation plates possibly.

So flow distribution is improved

obviously, with β=45°.

3.3.3 Flow characteristics and

thermal performance

Figure 7(a) shows ƒ and j for inclination

angles with 15°≤β≤75° and

p/h=5.0. The value of ƒ increases

monotonically with increasing β.

For β60°.

(a) p/h=5.0 (b) p/h=3.3 (c) p/h=2.5 (d) p/h=2.0

| Fig. 8.

Bulk flow patterns for p/h=5.0, 3.3, 2.5 and 2.0.

Figure 7(b) shows JF with respect

to β. For Re

atw Vol. 65 (2020) | Issue 1 ı January

RESEARCH AND INNOVATION 34

(a) ƒ and j factors

| Fig. 10.

ƒ, j and JF factors with respect to inclination and chevron corrugation.

inclination and chevron corrugations

on the flow characteristics and

thermal performance. It is mainly

because of their bulk flow pattern –

zig-zag flow. Most of working fluid

turns to the groove of the opposite

plate at corrugation contacts in the

zig-zag flow. So the structure difference

of inclination and chevron corrugations

has barely influence on flow

and heat transfer. Therefore, ƒ, j and

JF are almost constant.

4 Summary

Comparison between results of

numerical simulations and experimental

data has verified that CFD

simulation is reliable for studies on

the corrugation CPSHE. The RNG k-ε

turbulence model has been validated

more preciously than the laminar

model in CPSHE at low Reynolds

number from 5 to 50. The corrugation

inclination angle β, ratio of pitch to

height p/h and corrugation styles

have been taken as major parameters

of the circular corrugated plate

influencing the performance of heat

transfer. Some conclusions are obtained

as follow:

(1) When Reynolds number is low

( 5-50) and p/h=5, the bulk flow

pattern is zig-zag flow, no matter

how much the corrugation angle

is.

(2) Flow maldistribution exists in

every channel, and it is at a

minimum with β=45°.

(3) The flow resistance and thermal

performance increases with increasing

β. When β>60°, increasing

rate of thermal performance is

low. The comprehensive performance

with β= 45° is the best at

the Re range from 5 to 50.

(4) The flow resistance and thermal

performance decreases with

reducing p/h. The comprehensive

performance with p/h=3.3 is

the best.

(5) There is nearly no difference

between inclination and chevron

corrugations in CPSHE at low

Reynolds number.

Nomenclature

ɑ [m 2 /s] Coefficient of thermal Diffusion

B [m] Plate width

De [m] Hydraulic diameter

Ƒ [-] Friction factor

G k [-] Generation of turbulence kinetic energy

H [m] Corrugation height

I [-] Turbulence intensity

J [-] Colburn factor

L [m] Corrugation length

Nu [-] Nusselt number

P, ΔP [kPa] Pressure, Pressure drop

Pr [-] Prandtl number

T [K] Temperature

u [m/s] Fluid velocity

v [m 2 /s] Kinematic viscosity

Greek symbols

Α [-] Turbulence Prandtl number

β [°] Inclination angle

Ρ [kg/m 3 ] Density

Subscripts

ε [-] ε equation

k [-] k equation

1 [-] x-component

2 [-] y-component

3 [-] z-component

o, w [-] Lubricating-oil, wall

References

[1] W.W. Focke, P.G. Knibbe, Flow visualization in parallel-plate

ducts with corrugated walls, J. Fluid Mech., 165 (1986):

73–77.

[2] G. Gaiser, V. Kottke, Flow phenomena and local heat and mass

transfer in corrugated passages, Chem. Eng. Technol., 12

(1989):400–405.

[3] A. Muley, R.M. Manglik, Experimental study of turbulent flow

heat transfer and pressure drop in a plate heat exchanger with

chevron plates, Journal of Heat Transfer, 121(1999):110-117.

[4] W.W. Focke, J. Zachariades, I. Olivier, The effect of the

corrugation inclination angle on the thermo hydraulic

performance of plate heat exchangers, Int. J. Heat Mass Transfer

28 (1985): 1469–1479.

[5] A.G. Kanaris, A.A. Mouza, S.V. Paras, Flow and heat transfer

prediction in a corrugated plate heat exchanger using a CFD

code, Chem. Eng. Technol., 8 (2006): 923-930.

[6] J. Lee, K.S. Lee, Flow characteristic and thermal performance in

chevron type plate heat exchangers, International Journal of

Heat and Mass Transfer., 78(2014): 699-706.

[7] W. Li, H.X. Li, G.Q. Li, Numerical and experimental analysis

of composite fouling in corrugated plate heat exchangers.

International Journal of Heat and Mass Transfer, 63 (2013):

351-360.

[8] S.M. Lee, K.Y. Kim, Thermal performance of a double-faced

printed circuit heat exchanger with thin plates, Journal of

Thermophysics and Heat Transfer, 28 (2014): 251-257.

[9] Z.J. Luan, G.M. Zhang, Flow resistance and heat transfer characteristics

of a new-type plate heat exchanger.

Journal of Hydrodynamics, 20 (2008): 524-529.

[10] V. Yakhot, S.A. Orczag, Renormalization group analysis of

turbulence, Basic theory. Scient. Comput, 1 (1986): 3-11.

[11] J.Y. Yun, K.S. Lee, Influence of design parameters on the

heat transfer and flow friction characteristics of the heat

exchanger with slit fins, Int. J. Heat Mass Transfer, 43 (2000):

2529–2539.

[12] M.S. Kim, J. Lee, Correlations and optimization of a heat

exchanger with offset-strip fins, Int. J. Heat Mass Transfer,

54 (2011): 2073–2079.

[13] J. Lee, K.S. Lee, Correlations and shape optimization in a

channel with aligned dimples and protrusions, Int. J. Heat

Mass Transfer, 64 (2013): 444–451.

Authors

(b) JF factor

Shen Ya-jie

Gao Yong-heng

Zhan Yong-jie

CNNP Nuclear Power Operations

Management Co Ltd

Jiaxing, China

Research and Innovation

CFD Simulation of Flow Characteristics and Thermal Performance in Circular Plate and Shell Oil Coolers ı Shen Ya-jie, Gao Yong-heng and Zhan Yong-jie

atw Vol. 65 (2020) | Issue 1 ı January

Research on Neutron Diffusion and

Thermal Hydraulics Coupling Calculation

based on FLUENT and its Application

Analysis on Fast Reactors

Xuebei Zhang, Chi Wang and Hongli Chen

The neutron diffusion equation is defined based on the User Defined Function (UDF) and the User Defined Scalar

(UDS) functions of the FLUENT. The neutron diffusion equation is solved iteratively by using the solver of the FLUENT

with the Finite Volume Method (FVM). At the same time, the mass, momentum and energy equations are solved

iteratively. At each iteration, the power distribution (flux distribution) obtained by the iteration of the neutron diffusion

equation is transferred to the thermal-hydraulics calculation and is used as the heat source term. At the same time, the

temperature distribution obtained from the thermal-hydraulics calculation is transferred to the neutron diffusion

calculation and the macroscopic cross sections of the materials are corrected to realize the coupling calculation of the

neutron diffusion and the thermal-hydraulics under the same solver of the FLUENT without needing to develop the

interface program and the computational cost is saved. 2D-TWIGL benchmark problem is calculated by the FLUENT

solver to verify the feasibility for the neutron diffusion. Through the modeling and calculation of the 5 x 5 PWR assembly

model, the calculation results are compared with the results of other programs to verify the feasibility of the coupling

method and the correctness of data transfer. Then this coupling method is applied to calculate the hot assembly of a

modular lead-cooled fast reactor (M 2 LFR-1000) to verify that the thermal-hydraulics characteristics (the maximum

fuel temperature and the maximum cladding outer surface temperature) are all within the corresponding thermalhydraulics

design limits.

RESEARCH AND INNOVATION 35

Key words: neutron diffusion and

thermal- hydraulics coupling; UDF

and UDS functions; 5 x 5 PWR

assembly; M 2 LFR-1000 hot assembly.

Traditionally, the best estimation

procedure is generally used in reactor

design and reactor safety analysis.

With the improvement of computer

performance and the development of

parallel computing, the high confidence

simulation of reactor has been

paid more attention in the research of

reactor design, scheme optimization

and safety analysis. Only by considering

the multi-physical feedback

in reactor simulation, can high confidence

simulation be realized. And

the neutron diffusion and thermalhydraulics

coupling calculation is an

important part of multi-physics

coupling calculation [1-2]. The actual

operation of the reactor is a process

of neutrons and thermal reciprocal

feedback. Temperature coefficient

(fuel temperature coefficient and

moderator temperature coefficient) is

an important factor for reactivity

control in normal operation of reactor

[3]. To achieve accurate calculation of

reactor operation and transient conditions,

the effects of fuel temperature,

moderator temperature and

density on local neutron flux and

system reactivity must be considered.

Computational Fluid Dynamics

(CFD) program FLUENT can realize

the fine simulation of reactor core and

fuel assembly by coupling the mass

continuity equation, momentum

equation and energy conservation

equation. The UDS (User Defined

Scalar) in the FLUENT can solve a kind

of diffusion equation by using the

solver in Fluent. It has been widely

used in multi-phase flow coupling

calculation and flow-field and electric

field coupling calculation. H.G.Wang,

W.Q.Yang, P.Senior [4] used the UDS

to add the water diffusion equation in

air and solid phase to FLUENT. And the

hydrodynamic parameters of heat and

mass transfer between two phases

were added to the UDF of FLUENT

to simulate the complex gas-solid

multi phase process of batch fluidized

bed drying. P. Donoso- GarcaL, L.

Henrquez- Vargas [5] used the twodimensional

numerical simulation

method to simulate the turbulent state

of the adiabatic regenerative porous

medium burner coupled with thermoelectric

components. The time and

volume averaged transport equation

and the two order turbulence model

were adopted. The FLUENT was used

to simulate the burner through its UDF

(user-defined function) and UDS

(user- defined scalar) interface to

obtain additional terms involving

turbulence and thermal energy. The

flow field and electric field were

calculated considering the effects

of inlet velocity and composition,

thermal conductivity of porous media

and thermal insulation materials on

the burner. Y. Liu, Y. P. Liu, S. M. Tao

[6] established a three-dimensional

(3D) unsteady mathematical model of

alumina ball regenerator, and solved

it by commercial computational fluid

dynamics (CFD) software FLUENT

based on the porous medium hypothesis.

The standard K-e turbulence

model was combined with standard

wall function to simulate gas flow

and the momentum equation was

modified to consider the effect of

porous media on fluid flow. The user

defined function (UDF) program was

programmed in C language and

connected with FLUENT. The userdefined

scalar (UDS) transfer equation

of solid energy conservation was

defined. And the heat transfer and

thermo-physical properties between

gas and solid phases were calculated.

J. Jang, H. Arastoopour [7] used

ANSYS/FLUENT computational fluid

dynamics (CFD) program to simulate

the gas-solid two-phase flow pattern,

the mixing and drying process of drug

particles in three different scales of

bubbling fluidized bed dryers. The

capacity of water transfer and

simulation of drying process was

calculated. The user-defined scalar

transfer equation (UDS) was added to

FLUENT to simulate the flow pattern

and heat and mass transfer process

of drug drying process based on bubbling

fluidized bed. Based on the UDF

(User Defined Function) and UDS

(User Defined Scalar) functions of

FLUENT, Xi’an Jiaotong University

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atw Vol. 65 (2020) | Issue 1 ı January

RESEARCH AND INNOVATION 36

| Fig. 1.

The flow chart of coupling calculation.

developed the TASNAM program [8],

which is mainly used for numerical

calculation of neutron diffusion in

molten salt reactor. Based on the

user interface programming function

of commercial software CFX,

Naval Engineering University added

three-dimensional space-time neutron

dynamics model, coupled with

CFD thermal-hydraulics, and simulated

the local three- dimensional flow

behavior and three- dimensional physical

characteristics of PWR under

steady state [9].

Based on UDF and UDS functions

of FLUENT, this paper defines the

neutron diffusion equation and uses

the modeling tool (GAMBIT) and

solver in FLUENT to solve the neutron

diffusion equation iteratively, and

carries out thermal-hydraulics calculation

at the same time. In each

iteration, thermal power is transferred

to thermal- hydraulics calculation by

solving neutron diffusion equation.

The temperature obtained by thermalhydraulics

calculation is transferred to

the neutron diffusion calculation, and

the macroscopic cross sections of the

materials ware modified until the

convergence of the iterative calculation

of neutron diffusion equation

and thermal-hydraulics iteration is

achieved. The iterative flow chart of

coupling calculation is shown by

Figure 1. In order to verify the

correctness of the coupling calculation

method and data transfer, the

5 x 5 PWR assembly model [10] is

modeled and calculated, and the

results are compared with other

coupling programs. Then the hot

assembly of the M 2 LFR-1000 [11] is

modeled and calculated. And the

neutron flux, temperature distri bution

and the thermal-hydraulics characteristics

(the maximum fuel temperature

and the maximum cladding outer

surface temperature) on the steady

state has good agreement with the

results calculated by sub-channel

program (KMC-SUB) [12]. In this

paper, the calculation method and

mathematical model are introduced

in the section 2. Section 3 describes

the calculation of 2D-TWIGL [13]

benchmark problem by the FLUENT

solver. And the section 4 describes the

coupling calculation of 5 x 5 PWR

assembly. The calculation method is

applied to the hot assembly of the

M 2 LFR-1000 in section 5. Section 6

summarizes the general conclusion.

2 Calculation method and

mathematical model

2.1 The UDS module

of the FLUENT

The UDS module of the FLUENT can

define a kind of equation and solve it

by the inner solver with the finite

volume method. The form is shown in

formula (1):

(1)

The definitions of equations in the

FLUENT are shown in Table 1.

The transient neutron diffusion

equation is shown in equation (2):

(2)

The first term on the left side of the

equation (2) is unsteady state term,

the second term on the left side is

diffusion term, and the term on the

right side of the equation is the source

term.

For steady state calculation, the

unsteady term in the equation is

neglected. And the equation is shown

in equation (3):

(3)

For the equation (3), φ g (r) represents

the neutron flux, unit cm -2 s -1 , represents

the neutron diffusion coefficient,

unit cm, ∑ f represents the macro scopic

fission cross section, the unit is cm -1 ,

Name Expression Definition Corresponding functions in UDS

Unsteady-state term Discrete form of unsteady state term DEFINE_UDS_UNSTEADY

Convection term Flux ( ) DEFINE_UDS_FLUX

Diffusion term Diffusivity (Γ(T)) DEFINE_DIFFUSIVITY

Boundary condition Value ( ) Specified Value

Flux ( ) Specified Flux

| Tab. 1.

Definition of equation in the UDS of FLUENT.

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∑ g'→g represents the macroscopic

transfer cross section, the unit is cm -1 ,

∑ r represents the macroscopic removal

cross section, the unit is cm -1 , v

represents the average number of

neutrons emitted per fission, χ g represents

the fission spectrum.

The effective multiplication factor

is calculated by equation (4):

(4)

2.2 Numerical Method [14]

Finite Volume Method (FVM) is

widely used in CFD methodology to

discretize governing equations and is

adopted by almost all the popular CFD

softwares. FLUENT converts a general

transport equation to an algebraic

equation using a control-volumebased

technique which consists of

integrating the general transport

equation on each discrete control

volume. For a general scalar, ϕ the

integral form of a transport equation

on a control volume V can be illustrated

as follows [15]:

(5)

Where ρ is density, Γ ϕ is the effective

diffusion coefficient for the scalar ϕ,

S ϕ is the source term of ϕ per unit

volume.

Applying Equation (5) to each

control volume, the discretization

equation on each given cell is:

(6)

Where N is the number of the faces

enclosing a cell; ϕ f is the value of ϕ at

the cell face, A f is the surface area

vector, which means that its direction

is normal to the surface and | → A f | is

the area of the surface, ∇ ϕ f is the

gradient of ϕ at the face f and V is the

cell volume. The equations given

above can be applied to multidimensional,

unstructured meshes composed

of arbitrary polyhedral in

FLUENT.

For the steady state neutron

diffusion equation (3), applying equation

(6) to each control volume, the

discretization equation on each given

cell is:

(7)

Region Group D g (cm -1 ) ∑ a,g (cm -1 ) υ∑ f,g (cm -1 ) ∑ 1→2 (cm -1 )

1 1 1.4 0.01 0.007 0.01

2 0.4 0.15 0.2

2 1 1.4 0.01 0.007 0.01

2 0.4 0.15 0.2

3 1 1.3 0.008 0.003 0.01

2 0.5 0.05 0.006

| Tab. 2.

Section parameter3 of 2D-TWIGL.

2.3 Thermal-hydraulics

calculation model

The energy equation of coolant region

is shown in equation (8):

(8)

The equation of heat conduction in

fuel area is shown in equation (9):

(9)

(10)

γ is the release energy of each fission.

The heat conduction equation of the

cladding and gap is shown in equation

(11), (12) respectively:

(11)

(12)

2.4 Material macroscopic cross

section library

Under the condition of knowing

nuclear density and considering the

energy group emerging and the

influence of temperature on material

density, the nuclear database program

[16] developed by the author’s

research group calculates one set of

group-wise neutronics parameters

including the group-wise macroscopic

cross sections, the diffusion coefficients

(D g ), the neutron fission yields

(ν g ) and the fission spectrum (χ g ) at

400 K, 500 K, 600 K, 700 K, 800 K,

900 K, 1000 K, 1100 K, 1200 K, 1300 K,

1400 K, 1500 K and 1600 K of fuel, air

gap, cladding and coolant. The continuous

macroscopic cross sections

at 400 K to 1600 K of materials can get

through interpolation [17]. Then, the

macroscopic cross-section distribution

functions of temperature simply

shown by equation (13) are added to

FLUENT solver through UDF, and the

macroscopic cross-section of each

grid is updated by reading the new

temperature distribution of each grid

after each iteration.

(13)

3 Validation of the FLUENT

solver for neutron

diffusion

In this section, calculation results for

2D-TWIGL benchmark problem are

presented and compared with reference

values to prove that the FLUENT

solver is feasible to solve the neutron

diffusion based on the UDS and UDF

function. The meshes adopted in this

paper are generated by the general

mesh generation tool Gambit. The

mesh independent solutions are

obtained but not presented here.

3.1 2D-TWIGL seed blanket

problem

The 2D-TWIGL benchmark problem is

a simplified neutron kinetics model

with two neutron energy groups and

one delayed neutron precursor family.

A steady state calculation and two

transient calculations are included in

this problem.

The reactor core is a 160 cm square

consisting of three regions including

(1) perturbed seed region containing

primary fissile materials with timedependent

properties in the transient

situation; (2) unperturbed seed

regions containing primary fissile

materials with constant properties in

the transient situation; (3) a blanked

region also containing fissile materials

surrounding the whole core. Due to

the symmetry, one quadrant of the

reactor is modeled for the calculation,

as displayed in Figure 2. The group

constants are given in Table 2.

| Fig. 2.

2D-TWIGL 1/4 core model.

RESEARCH AND INNOVATION 37

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RESEARCH AND INNOVATION 38

| Fig. 3.

Meshes of the 2D-TWIGL by the solver

of FLUENT.

Mesh size is 0.1 cm and meshes are

produced by GAMBIT, a part of the

meshes is shown by Figure 3.

3.2 Result analysis

The results obtained by the FLUENT

solver are in good agreement with

the reference values. The effective

multiplication factor calculated under

steady state is 0.913306, which is very

close to the reference value 0.913214.

The relative error is 10 pcm. Figure 4

and Figure 5 shows the fast and

thermal neutron flux distributions on

the diagonal line of the calculation

domain respectively. Figure 6 shows

the comparison between the normalized

power and reference value of

the calculated assemblies. We can find

that the errors mainly occur on the

boundary and the interface of different

materials.

4 Validation of the

coupling method

4.1 Model introduction

In this paper, the neutron diffusion

and thermal-hydraulics coupling

(%)

1.258

1.250

-0.635

1.321

1.342

1.5897

1.293

1.301

0.6187

1.198

1.194

-0.333

1.259

1.264

0.3971

1.243

1.250

0.5631

2.187

2.195

0.3657

2.350

2.352

-0.085

2.380

2.378

-0.084

2.373

2.372

-0.042

1.870

1.878

0.4278

2.033

2.044

0.5410

2.123

2.132

0.4221

2.161

2.172

0.5090

2.176

2.187

0.5055

calculation method based on UDF and

UDS functions of the FLUENT is used

to calculate the 5 x 5 PWR assembly

model. The thermo-physical properties

of the materials are given by Table

3, the model structure and parameters

are given by Figure 7 and Table 4,

and the change of coolant density and

specific heat with tem perature is given

by equation (14) and equation (15).

1.380

1.3801

0.0072

1.614

1.617

0.1858

1.779

1.783

0.2248

1.883

1.888

0.2655

1.961

1.945

-0.815

1.967

1.971

0.2033

0.948

0.936

-1.265

1.148

1.138

-0.871

1.350

1.339

-0.814

1.500

1.488

-0.800

1.602

1.589

-0.811

1.663

1.649

-0.841

1.691

1.676

-0.887

0.260

0.2593

-0.269

0.343

0.340

-0.874

0.432

0.427

-1.157

0.509

0.503

-1.178

0.568

0.561

-1.232

0.609

0.602

-1.149

0.635

0.628

-1.102

0.647

0.639

-1.236

0.093

0.092

-1.075

0.157

0.1566

-0.254

0.220

0.2198

-0.090

0.279

0.278

-0.358

0.329

0.328

-0.303

0.368

0.367

-0.271

0.396

0.395

-0.252

0.414

0.412

-0.483

0.422

0.420

-0.473

0.010

0.0099

-1.000

0.030

0.0302

0.666

0.051

0.0517

1.3725

0.072

0.0716

-0.555

0.091

0.0906

-0.439

0.108

0.107

-0.925

0.121

0.120

-0.826

0.130

0.129

-0.769

0.136

0.134

-1.470

0.139

0.137

-1.438

| Fig. 6.

Normalized power diagram for 2D-TWIGL assemblies.

| Fig. 4.

Fast neutron flux.

| Fig. 7.

Radial and axial Geometric structure

of the assembly.

| Fig. 8.

Radial mesh of the 5 x 5 assembly fuel rod.

| Fig. 5.

Thermal neutron flux.

| Fig. 9.

Radial mesh of the 5 x 5 assembly.

| Fig. 10.

Mesh quality check.

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Materials

Density

(g/cm3)

| Tab. 3.

Thermal properties of the assembly materials.

Thermal conductivity

(W/m.K)

Specific heat capacity

(J/kg.K)

Fuel (UO2) 10.3 3.0 310

Cladding (Zircaloy-2) 6.5 11.0 330

Viscosity

(Pa.s)

Coolant (Water) 0.53 0.00009177

Gap (Helium) 0.0001625 0.152 5193

Fuel pin radius

Cladding inner radius

Cladding outer radius

Pitch

Fuel height

Bottom reflector height

Top reflector height

Fuel

Coolant

4.1 mm

4.2 mm

4.8 mm

12.5 mm

3 m

0.2 m

UOX

(2 %, 4 %

enrichment)

Water,

1000 ppm

boron

RESEARCH AND INNOVATION 39

Gap

Cladding

Power

Helium,

0.1 MPa

Zircaloy-2

12.5 MW

| Fig. 11.

Axial power density distribution the fuel rod center.

Energy group 2

Energy boundary

0.625 eV

| Tab. 4.

Size and material parameters of the assembly.

4.2 Modeling and

mesh generation

The gambit is used to model and mesh

the fuel assembly model of 5 x 5 PWR.

The radial mesh of fuel rod is shown

in Figure 8. The radial mesh of fuel

assembly is shown in Figure 9. The

same meshes are used for neutron

diffusion and thermal-hydraulics

calculation. The axial mesh size is

0.1 m and the total mesh number of

the assembly model is 3.10E+06. The

mesh quality checking tool in the

Gambit is used to check the assembly

meshes. As shown in Figure 10, there

are 99.32 % of the meshes whose

EquiSize Skew ranges from 0 to 0.4.

| Tab. 5.

Boundary conditions of the coupling calculation of the 5 x 5 assembly.

4.3 Boundary conditions

The coupled calculation boundary

conditions are shown in Table 5.

4.4 Calculation results and

analysis

The reference value of effective multiplication

factor of the module is

1.17109 [10], and the effective multiplication

factor calculated in this

paper is 1.17100. Figure 11 shows the

axial power density distribution of

fuel rod center with 2 % and 4 %

enrichment when the neutron diffusion

and thermal- hydraulics calculation

converge. The blue and black

Field Boundary Type Value

Temperature (T) Inlet Constant value 540 K

Outlet

Zero gradient

Neutron flux (f) Inlet Extrapolation boundary Gradient on boundary

Outlet Extrapolation boundary Gradient on boundary

Pressure (P) Inlet Zero gradient

Outlet Constant value 15.5 MPa

Velocity (U) Inlet Constant value (0,0,3) m/s

Outlet

Zero gradient

(kg/m 3 ) (14)

(J/kg.K) (15)

points are the results of other coupling

programs. The red and green lines

are the results of this paper. The

maximum power density of the 4 %

en richment fuel rod center is about

4.25E8 W/m 3 and the maximum

power density of the 2 % enrichment

fuel rod center is about 2.50E+08 W/

m 3 in this paper. And the reference

maximum power density of the 4 %

enrichment fuel rod center is about

4.178E+08 W/m 3 and the reference

maximum power density of the 2 %

enrichment fuel rod center is about

2.45E+08 W/m 3 .

It can be seen from Figure 11 that

the calculation deviation is mainly at

the inlet and outlet of the assembly

model. The reference value of the

power density increases slightly at the

inlet and outlet of the assembly, which

is mainly due to the influence of the

upper and lower reflectors, which

make some neutrons be reflected into

the fuel area, then cause a slight

increase of the power density. In this

paper, the influence of the upper and

lower reflectors is neglected due

to considering the convenience of

modeling and meshing by the Gambit.

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RESEARCH AND INNOVATION 40

| Fig. 12.

Temperature distribution of fuel rod

outer diameter.

| Fig. 13.

Temperature distribution of fuel

cladding inner diameter.

Figure 12 gives the temperature

distribution of the fuel pellet outer

diameter, Figure 13 gives the temperature

distribution of the fuel cladding

inner diameter, Figure 14 gives the

temperature distribution of the fuel

cladding outer diameter and Figure

15 gives the temperature distribution

of the assembly inlet and outlet.

Figure 16 shows the axial coolant

temperature distribution of the adjacent

fuel rod center and diagonal fuel

rod center, Figure 17 shows the

temperature distribution along the X

axis direction (z=0.0 m, y=0.0 m,

| Fig. 14.

Temperature distribution of fuel

cladding outer diameter.

z=0.0 m is symmetry axis). And they

are all compared with reference

values calculated by other coupling

programs and in good agreement with

them. The maximum temperature of

the 4 % enrichment fuel rod center is

1506.97 K, and the maximum temperature

of the 2 % enrichment fuel

rod center is 1066.42 K. And the

reference maximum temperature

of the 4 % enrichment fuel rod center

is 1502.22 K, and the reference

maximum temperature of the 2 % enrichment

fuel rod center is 1047.60 K.

Through the coupling calculation of

| Fig. 16.

Axial coolant temperature distribution of the contiguous fuel rod center and diagonal fuel rod center.

| Fig. 17.

Temperature distribution in the X axis direction (z = 0.0 m, y = 0.0 m).

| Fig. 15.

Temperature distribution of assembly

inlet and outlet.

the 5 x 5 PWR assembly, it is proved

that this coupling method is feasible

and the data transfer is correct.

5 Application of the

coupling method

The coupling calculation of 5 x 5 PWR

assembly model proves that it is

feasible to realize the coupling calculation

of neutron diffusion and

thermal- hydraulics by utilizing UDF

and UDS functions of FLUENT. Now

the M 2 LFR-1000 hot assembly model

is calculated by this coupling method.

5.1 Model description

M 2 LFR-1000 is a modular lead-cooled

fast reactor. The structure of components

and fuel rods is given by

Figure 18 and Figure 19 respectively.

The fuel rods in the core fuel assembly

are arranged in a regular triangular

matrix, and the bundles are hexagonal.

The bundles are wrapped in the

assembly box with a thickness of

4 mm. The distance between the fuel

rods is 14 mm. Each fuel assembly

contains 169 fuel rods. The inner

margin of the fuel assembly box is

185 mm, the outer margin is 193 mm,

and the component center distance is

198 mm. The core of the fuel pellet has

a central hole with a diameter of

1.9 mm, which can reduce the core

temperature and improve the core

safety margin under the condition of

the same linear power density of the

fuel rod. The outer diameter of the

fuel pellets is 8.6 mm, and the MOX

fuel is added with a small amount of

MA or without MA.

There is a gap of 0.15 mm between

fuel pellet and cladding, which is filled

with He of ~0.5 MPa. The gas pressure

is higher than the operating

pressure of the primary circulation.

This can improve the gap heat conduction

between fuel pellet and cladding

and provide inert environment.

On the other hand, it can prevent

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| Fig. 18.

Fuel assembly cross section (unit: mm).

| Fig. 19.

Fuel rod cross section.

Nuclide

Nucleon density

(b - cm - )

| Tab. 6.

The nuclide composition of the FMS T91.

Nuclide

| Tab. 7.

Boundary conditions of the coupling calculation of the M 2 LFR-1000 assembly.

Nucleon density

(b -1 cm -1 )

C 3.8900E-04 N 1.6600E-04

Cr 7.8690E-03 P 3.0200E-05

Ni 1.5948E-04 S 7.2845E-06

Mn 3.8300E-04 Cu 7.3600E-05

Mo 4.6270E-04 V 1.9700E-04

Si 5.8230E-04 Al 6.9300E-05

Nb 4.0200E-05 Fe 7.4232E-02

Field Boundary Type Value

Temperature (T) Inlet Constant value 673 K

Neutron flux (f) Outlet Extrapolation boundary Gradient on boundary

Inlet Extrapolation boundary Gradient on boundary

Pressure (P) Outlet Constant value 101 kPa

Velocity (U) Inlet Constant value (0,0,1.66) m/s

RESEARCH AND INNOVATION 41

cladding from contacting with pellet

due to external pressure and creep

collapse. The cladding damage can

also be tested. The fuel cladding is

FMS T91 with thickness of 0.55 mm,

the diameter of the whole fuel rod is

10.0 mm, and the length of the active

zone of the fuel rod is 1000 mm.

FMS T91 with good comprehensive

performance is selected as the core

structure and cladding material.

The nuclide composition of FMS T91

is shown in Table 6 [18]. And the

coolant is Pb.

The temperature of core inlet coolant

is set as 673.15 K and the temperature

of core outlet coolant is set

as 753.15 K. Under all operational

con ditions including design basis

accidents (DBAs), the maximum fuel

pellet temperature should be lower

than 2946.15 K [19]. Under normal

con ditions, the maximum cladding

temperature should be lower than

823.15 K [20] with sufficient safety

margin.

Considering the computational

cost, the 1/6 assembly is selected to

be modeled and calculated. The

assembly is the hot assembly and the

power is 3.6 MW. The M 2 LFR-1000

assembly is modeled and meshed by

the Gambit. The axial mesh size is

0.05 m, and the total mesh number is

3.40E+06. The radial mesh of the fuel

rod and the assembly are shown by

Figure 20 and Figure 21 respectively.

The meshes are checked by Gambit’s

grid quality checking tool. There are

96.67 % of meshes whose EquiSize

Skew ranges from 0 to 0.4.

| Fig. 20.

Radial mesh of the M 2 LFR-1000 fuel rod.

5.3 Boundary conditions and

properties of the materials

[21]

The boundary conditions are showed

by Table 7.

5.4 Calculation results and

analysis

Figure 22 and Figure 23 show the

unnormalized fast neutron flux distribution

on the outlet and the unnormalized

thermal neutron flux

distribution on the outlet respectively.

Because of the fission in the fuel

region, the fast neutron is mainly in

the fuel region and the thermal

neutron is mainly in the coolant

region. And great changing of the fuel

temperature which makes the macroscopic

cross sections of fuel region

change greatly leads to apparent

changing of the fast neutron flux and

thermal neutron flux distribution in

this region. However, for the coolant

region, the operation temperature is

673 K to 753 K and the macroscopic

cross sections of Pb hardly change.

| Fig. 21.

Radial mesh of the M 2 LFR-1000 assembly.

| Fig. 22.

Fast neutron flux distribution on the outlet.

| Fig. 23.

Thermal neutron flux distribution on the outlet.

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So the fast neutron flux and thermal

neutron flux dis tribution on the

coolant change a little.

Thermal conductivity of the MOX (W/m.K)

(16)

Density of the MOX (kg/m 3 )

(17)

Thermal conductivity of the T91 (W/m.K)

(18)

Specific heat capacity of the T91 (J/kg.K)

Thermal conductivity of the Pb (W/m.K)

Density of the Pb (kg/m 3 )

(19)

(20)

(21)

Specific heat capacity of the Pb (J/kg.K)

(22)

Viscosity of the Pb (Pa.s)

(23)

Figure 24, Figure 25, Figure 26 show

the unnormalized fast neutron flux,

unnormalized thermal flux and

temperature distribution on the outer

boundary respectively. Figure 27

shows the fuel pellet centerline

temperature distribution along Z axis

direction and there is the peak

temperature when Z=0.6 m. And the

maximum fuel pellet centerline

temperature deviation is 21 K which

occurs on the outlet. Figure 28 shows

the assembly temperature distribution

along Y axis direction when Z=0.6 m.

Figure 29 shows the coolant temperature

distribution along Z axis

direction. Figure 30 displays the

cladding outer surface temperature

dis tribution along Z axis direction. It

is obvious that the central hole makes

the central fuel pellet temperature

distribution flat, decreases the peak

temperature of fuel pellet and improve

the safety margin. And there is

the maximum fuel temperature in the

central fuel rod of the hot assembly.

The coolant outlet temperature is

755.11 K. The maximum fuel temperature

is 1643.41 K and the maximum

cladding outer surface temperature is

773.83 K calculated by coupling calculation.

And the maximum fuel

temperature is 1647.17 K and the

maximum cladding outer surface

temperature is 777.28 K calculated by

sub-channel code (KMC-SUB) which

are all within the corresponding

thermal- hydraulics design limits.

| Fig. 24.

Fast neutron flux distribution

on the outer boundary.

| Fig. 25.

Thermal neutron flux distribution

on the outer boundary.

| Fig. 27.

The fuel pellet centerline temperature distribution along Z axis direction.

| Fig. 26.

Temperature distribution

on the outer boundary.

6 Conclusion

In this study, based on the UDF and

UDS function of the FLUENT, the

neutron diffusion equation is defined.

There is no requirement to develop

the interface program of the coupling

calculation. Then the assembly is fine

modeled and the solver in the FLUENT

is used to solve the neutron diffusion

equation. The thermal-hydraulics

calculation is carried out at the same

time. Therefore the coupling calculation

between neutron diffusion

and thermal-hydraulics is achieved on

the same solver of the FLUENT. In

order to achieve the convenient data

transfer, the neutron diffusion and

thermal-hydraulics calculation use

the same meshes.

Through calculating the 2D-TWIGL

benchmark problem by the FLUENT

solver based on the Finite Volume

Method (FVM), and comparing the

effective multiplication factor, neutron

flux and power with reference

values to verify that the solver can

be used to calculate the neutron

diffusion. The errors mainly occur on

Research and Innovation

Research on Neutron Diffusion and Thermal Hydraulics Coupling Calculation based on FLUENT and its Application Analysis on Fast Reactors ı Xuebei Zhang, Chi Wang and Hongli Chen

atw Vol. 65 (2020) | Issue 1 ı January

| Fig. 28.

Temperature distribution along the Y axis direction (Z=0.6m, X=0.0m).

| Fig. 29.

The coolant temperature distribution along Z axis direction.

| Fig. 30.

The cladding outer surface temperature distribution along Z axis direction.

the boundary and the interface of

different materials. So further work is

needed to refine the mesh on the

boundary and interface areas.

Through modeling and calculating

the 5x5 PWR assembly:

(1) The effective multiplication factor

(1.17100) which has good agreement

with the reference value

(1.17109), axial power density

distribution of the fuel rod center,

temperature distribution of the fuel

pellet outer diameter, tem perature

distribution of the fuel cladding

inner diameter, tem perature distribution

of the fuel cladding outer

diameter, temperature distribution

of the assembly inlet and outlet are

obtained on the steady state.

(2) The axial coolant temperature distribution

of the adjacent fuel rod

center and diagonal fuel rod center,

and the temperature distribution

along the X axis direction

(z = 0.0 m, y = 0.0 m) and axial

power density distribution of the

fuel rod center are compared with

reference values calculated by

other coupling programs and in

good agreement with them. Therefore

this coupling method is feasible

to achieve neutron diffusion

and thermal- hydraulics coupling.

And the correctness of the data

transfer is verified.

(3) The reference value of the power

density increases slightly at the

inlet and outlet of the assembly,

which is mainly due to the influence

of the upper and lower reflectors,

which make some neutrons be

reflected into the fuel area, and

then cause a slight increase of the

power density. In this paper, the

influence of the upper and lower

reflectors is neglected due to considering

the convenience of modeling

and meshing by the Gambit.

Therefore the power density distribution

is flat at the inlet and

outlet of the assembly. Further

work is needed to add the modeling

of the upper and lower reflectors.

Then the coupling method is applied

to a modular lead-cooled fast reactor

(M 2 LFR-1000). Through modeling

and calculating the hot assembly of

the M 2 LFR-1000:

(1) The neutron flux and temperature

distribution of the hot assembly are

obtained on the steady state. The

fast neutron is mainly in the fuel

region and the thermal neutron is

mainly in the coolant region. And

the fast neutron flux and thermal

neutron flux distribution on the

coolant region change a little. But

they change a lot on the fuel

region.

(2) It is obvious that the central hole

makes the central fuel pellet

temperature distribution flat,

decreases the peak temperature of

fuel pellet and improves the safety

margin at the same power density.

(3) The maximum fuel temperature

and the maximum cladding outer

surface temperature are obtained

by the coupling calculation and are

compared with the reference

values calculated by sub-channel

code (KMC-SUB). The error of

maximum fuel temperature is

1.25 K and error of maximum

cladding outer surface temperature

is 2.68 K. The coolant outlet

temperature is 755.11 K which is

very close to the design value

(753.15 K). And these thermalhydraulics

characteristics are all

within the corresponding thermalhydraulics

design limits.

RESEARCH AND INNOVATION 43

Research and Innovation

Research on Neutron Diffusion and Thermal Hydraulics Coupling Calculation based on FLUENT and its Application Analysis on Fast Reactors ı Xuebei Zhang, Chi Wang and Hongli Chen

atw Vol. 65 (2020) | Issue 1 ı January

RESEARCH AND INNOVATION 44

References:

[1] J.A. Kulesza, F. Franceschini, T.M. Evans, et al, Overview of

the consortium for the advanced simulation of light water reactor

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(NURESIM) [R]. Integrated Project NUCTECH-2004-3. 4. 3.

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[3] Z.S. Xie. Physical Analysis of Nuclear Reactor [M]. Xi’an: Xi’an

Jiao Tong University press, 2004.

[4] H.G. Wang, W.Q. Yang, P. Senior et al. Investigation of batch

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[5] P. Donoso-GarcíaL, L. Henríquez-Vargas. Numerical study of

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[6] Y. Liu, Y.P. Liu, S.M. Tao et al. Three-dimensional analysis of

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balls [J]. Applied Thermal Engineering, 2014, 69:113-122.

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[7] J. Jang, H. Arastoopour. CFD simulation of a pharmaceutical

bubbling bed drying process at three different scales [J].

Powder Technology, 2014, 263:14-25.

doi: 10.1016/j.powtec.2014.04.054

[8] D.L. Zhang a, b, S.Z. Qiu a, b,*, G.H. Sua, b, C.L. Liu b. Development

of a steady state analysis code for a molten salt reactor

[J]. Annals of Nuclear Energy, 2009.36:590-603.

[9] X.W. Gui, Q. Cai, Y.Q. Chen. Study on coupling of local threedimension

flow model based on CFD method and space-time

neutron kinetics model. Chinese Journal of Nuclear Science

and Enginee ring, 2010 (3): 216-222.

[10] Klas Jareteg, Paolo Vinai, Christophe Demazière. Fine-mesh

deterministic modeling of PWR fuel assemblies:Proof-ofprinciple

of coupled neutronic/thermal–hydraulic calculations

[J]. Annals of Nuclear Energy, 2014, 68:247-256.

[11] Chen H, Zhang X, Zhao Y, et al. Preliminary design of a

medium-power modular lead-cooled fast reactor with the

application of optimization methods. Int J Energy Res. 2018;

42:3643–3657.

[12] Li S, Cao L, Khan MS, Chen H. Development of a sub-channel

thermal hydraulic analysis code and its application to lead

cooled fast reactor. Appl Therm Eng. 2017; 117:443-451.

[13] Ahmad Pirouzmand-Abolhasan Nabavi. Simulation of

nuclear reactor dynamics equations using reconfigurable

computing [J]. Progress in Nuclear Energy, 2016.89:197-203.

[14] Jian Ge, Dalin Zhang, Wenxi Tian, et al. Steady and transient

solutions of neutronics problems based on finite volume

method(FVM) with a CFD code[J]. Progress in Nuclear Energy,

2015, 85: 366-374.

[15] ANSYS, 2013. ANSYS FLUENT Theory Guide, Release 15.0.

[16] X.B. Zhou, Y.S. Zhao, H.T Fan et al. Development and preliminary

test of date library ANDL-ADS for accelerator-driven

systems [J]. Nuclear Techniques, 2018, 41(03):65-70.

[17] G.W. Bi. Interpolation method development for temperature

based neutron cross-sections. Beijing: Tsinghua University,

2008.

[18] David Jaluvka. Development of a Core Management Tool for

the MYRRHA Irradiation

Research Facility [D]. KU Leuven, 2015.

[19] Carbajo JJ, Yoder GL, Popov SG, Ivanov VK. A review of the

thermophysical properties of MOX and UO2 fuels. J Nucl

Mater. 2001;299(3):181-198.

[20] Chen Z. Thermal-hydraulics design and safety analysis of a

100MWth small natural circulation lead cooled fast reactor

SNCLFR-100. University of Science and Technology of China,

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[21] Popov, S.G., Carbajo, J.J., Ivanov, V.K., Yoder, G.L., 2000.

Thermophysical properties of MOX and UO2 fuels including

the effects of irradiation. ORNL Report TM-2000/351.

Authors

Xuebei Zhang

Chi Wang

Hongli Chen

School of Physics,

University of Science & Technology

of China

Hefei 230027

China

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atw Vol. 65 (2020) | Issue 1 ı January

Programme Overview

PROGRAMME OVERVIEW

Plenarsitzung | Plenary Session

5. Mai 2020

09:00 D/E

Begrüßung und Eröffnungsansprache

| Welcome and Opening Address

Dr. Joachim Ohnemus

Vorsitzender des Vorstands, KernD

11:25 DE

System-Know how – der Schlüssel für die Zukunft

der nuklearen Kompetenz | System-oriented knowhow

– the key to the future of nuclear competence

Wolfgang Däuwel

Framatome GmbH, Germany

KERNTECHNIK 2020

Politik | Policy

09:15 D

Sicherer Kernkraftwerksbetrieb: Wie kann

Deutschland nach 2022 international noch Gehör

finden? | Safe Operation of Nuclear Power Plants:

How Can Germany Still Be Heard Internationally After

2022?

Andreas Feicht

Staatssekretär im Bundesministerium für Wirtschaft und Energie

(BMWi)

09:35 D

Kernenergiepolitik in der Schweiz – Wie geht es

weiter? | Nuclear Energy Policy in Switzerland –

What's Next?

Hans-Ulrich Bigler

Präsident, Nuklearforum Schweiz

09:55 D

Wirtschaftsstandort Deutschland – Welchen Beitrag

kann die kerntechnische Industrie leisten?

| Business Location Germany – What Contribution Can

Be Made by the Nuclear Industry?

Karlheinz Busen MdB

Stellvertretendes Mitglied im Ausschuss für Umwelt, Naturschutz und

nukleare Sicherheit, Deutscher Bundestag

Endlagerung | Waste Management

11:45 E

Creating Public Acceptance for a Final Repository

Jussi Heinonen

Director of the Nuclear Waste and Material Regulation Department,

STUK – Radiation and Nuclear Safety Authority, Finland

12:05 D

Ansprache

Karsten Möring MdB

Ordentliches Mitglied im Ausschuss für Umwelt, Naturschutz und

nukleare Sicherheit, Deutscher Bundestag

12:25 D

Aktueller Stand im Standortauswahlverfahren

(Arbeitstitel)

Steffen Kanitz

Mitglied der Geschäftsführung, Bundesgesellschaft für Endlagerung

mbH (BGE)

12:45 D

Verleihung der Ehrenmitgliedschaft der KTG

| Award of the Honorary Membership of KTG

Präsentiert von Frank Apel

Vorsitzender der KTG

13:00-14:00 Lunch

Wirtschaft | Economy

10:15 D/E

Restbetrieb und Rückbau in Nord- und

Süddeutschland | Dismantling and Last Years of

Operation in Northern and Southern Germany

Dr. Guido Knott

CEO, PreussenElektra GmbH

10:35 Pause

Kompetenz | Competence

11:05 D/E

Kerntechnische Ausbildung – Ein Grund zur Sorge?

| Nuclear Education – a Cause of Concern?

Prof. Dr. Jörg Starflinger

Geschäftsführender Direktor, Institut für Kernenergetik und

Energiesysteme (IKE), Universität Stuttgart

Technisch-wissenschaftliches Programm

14:00

j Themenblock Kompetenz & Innovation

j Themenblock Sicherheit & Betrieb

j Themenblock Rückbau & Abfallbehandlung

j Themenblock Zwischen- und Endlagerung

j Young Scientists‘ Workshop

15:30 Pause

16:00-ca.17:30

Fortsetzung Programm

18:30- 23:00

KernD-Empfang und Gesellschaftsabend

in der Ausstellung

KERNTECHNIK 2020

Programme Overview

atw Vol. 65 (2020) | Issue 1 ı January

KERNTECHNIK 2020

Programmstruktur nach Sessions

Themenblock

Kompetenz &

Innovation

j CFD Simulations

for Reactor Safety

Relevant Objectives

j Know-how, New Build

and Innovation

j Reactor Physics,

Thermo and Fluid

Dynamics

j Young Scientists'

Workshop

j CAMPUS Kerntechnik

Themenblock

Sicherheit &

Betrieb

j Radiation Protection

j What is an Accident

Tolerant Fuel?

j Operation and

Safety of Nuclear

Installations, Fuel

Themenblock

Rückbau &

Abfallbehandlung

j Experiences on

Post-Operation and

Decommissioning

j Decommissioning of

Nuclear Installations

Themenblock

Zwischen- &

Endlagerung

j N.N.

j Radioactive Waste Management, Storage

and Disposal

Programmstruktur nach Tagen

Montag

, Gremiensitzungen KernD

, Gremiensitzungen KTG

, Get-together KTG

Dienstag

, Industrieausstellung

, Plenarvorträge

, Themenblock

Kompetenz & Innovation

, Themenblock

Sicherheit & Betrieb

, Themenblock

Rückbau & Abfallbehandlung

, Themenblock

Zwischen- & Endlagerung

, Young Scientists‘ Workshop

, Gesellschaftsabend

Mittwoch

, Industrieausstellung

, Themenblock

Kompetenz & Innovation

, Themenblock

Sicherheit & Betrieb

, Themenblock

Rückbau & Abfallbehandlung

, Themenblock

Zwischen- & Endlagerung

, Young Scientists‘ Workshop

mit Preisverleihung

, CAMPUS Kerntechnik

#51KT

www.kerntechnik.com

KERNTECHNIK 2020

Programme Overview

atw Vol. 65 (2020) | Issue 1 ı January

Inside

Nachwuchstagung Kerntechnik 2019 in Essen-Kupferdreh

Die jährliche Nachwuchstagung Kerntechnik der Jungen

Generation der KTG fand in diesem Jahr beim Simulatorzentrum

KSG | GfS in Essen-Kupferdreh statt. 35 Junge

Nachwuchswissenschaftler, Studenten und interessierte

Mitarbeiter von Unternehmen aus der Kerntechnik hatten

die Möglichkeit, einen Blick „über den Tellerrand“ zu

erhalten.

KTG INSIDE

In elf Vorträgen spannte sich das breite Themenfeld von

der Kraftwerksimulation, der Reaktorsicherheit, dem

Strahlenschutz bis hin zu Innovationen in der Kerntechnik,

Entsorgungsthemen, dem Knowledge-Transfer am CERN

und der Arbeit anderer Jungen Generationen der Kerntechnischen

Gesellschaften Europas. Im Rahmen eines Impulsvortrags

und anschließender Diskussion wurden die

Themen Diversität und moderne Teamarbeit im Berufsalltag

besprochen.

Wie in jedem Jahr freuen wir uns insbesondere auf die

spannenden Besichtigungen, die oft einzigartige Highlights

darstellen. So wurde uns das weltweit einzigartige

Reaktor-Glasmodell mitsamt unterschiedlichen Szenarien

vorgeführt, Störfälle im Kraftwerks-Simulator des KKW

Brokdorf geprobt und die Behälter-Fertigungsstätte der

GNS besichtigt. Beim Vorabendtreffen und gemeinsamen

Dinner mit Speis und Trank gab es reichlich Gelegenheiten

für Austausch und Netzwerken.

Unser herzlichster Dank gilt dem Simulatorzentrum

KSG | GfS sowie der GNS für ihre Vorträge, Führungen

und Gastfreundschaft. Ebenso danken wir allen Referenten

für die spannenden Vorträge sowie den Teilnehmern für

das rege Interesse und die Teilnahme bei der Nachtagung

Kerntechnik 2019!

Vorstand der Jungen Generation in der KTG

Herzlichen Glückwunsch!

Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag

und wünscht ihnen weiterhin alles Gute!

Februar 2020

55 Jahre | 1965

18. Sven Lehmann, Adenbüttel

65 Jahre | 1955

1. Wolfgang Filbert

80 Jahre | 1940

9. Dr. Gerhard Preusche, Herzogenaurach

13. Dr. Hans-Ulrich Fabian, Gehrden

81 Jahre | 1939

8. Dr. Herbert Spierling, Dietzenbach

22. Dr. Manfred Schwarz, Dresden

86 Jahre | 1934

9. Dr. Horst Keese, Rodenbach

12. Dipl.-Ing. Horst Krause, Radebeul

91 Jahre | 1929

20. Dr. Helmut Hübel, Bensberg

Wenn Sie künftig eine

Erwähnung Ihres

Geburtstages in der

atw wünschen, teilen

Sie dies bitte der KTG-

Geschäftsstelle mit.

KTG Inside

Verantwortlich

für den Inhalt:

Die Autoren.

75 Jahre | 1945

1. Prof. Alfred Voß, Aidlingen

23. Dipl.-Ing. Victor Teschendorff, München

28. Dr. Günther Dietrich, Holzwickede

76 Jahre | 1944

26. Dr. Ivar Kalinowski, Ohrum

77 Jahre |1943

5. Dr. Joachim Banck, Heusenstamm

20. Ing. Leonhard Irion, Rückersdorf

28. Dr. Klaus Tägder, Sankt Augustin

83 Jahre | 1937

6. Dipl.-Ing. Heinrich Moers, Winter Park/

USA

11. Dr. Günter Keil, Sankt Augustin

18. Dipl.-Ing. Hans Wölfel, Heidelberg

84 Jahre | 1936

6. Dr. Ashu-Tosh Bhattacharyya, Erkelenz

17. Dr. Helfrid Lahr, Wedemark

Nachträgliche

Geburtstagsnennungen:

Dezember 2019

76 Jahre | 1943

7. Dipl.-Ing. Nobert Bauer, Limburgerhof

Januar 2020

77 Jahre | 1942

6. Dipl.-Ing. Günter Höfer, Mainhausen

Lektorat:

Natalija Cobanov,

Kerntechnische

Gesellschaft e. V.

(KTG)

Robert-Koch-Platz 4

10115 Berlin

T: +49 30 498555-50

F: +49 30 498555-51

E-Mail:

natalija.cobanov@

ktg.org

www.ktg.org

KTG Inside

atw Vol. 65 (2020) | Issue 1 ı January

STATISTICS

Nuclear Power Plants:

2019 atw Compact Statistics

Editorial

At the end of the last year 2019 (key date: 31 December 2019), nuclear power plants were operating in 31 countries

worldwide (cf. Table 1). In total, 449 nuclear power plants were operating on the key date. This means that the number

decreased by 2 units compared to the previous year’s number on 31 December 2018 (451, the highest number of units

since the first start of an commercial nuclear power plant in 1956), due to first criticalities on the one hand and

shut-downs on the other. The gross power output of these nuclear power plant units amounted to around 425 GWe*,

the net power output was approximately 401 GWe. This means that the available gross capacity and the available net

capacity did not significantly changed compared with the previous year's numbers. The highest capacity since the first

grid connection of a commercial nuclear power plant was available in 2019 (425,959 MWe gross, 401,177 MWe net).

Four (4) nuclear power plants started (nuclear) operation 1

in two countries in 2018. These units reached initial

criticality (C), were synchronized with the grid (G) and

started commercial operation (O) for the first time in

2019 (cf. Table 1): China: Taishan 2 (1750 MW, PWR),

Yangjiang 6 (1086 MW, PWR); Korea, Rep.: Shin Kori 4

(PWR, 1400 MW); Russia: Novovoronezh 2-2 (1200 MW,

VVER-PWR).

No unit resumed operation in 2019 in Japan after the

long-term shut-down of all reactors and safety evaluations

after the Fukushima accidents in 2011. In total 51 reactors

were in operation and shut-down in 2011, 9 resumed operation

until today.

Six (6) nuclear power plant units were definitively

per manently shut-down worldwide in five (5) countries in

2019. In Germany the Philippsburg 2 (1468 MW, PWR)

unit was shut-down due to the revised Atomic Act (2011)

and the termination of the license for power production.

In Japan the Genkai 2 (559 MW, PWR) plant ceased

operation. In Russia the LWGR-type unit Bilibinsk 1

(12 MW, LWGR) was shut-down. The plant supplied the

local area with electricity and heat. Three further units are

still in operation and will be shut-down in the coming

years. The barge Akademic Lomonosov with two nuclear

reactors will supply the region in the future. In Taiwan,

China, the Chin Shan 2 (636 MW, BWR) plant and in the

USA the Pilgrim 1 (712 MW, BWR) and Three Mile Island

1(1021 MW, PWR) reactor were shut down.

Five new projects (the same number as in the previous

year 2018) started with an official announcement and first

preparations for construction or the first concrete and

further build activities. In China three additional new

build projects started with Changjiang 3 (1170 MW, PWR),

Changjiang 4 (1170 MW, PWR), and Zhangzhou 1 (1212

MW, PWR), the Islamic Republic of Iran started the new

build of the second unit at Bushehr (1127 MW, VVER-

PWR) and in Russia one additional project started with the

Kursk II-2 project (1255 MW, VVER-PWR). At the Kursk

site four RMBK reactors are in operation which should be

replaced by modern GEN III+ PWR technology units.

In total 54 reactors are under construction worldwide

in 18 countries. The total gross capacity of this projects is

about 58 GW*, the net capacity 55 GW, in other words the

number was higher (1 unit) compared to the previous year

number due to the four (4) operation starts and five (5)

new build projects. Compared with the millennium change

1999/2000 this means that the number of projects under

construction has risen, when 30 nuclear power plants were

under construction worldwide.

Active construction projects (numbers in brackets)

listed are: Argentina (1), Bangladesh (2), Belarus (2),

Brazil (1), China (12), Finland (1), France (1), India (7),

Iran (1), Japan (2), Republic of Korea (4), Pakistan (2),

Russia (6), Slovak Republic (2), Taiwan (2), Turkey (1),

the USA (2), the United Arab Emirates (4) and the United

Kingdom (1).

In addition, there are about 200 nuclear power plant

units in 25 countries worldwide that are in an advanced

planning stage, others are in the pre-planning phase

( status: 31 December 2019).

Country Location/

Station name

Argentina

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Atucha 1 p D2O-PWR 357 341 1974

Embalse p Candu 648 600 1983

Atucha 2 p D2O-PWR 745 692 2014

CAREM25 P PWR 29 25 (2022)

Armenia

Metsamor 2 p VVER-PWR 408 376 1980

Belarus

Belarusian 1 P VVER-PWR 1 194 1 109 (2020)

Belarusian 2 P VVER-PWR 1 194 1 109 (2021)

Bangladesh

Rooppur 1 P VVER-PWR 1 200 1 080 (2023)

Rooppur 1 P VVER-PWR 1 200 1 080 (2024)

Belgium

Doel 1 p PWR 454 433 1975

Doel 2 p PWR 454 433 1975

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Doel 3 p PWR 1 056 1 006 1982

Doel 4 p PWR 1 090 1 039 1985

Tihange 1 p PWR 1 009 962 1975

Tihange 2 p PWR 1 055 1 008 1983

Tihange 3 p PWR 1 094 1 046 1985

Brazil

Angra 1 p PWR 640 609 1984

Angra 2 p PWR 1 350 1 275 1999

Angra 3 P PWR 1 300 1 245 (2021)

Bulgaria

Kozloduj 5 p VVER-PWR 1 000 953 1987

Kozloduj 6 p VVER-PWR 1 000 953 1989

Canada

Bruce 1 p Candu 824 772 1977

Bruce 2 p Candu 786 734 1977

Bruce 3 p Candu 805 730 1977

Statistics

Nuclear Power Plants: 2019 atw Compact Statistics

atw Vol. 65 (2020) | Issue 1 ı January

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Bruce 4 p Candu 805 750 1979

Bruce 5 p Candu 872 817 1985

Bruce 6 p Candu 891 822 1984

Bruce 7 p Candu 872 817 1986

Bruce 8 p Candu 845 817 1987

Darlington 1 p Candu 934 878 1993

Darlington 2 p Candu 934 878 1990

Darlington 3 p Candu 934 878 1993

Darlington 4 p Candu 934 878 1993

Pickering 1 p Candu 542 515 1971

Pickering 4 p Candu 542 515 1973

Pickering 5 p Candu 540 516 1983

Pickering 6 p Candu 540 516 1984

Pickering 7 p Candu 540 516 1985

Pickering 8 p Candu 540 516 1986

Point Lepreau p Candu 705 660 1983

China

CEFR p SNR 25 20 2011

Changjiang 1 p PWR 650 610 2015

Changjiang 2 p PWR 650 601 2016

Fangchenggang 1 p PWR 1 080 1 000 2015

Fangchenggang 2 p PWR 1 088 1 000 2016

Fangjiashan 1 p PWR 1 080 1 000 2014

Fangjiashan 2 p PWR 1 080 1 000 2014

Fuqing 1 p PWR 1 087 1 000 2014

Fuqing 2 p PWR 1 087 1 000 2015

Fuqing 3 p PWR 1 089 1 000 2016

Fuqing 4 p PWR 1 089 1 089 2017

Guandong 1 p PWR 984 944 1993

Guandong 2 p PWR 984 944 1994

Haiyang 1 p PWR 1 180 1 100 2018

Haiyang 2 p PWR 1 180 1 100 2018

Hongyanhe 1 p PWR 1 080 1 000 2013

Hongyanhe 2 p PWR 1 080 1 000 2013

Hongyanhe 3 p PWR 1 080 1 000 2014

Hongyanhe 4 p PWR 1 119 1 000 2016

Lingao 1 p PWR 990 938 2002

Lingao 2 p PWR 990 938 2002

Lingao II-1 p PWR 1 087 1 000 2010

Lingao II-2 p PWR 1 087 1 000 2011

Ningde 1 p PWR 1 087 1 000 2012

Ningde 2 p PWR 1 080 1 000 2014

Ningde 3 p PWR 1 080 1 000 2015

Ningde 4 p PWR 1 089 1 018 2016

Qinshan 1 p PWR 310 288 1992

Qinshan II-1 p PWR 650 610 2002

Qinshan II-2 p PWR 650 610 2004

Qinshan II-3 p PWR 642 610 2010

Qinshan II-4 p PWR 642 610 2011

Qinshan III-1 p Candu 728 665 2002

Qinshan III-2 p Candu 728 665 2003

Sanmen 1 p PWR 1 180 1 100 2018

Sanmen 2 p PWR 1 180 1 100 2018

Taishan 1 p PWR 1 750 1 660 2018

Taishan 2 [1] p PWR 1 750 1 660 2019

Tianwan 1 p VVER-PWR 1 060 990 2005

Tianwan 2 p VVER-PWR 1 060 990 2007

Tianwan 3 p VVER-PWR 1 126 1 060 2017

Tianwan 4 p VVER-PWR 1 126 1 060 2018

Yangjiang 1 p PWR 1 080 1 000 2013

Yangjiang 2 p PWR 1 080 1 000 2015

Yangjiang 3 p PWR 1 080 1 000 2015

Yangjiang 4 p PWR 1 086 1 000 2016

Yangjiang 5 p PWR 1 080 1 000 2018

Yangjiang 6 [1] p PWR 1 080 1 000 2019

Changjiang 3 [2] P PWR 1 170 1 090 (2024)

Changjiang 4 [2] P PWR 1 170 1 090 (2025)

Fangchenggang 3 P PWR 1 080 1 000 (2020)

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Fangchenggang 4 P PWR 1 080 1 000 (2022)

Fuqing 5 P PWR 1 087 1 000 (2020)

Fuqing 6 P PWR 1 087 1 000 (2020)

Hongyanhe 5 P PWR 1 080 1 000 (2020)

Hongyanhe 6 P PWR 1 080 1 000 (2021)

Shidaowan 1 P HTGR 211 200 (2020)

Tianwan 5 P VVER-PWR 1 118 1 000 (2020)

Tianwan 6 P VVER-PWR 1 118 1 000 (2022)

Zhangzhou 4 [2] P PWR 1 212 1 126 (2024)

Czech Republic

Dukovany 1 p VVER-PWR 500 473 1985

Dukovany 2 p VVER-PWR 500 473 1986

Dukovany 3 p VVER-PWR 500 473 1987

Dukovany 4 p VVER-PWR 500 473 1987

Temelín 1 p VVER-PWR 1 077 1 027 1999

Temelín 2 p VVER-PWR 1 056 1 006 2002

Finland

Loviisa 1 p VVER-PWR 520 496 1977

Loviisa 2 p VVER-PWR 520 496 1981

Olkiluoto 1 p BWR 890 860 1979

Olkiluoto 2 p BWR 890 860 1982

Olkiluoto 3 P PWR 1 600 1 510 (2020)

France

Belleville 1 p PWR 1 363 1 310 1987

Belleville 2 p PWR 1 363 1 310 1988

Blayais 1 p PWR 951 910 1981

Blayais 2 p PWR 951 910 1982

Blayais 3 p PWR 951 910 1983

Blayais 4 p PWR 951 910 1983

Bugey 2 p PWR 945 910 1978

Bugey 3 p PWR 945 910 1978

Bugey 4 p PWR 917 880 1979

Bugey 5 p PWR 917 880 1979

Cattenom 1 p PWR 1 362 1 300 1986

Cattenom 2 p PWR 1 362 1 300 1987

Cattenom 3 p PWR 1 362 1 300 1990

Cattenom 4 p PWR 1 362 1 300 1991

Chinon B-1 p PWR 954 905 1982

Chinon B-2 p PWR 954 905 1983

Chinon B-3 p PWR 954 905 1986

Chinon B-4 p PWR 954 905 1987

Chooz B-1 p PWR 1 560 1 500 1996

Chooz B-2 p PWR 1 560 1 500 1997

Civaux 1 p PWR 1 561 1 495 1997

Civaux 2 p PWR 1 561 1 495 1999

Cruas Meysse 1 p PWR 956 915 1983

Cruas Meysse 2 p PWR 956 915 1984

Cruas Meysse 3 p PWR 956 915 1984

Cruas Meysse 4 p PWR 956 915 1984

Dampierre 1 p PWR 937 890 1980

Dampierre 2 p PWR 937 890 1980

Dampierre 3 p PWR 937 890 1981

Dampierre 4 p PWR 937 890 1981

Fessenheim 1 p PWR 920 880 1977

Fessenheim 2 p PWR 920 880 1977

Flamanville 1 p PWR 1 382 1 330 1985

Flamanville 2 p PWR 1 382 1 330 1986

Golfech 1 p PWR 1 363 1 310 1990

Golfech 2 p PWR 1 363 1 310 1993

Gravelines B-1 p PWR 951 910 1980

Gravelines B-2 p PWR 951 910 1980

Gravelines B-3 p PWR 951 910 1980

Gravelines B-4 p PWR 951 910 1981

Gravelines C-5 p PWR 951 910 1984

Gravelines C-6 p PWR 951 910 1985

Nogent 1 p PWR 1 363 1 310 1987

Nogent 2 p PWR 1 363 1 310 1988

Paluel 1 p PWR 1 382 1 330 1984

STATISTICS

Statistics

Nuclear Power Plants: 2019 atw Compact Statistics

atw Vol. 65 (2020) | Issue 1 ı January

STATISTICS

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Paluel 2 p PWR 1 382 1 330 1984

Paluel 3 p PWR 1 382 1 330 1985

Paluel 4 p PWR 1 382 1 330 1986

Penly 1 p PWR 1 382 1 330 1990

Penly 2 p PWR 1 382 1 330 1992

St. Alban 1 p PWR 1 381 1 335 1986

St. Alban 2 p PWR 1 381 1 335 1987

St. Laurent B-1 p PWR 956 915 1981

St. Laurent B-2 p PWR 956 915 1981

Tricastin 1 p PWR 955 915 1980

Tricastin 2 p PWR 955 915 1980

Tricastin 3 p PWR 955 915 1980

Tricastin 4 p PWR 955 915 1981

Flamanville 3 P PWR 1 600 1 510 (2021)

Germany

Brokdorf p PWR 1 480 1 410 1986

Emsland p PWR 1 406 1 335 1988

Grohnde p PWR 1 430 1 360 1985

Gundremmingen C p BWR 1 344 1 288 1985

Isar 2 p PWR 1 485 1 410 1988

Neckarwestheim II p PWR 1 400 1 310 1989

Philippsburg 2 [6] j PWR 1 468 1 402 1985

Hungary

Paks 1 p VVER-PWR 500 470 1983

Paks 2 p VVER-PWR 500 473 1984

Paks 3 p VVER-PWR 500 473 1986

Paks 4 p VVER-PWR 500 473 1987

India

Kaiga 1 p Candu (IND) 220 202 2001

Kaiga 2 p Candu (IND) 220 202 1999

Kaiga 3 p Candu (IND) 220 202 2007

Kaiga 4 p Candu (IND) 220 202 2010

Kakrapar 1 p Candu (IND) 220 202 1993

Kakrapar 2 p Candu (IND) 220 202 1995

Kudankulam 1 p VVER-PWR 1 000 917 2013

Kudankulam 2 p VVER-PWR 1 000 917 2016

Madras Kalpakkam 1 p Candu (IND) 220 205 1984

Madras Kalpakkam 2 p Candu (IND) 220 205 1986

Narora 1 p Candu (IND) 220 202 1992

Narora 2 p Candu (IND) 220 202 1991

Rajasthan 1 p Candu 100 90 1973

Rajasthan 2 p Candu 200 187 1981

Rajasthan 3 p Candu (IND) 220 202 1999

Rajasthan 4 p Candu (IND) 220 202 2000

Rajasthan 5 p Candu (IND) 220 202 2009

Rajasthan 6 p Candu (IND) 220 202 2010

Tarapur 1 p BWR 160 150 1969

Tarapur 2 p BWR 160 150 1969

Tarapur 3 p Candu (IND) 540 490 2006

Tarapur 4 p Candu (IND) 540 490 2005

Kakrapar 3 P Candu (IND) 700 640 (2021)

Kakrapar 4 P Candu (IND) 700 640 (2020)

PFBR (Kalpakkam) P SNR 500 470 (2020)

Kudankulam 3 P VVER-PWR 1 000 917 (2023)

Kudankulam 4 P VVER-PWR 1 000 917 (2023)

Rajasthan 7 P Candu (IND) 700 630 (2020)

Rajasthan 8 P Candu (IND) 700 630 (2021)

Iran

Bushehr 1 p VVER-PWR 1 000 953 2011

Bushehr 2 [2] P VVER-PWR 1 127 1 057 (2025)

Japan

Fukushima Daini 1 p BWR 1 100 1 067 1982

Fukushima Daini 2 p BWR 1 100 1 067 1984

Fukushima Daini 3 p BWR 1 100 1 067 1985

Fukushima Daini 4 p BWR 1 100 1 067 1987

Genkai 3 p PWR 1 180 1 127 1994

Genkai 4 p PWR 1 180 1 127 1997

Hamaoka 3 p BWR 1 100 1 056 1987

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Hamaoka 4 p BWR 1 137 1 092 1993

Hamaoka 5 p BWR 1 267 1 216 2004

Higashidori 1 p BWR 1 100 1 067 2005

Ikata 3 p PWR 890 846 1994

Kashiwazaki Kariwa 1 p BWR 1 100 1 067 1985

Kashiwazaki Kariwa 2 p BWR 1 100 1 067 1990

Kashiwazaki Kariwa 3 p BWR 1 100 1 067 1993

Kashiwazaki Kariwa 4 p BWR 1 100 1 067 1994

Kashiwazaki Kariwa 5 p BWR 1 100 1 067 1990

Kashiwazaki Kariwa 6 p BWR 1 356 1 315 1996

Kashiwazaki Kariwa 7 p BWR 1 356 1 315 1997

Mihama 3 p PWR 826 781 1976

Ohi 3 p PWR 1 180 1 127 1991

Ohi 4 p PWR 1 180 1 127 1993

Onagawa 1 p BWR 524 496 1984

Onagawa 2 p BWR 825 796 1995

Onagawa 3 p BWR 825 798 2002

Sendai 1 p PWR 890 846 1984

Sendai 2 p PWR 890 846 1985

Shika 1 p BWR 540 505 1993

Shika 2 p BWR 1 358 1 304 2005

Shimane 2 p BWR 820 791 1989

Takahama 1 p PWR 826 780 1974

Takahama 2 p PWR 826 780 1975

Takahama 3 p PWR 870 830 1985

Takahama 4 p PWR 870 830 1985

Tokai 2 p BWR 1 100 1 067 1978

Tomari 1 p PWR 579 550 1989

Tomari 2 p PWR 579 550 1991

Tomari 3 p PWR 912 866 2009

Tsuruga 2 p PWR 1 160 1 115 1986

Shimane 3 P BWR 1 375 1 325 (2022)

Ohma P BWR 1 385 1 325 (2023)

Genkai 2 [6] j PWR 559 529 1981

Korea (Republic)

Kori 2 p PWR 676 639 1983

Kori 3 p PWR 1 042 1 003 1985

Kori 4 p PWR 1 041 1 001 1986

Shin Kori 1 p PWR 1 048 996 2010

Shin Kori 2 p PWR 1 045 993 2011

Shin Kori 3 p PWR 1 400 1 340 2016

Shin Kori 4 [1] p PWR 1 400 1 340 2019

Hanul 1 p PWR 1 003 960 1988

Hanul 2 p PWR 1 008 962 1989

Hanul 3 p PWR 1 050 994 1998

Hanul 4 p PWR 1 053 998 1998

Hanul 5 p PWR 1 051 996 2003

Hanul 6 p PWR 1 051 996 2004

Wolsong 1 p Candu 687 645 1983

Wolsong 2 p Candu 678 653 1997

Wolsong 3 p Candu 698 675 1999

Wolsong 4 p Candu 703 679 1999

Shin Wolsong 1 p PWR 1 043 991 2012

Shin Wolsong 2 p PWR 1 000 960 2015

Hanbit 1 p PWR 996 953 1986

Hanbit 2 p PWR 993 945 1987

Hanbit 3 p PWR 1 050 997 1995

Hanbit 4 p PWR 1 049 997 1996

Hanbit 5 p PWR 1 053 997 2001

Hanbit 6 p PWR 1 052 995 2002

Shin Kori 5 P PWR 1 400 1 340 (2022)

Shin Kori 6 P PWR 1 400 1 340 (2024)

Shin Hanul 1 P PWR 1 400 1 340 (2020)

Shin Hanul 2 P PWR 1 400 1 340 (2022)

Kori 2 j PWR 676 639 1983

Mexico

Laguna Verde 1 p BWR 820 765 1990

Laguna Verde 2 p BWR 820 765 1995

Statistics

Nuclear Power Plants: 2019 atw Compact Statistics

atw Vol. 65 (2020) | Issue 1 ı January

Country Location/

Station name

Netherlands

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Borssele p PWR 515 482 1973

Pakistan

Kanupp 1 p Candu 137 909 1972

Chasnupp 1 p PWR 325 300 2000

Chasnupp 2 p PWR 325 300 2011

Chasnupp 3 p PWR 340 315 2016

Chasnupp 4 p PWR 340 315 2017

Kanupp 2 P PWR 1 100 1 014 (2021)

Kanupp 3 P PWR 1 100 1 014 (2022)

Romania

Cernavoda 1 p Candu 706 650 1996

Cernavoda 2 p Candu 706 655 2007

Russia

Balakovo 1 p VVER-PWR 1 000 953 1986

Balakovo 2 p VVER-PWR 1 000 953 1988

Balakovo 3 p VVER-PWR 1 000 953 1990

Balakovo 4 p VVER-PWR 1 000 953 1993

Beloyarsky 3 p FBR 600 560 1981

Beloyarsky 4 p FBR 800 750 2014

Bilibino 2 p LWGR 12 11 1975

Bilibino 3 p LWGR 12 11 1976

Bilibino 4 p LWGR 12 11 1977

Kalinin 1 p VVER-PWR 1 000 953 1985

Kalinin 2 p VVER-PWR 1 000 953 1987

Kalinin 3 p VVER-PWR 1 000 953 2004

Kalinin 4 p VVER-PWR 1 000 953 2011

Kola 1 p VVER-PWR 440 411 1973

Kola 2 p VVER-PWR 440 411 1975

Kola 3 p VVER-PWR 440 411 1982

Kola 4 p VVER-PWR 440 411 1984

Kursk 1 p LWGR 1 000 925 1977

Kursk 2 p LWGR 1 000 925 1979

Kursk 3 p LWGR 1 000 925 1984

Kursk 4 p LWGR 1 000 925 1986

Leningrad 2 p LWGR 1 000 925 1976

Leningrad 3 p LWGR 1 000 925 1980

Leningrad 4 p LWGR 1 000 925 1981

Leningrad II-1 p VVER-PWR 1 187 1 085 2018

Novovoronezh 4 p VVER-PWR 417 385 1973

Novovoronezh 5 p VVER-PWR 1 000 953 1981

Novovoronezh II-1 p VVER-PWR 1 000 955 2016

Novovoronezh II-2 [1] p VVER-PWR 1 000 955 2019

Rostov 1 p VVER-PWR 1 000 953 2001

Rostov 2 p VVER-PWR 1 000 953 2010

Rostov 3 p VVER-PWR 1 000 950 2014

Rostov 4 p VVER-PWR 1 030 980 2017

Smolensk 1 p LWGR 1 000 925 1983

Smolensk 2 p LWGR 1 000 925 1985

Smolensk 3 p LWGR 1 000 925 1990

Akademik Lomonosov I P PWR 40 35 (2020)

Baltic 1 (Kaliningrad) P VVER-PWR 1 170 1 080 (2024)

Kursk II-1 P VVER-PWR 1 255 1 175 (2024)

Kursk II-2 [2] P VVER-PWR 1 255 1 175 (2025)

Leningrad II-2 P VVER-PWR 1 170 1 085 (2021)

Bilibino 1 [6] j LWGR 12 11 1974

Slovakia

Bohunice 3 p VVER-PWR 505 472 1985

Bohunice 4 p VVER-PWR 505 472 1985

Mochovce 1 p VVER-PWR 470 436 1998

Mochovce 2 p VVER-PWR 470 436 1999

Mochovce 3 P VVER-PWR 440 408 (2020)

Mochovce 4 P VVER-PWR 440 408 (2020)

Slovenia

Krsko p PWR 727 696 1983

South Africa

Koeberg 1 p PWR 970 930 1984

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Koeberg 2 p PWR 970 930 1985

Spain

Almaraz 1 p PWR 1 049 1 011 1981

Almaraz 2 p PWR 1 044 1 006 1983

Ascó 1 p PWR 1 033 995 1984

Ascó 2 p PWR 1 027 997 1985

Cofrentes p BWR 1 092 1 064 1985

Trillo 1 p PWR 1 066 1 002 1988

Vandellos 2 p PWR 1 087 1 045 1987

Sweden

Forsmark 1 p BWR 1 022 984 1980

Forsmark 2 p BWR 1 158 1 120 1981

Forsmark 3 p BWR 1 212 1 170 1985

Oskarshamn 3 p BWR 1 450 1 400 1985

Ringhals 1 p BWR 910 878 1976

Ringhals 2 p PWR 847 807 1975

Ringhals 3 p PWR 1 117 1 064 1981

Ringhals 4 p PWR 990 940 1983

Switzerland

Beznau 1 p PWR 380 365 1969

Beznau 2 p PWR 380 365 1972

Gösgen p PWR 1 060 1 010 1979

Leibstadt p BWR 1 275 1 220 1984

Mühleberg p BWR 390 373 1973

Taiwan, China

Kuosheng 1 p BWR 985 948 1981

Kuosheng 2 p BWR 985 948 1983

Maanshan 1 p PWR 951 890 1984

Maanshan 2 p PWR 951 890 1985

Lungmen 1 P BWR 1 356 1 315 (2021)

Lungmen 2 P BWR 1 356 1 315 (2022)

Chin Shan 2 [6] j BWR 636 604 1979

Turkey

Akkuyu 1 P VVER-PWR 1 200 1 114 (2023)

United Arab Emirates

Barakah 1 P PWR 1 400 1 340 (2020)

Barakah 2 P PWR 1 400 1 340 (2021)

Barakah 3 P PWR 1 400 1 340 (2022)

Barakah 4 P PWR 1 400 1 340 (2023)

United Kingdom

Dungeness B-1 p AGR 615 520 1985

Dungeness B-2 p AGR 615 520 1986

Hartlepool-1 p AGR 655 595 1984

Hartlepool-2 p AGR 655 585 1985

Heysham I-1 p AGR 625 585 1984

Heysham I-2 p AGR 625 575 1985

Heysham II-1 p AGR 682 595 1988

Heysham II-2 p AGR 682 595 1989

Hinkley Point B-1 p AGR 655 610 1976

Hinkley Point B-2 p AGR 655 610 1977

Hunterston B-1 p AGR 644 460 1976

Hunterston B-2 p AGR 644 430 1977

Sizewell B p PWR 1 250 1 191 1995

Torness Point 1 p AGR 682 595 1988

Torness Point 2 p AGR 682 595 1989

Hinkley Point C-1 P PWR 1 720 1 630 (2025)

Ukraine

Khmelnitski 1 p VVER-PWR 1 000 950 1985

Khmelnitski 2 p VVER-PWR 1 000 950 2004

Rovno 1 p VVER-PWR 402 363 1981

Rovno 2 p VVER-PWR 416 377 1982

Rovno 3 p VVER-PWR 1 000 950 1987

Rovno 4 p VVER-PWR 1 000 950 2004

Zaporozhe 1 p VVER-PWR 1 000 950 1985

Zaporozhe 2 p VVER-PWR 1 000 950 1985

Zaporozhe 3 p VVER-PWR 1 000 950 1987

Zaporozhe 4 p VVER-PWR 1 000 950 1988

Zaporozhe 5 p VVER-PWR 1 000 950 1988

STATISTICS

Statistics

Nuclear Power Plants: 2019 atw Compact Statistics

atw Vol. 65 (2020) | Issue 1 ı January

STATISTICS

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Zaporozhe 6 p VVER-PWR 1 000 950 1989

South Ukraine 1 p VVER-PWR 1 000 950 1983

South Ukraine 2 p VVER-PWR 1 000 950 1985

South Ukraine 3 p VVER-PWR 1 000 950 1989

USA

Arkansas Nuclear One 1 p PWR 969 903 1974

Arkansas Nuclear One 2 p PWR 1 006 943 1980

Beaver Valley 1 p PWR 955 923 1976

Beaver Valley 2 p PWR 957 923 1987

Braidwood 1 p PWR 1 289 1 225 1988

Braidwood 2 p PWR 1 289 1 225 1988

Browns Ferry 1 p BWR 1 200 1 152 1974

Browns Ferry 2 p BWR 1 193 1 152 1975

Browns Ferry 3 p BWR 1 232 1 190 1977

Brunswick 1 p BWR 1 074 1 002 1977

Brunswick 2 p BWR 1 075 1 002 1975

Byron 1 p PWR 1 307 1 225 1985

Byron 2 p PWR 1 304 1 225 1987

Callaway p PWR 1 316 1 236 1985

Calvert Cliffs 1 p PWR 935 918 1975

Calvert Cliffs 2 p PWR 939 911 1977

Catawba 1 p PWR 1 286 1 205 1985

Catawba 2 p PWR 1 286 1 205 1986

Clinton 1 p BWR 1 175 1 138 1987

Comanche Peak 1 p PWR 1 283 1 215 1990

Comanche Peak 2 p PWR 1 283 1 215 1993

Donald Cook 1 p PWR 1 266 1 152 1975

Donald Cook 2 p PWR 1 210 1 133 1978

Columbia (WNP 2) p BWR 1 244 1 200 1984

Cooper p BWR 844 801 1974

Davis Besse 1 p PWR 971 925 1978

Diablo Canyon 1 p PWR 1 236 1 159 1985

Diablo Canyon 2 p PWR 1 246 1 164 1985

Dresden 2 p BWR 1 057 1 009 1970

Dresden 3 p BWR 1 057 1 009 1971

Duane Arnold p BWR 737 680 1975

Farley 1 p PWR 933 888 1977

Farley 2 p PWR 934 888 1981

Fermi 2 p BWR 1 317 1 217 1988

FitzPatrick p BWR 918 882 1975

Ginna p PWR 713 614 1970

Grand Gulf 1 p BWR 1 516 1 440 1985

Hatch 1 p BWR 891 857 1974

Hatch 2 p BWR 905 865 1979

Hope Creek 1 p BWR 1 360 1 291 1986

Indian Point 2 p PWR 1 348 1 299 1974

Indian Point 3 p PWR 1 051 1 012 1976

La Salle 1 p BWR 1 242 1 170 1984

La Salle 2 p BWR 1 238 1 170 1984

Limerick 1 p BWR 1 203 1 139 1986

Limerick 2 p BWR 1 199 1 139 1990

McGuire 1 p PWR 1 358 1 220 1981

McGuire 2 p PWR 1 358 1 220 1984

Millstone 2 p PWR 946 91 0 1975

Millstone 3 p PWR 1 308 1 253 1986

Monticello p BWR 734 685 1971

Nine Mile Point 1 p BWR 671 642 1969

Nine Mile Point 2 p BWR 1 302 1 259 1988

North Anna 1 p PWR 1 035 980 1978

North Anna 2 p PWR 1 033 980 1980

Oconee 1 p PWR 955 887 1973

Oconee 2 p PWR 955 887 1974

Oconee 3 p PWR 961 893 1974

Country Location/

Station name

Status Reactor

type

Capacity

gross

[MW]

Capacity

net

[MW]

1 st

Criticality

[Year]

Palisades p PWR 870 812 1971

Palo Verde 1 p PWR 1 528 1 403 1986

Palo Verde 2 p PWR 1 524 1 403 1988

Palo Verde 3 p PWR 1 524 1 403 1986

Peach Bottom 2 p BWR 1 233 1 160 1974

Peach Bottom 3 p BWR 1 233 1 160 1974

Perry 1 p BWR 1 397 1 312 1987

Point Beach 1 p PWR 696 643 1970

Point Beach 2 p PWR 696 643 1972

Prairie Island 1 p PWR 642 593 1973

Prairie Island 2 p PWR 641 593 1974

Quad Cities 1 p BWR 1 061 1 009 1973

Quad Cities 2 p BWR 1 061 1 009 1973

RiverBend 1 p BWR 1 073 1 036 1986

Robinson 2 p PWR 855 769 1971

Salem 1 p PWR 1 276 1 170 1977

Salem 2 p PWR 1 303 1 170 1981

Seabrook 1 p PWR 1 330 1 242 1990

Sequoyah 1 p PWR 1 259 1 221 1981

Sequoyah 2 p PWR 1 279 1 221 1982

Shearon Harris 1 p PWR 983 951 1987

South Texas 1 p PWR 1 410 1 354 1988

South Texas 2 p PWR 1 410 1 354 1989

St. Lucie 1 p PWR 1 122 1 080 1976

St. Lucie 2 p PWR 1 135 1 080 1983

Virgil C. Summer p PWR 1 071 1 030 1984

Surry 1 p PWR 900 848 1972

Surry 2 p PWR 900 848 1973

Susquehanna 1 p BWR 1 374 1 298 1983

Susquehanna 2 p BWR 1 374 1 298 1985

Turkey Point 3 p PWR 885 835 1972

Turkey Point 4 p PWR 885 835 1973

Vogtle 1 p PWR 1 223 1 160 1987

Vogtle 2 p PWR 1 226 1 160 1989

Waterford 3 p PWR 1 250 1 200 1985

Watts Bar 1 p PWR 1 370 1 270 1996

Watts Bar 2 p PWR 1 240 1 180 2016

Wolf Creek p PWR 1 351 1 268 1984

Vogtle 3 P PWR 1 080 1 000 (2021)

Vogtle 4 P PWR 1 080 1 000 (2022)

Pilgrim [6] j BWR 712 670 1972

Three Mile Island 1 [6] j PWR 1 021 976 1974

1) Start of nuclear operation (first criticality: C, first grid connection: G, commercial

operation: O): 4 units in 3 countries in 2019: China: Taishan 2 (1750 MW, PWR,

CGO), Yangjiang 6 (1086 MW, PWR, CGO); Korea: Shin Kori 4 (1400 MW, PWR,

CGO); Russia: Novovoronezh 2-2 (1200 MW, PWR, CGO).

2) Start of construction (first concrete or official announcement and first preparations

for construction), 5 units 3 countries in 2019: China: Changjiang 3 (1170 MW,

PWR), Changjiang 4 (1170 MW, PWR), Zhangzhou 1 (1212 MW, PWR); Iran:

Bushehr 2 (1127 MW, VVER-PWR); Russia: Kursk 2-2 (1255 MW, VVER-PWR).

3) Project under construction (finally) cancelled: none.

4) Resumed operation: none.

5) Nuclear power plant taken in long-term shutdown: none.

6) Nuclear power plants permanently shutdown: 6 units in 5 countries in 2019: Germany:

Philippsburg 2 (1468 MW, PWR); Japan: Genkai 2 (559 MW, BWR); Russia:

Bilibinsk 1 (12 MW, LWGR); Taiwan: China, Chin Shan 2 (636 MW, BWR); USA: Pilgrim

1 (712 MW, BWR), Three Mile Island 1 (1021 MW, PWR).

(All capacity data in MWe gross)

AGR: Advanced Gas-cooled Reactor, BWR: Boiling water reactor, Candu: CANada

Deuterium Uranium reactor (IND: Indian type), D2O-PWR: heavy water moderated,

pressurised water reactor, PWR: pressurised water reactor, GGR: gas-graphite

reactor, LWGR/GLWR: light water cooled graphite moderated reactor (Russian type

RBMK), FBWR: advanced boiling water reactor, FBR: fast breeder reactor

| Tab. 1.

Nuclear power plant units worldwide on 31.12.2019 in operation (p), under construction (P), in lay-up operation/long-term shutdown (s) or permanently shut-down in 2019 (j)

[Sources: Operators, IAEO]. All information and data refer to the year 2019. Data have been updated with reference to the sources

Statistics

Nuclear Power Plants: 2019 atw Compact Statistics

atw Vol. 65 (2020) | Issue 1 ı January

Zum Zum Tode Tode von von

Prof. Dr. Dr. Adolf Birkhofer

Am 9. November 2019 ist Prof. Dr. Dr. h. c. mult. Adolf Birkhofer im Alter von 85 Jahren in

München verstorben. Mit ihm ging eine in vielfacher Weise beeindruckende und hochgeschätzte

Persönlichkeit – als international anerkannter Wissenschaftler, engagierter Hochschullehrer sowie

als Mitbegründer und langjähriger technisch-wissenschaftlicher Geschäftsführer der Gesellschaft

für Anlagen- und Reaktorsicherheit (GRS).

Geboren in München, studierte Adolf

Birkhofer zunächst Elektrotechnik an

der damaligen Technischen Hochschule

München (THM) und später

Theoretische Physik an der Universität Innsbruck. Nach

beruflichen Stationen bei Siemens & Halske sowie beim

Technischen Überwachungs-Verein Bayern folgten 1963

seine Promotion in Innsbruck und 1967 seine Habilitation

an der THM. Bereits 1963 war er zum Institut für Messund

Regeltechnik der THM gewechselt. Mit dem Laboratorium

für Reaktorregelung und Anlagensicherung (LRA)

baute er dort eine der beiden Vorläuferorganisationen der

GRS auf. Im Jahr 1971 übernahm er die Leitung des LRA

und wurde zum außerordentlichen Professor für Reaktordynamik

und Reaktorsicherheit berufen. Mitte der 1970er

Jahre setzte er sich dafür ein, das LRA mit dem Institut für

Reaktorsicherheit in Köln zur GRS zusammenzuschließen.

Von 1977 bis Ende 2001 leitete er die fachliche Arbeit der

GRS und sorgte dafür, dass sie sich schnell auch über die

Grenzen Deutschlands hinaus höchstes Ansehen erarbeitet

hat.

Die große technische Expertise Adolf Birkhofers wird

nicht nur an seiner Autorschaft an rund 200 wissenschaftlichen

Veröffentlichungen deutlich. Noch wesentlicher

ist, dass er ganz maßgeblich die Entwicklung von

grundlegenden Methoden und Konzepten vorangetrieben

hat, die das moderne Verständnis von Reaktorsicherheit

geformt, bis heute Gültigkeit und eine führende Rolle

Deutschlands auf dem Gebiet der Reaktorsicherheit

begründet haben. Dazu zählt beispielsweise die

Erarbeitung der „Deutschen Risikostudie Kernkraftwerke“

(Phasen A und B) durch die GRS. Bis heute zitiert, legen

diese Studien eine wesentliche Grundlage für die Entwicklung

der Probabilistische Sicherheitsanalyse. Prägend

bis heute für die GRS war seine Forderung, dass das

Verhindern von Störfällen stets Priorität gegenüber der

Begrenzung ihrer Auswirkungen haben muss.

Als international hochgeschätzter Fachmann wurde

Adolf Birkhofer in zahlreiche Fachgremien und Kommissionen

berufen. In Deutschland gehörte er als Berater des

Bundesumweltministeriums über drei Jahrzehnte der

Reaktor-Sicherheitskommission (RSK) an, davon viele

Jahre als deren Vorsitzender. Die Entwicklung des

deutschen nuklearen Sicherheitskonzeptes hat er dabei

mit Unterstützung durch die RSK ganz wesentlich

gestaltet. Als Mitglied und zeitweiliger Vorsitzender der

„International Nuclear Safety Group“ (INSAG) der

Internationalen Atomenergie-Organisation wirkte er an

der Erarbeitung des 1996 publizierten Reports „Defence in

Depth in Nuclear Safety“ (INSAG-10) mit, der bis heute als

Standard für das gleichnamige Sicherheitskonzept gilt.

Auf internationaler Ebene war er nicht nur in der IAEO

aktiv, sondern saß beispielsweise auch über mehrere Jahre

dem „Committee on the Safety of Nuclear Installations“

(CSNI) der OECD vor.

Eine besonders enge berufliche und persönliche

Beziehung verband ihn mit Frankreich. So ist es seinem

Einsatz zu verdanken, dass die GRS und ihr damaliges

französisches Pendant, das Institut de Protection et de

Sûreté Nucleaire (IPSN; heute IRSN) im Jahr 1989 eine

Vereinbarung über eine weitreichende Zusammenarbeit

schlossen. Diese bis heute andauernde Partnerschaft

bildete die Grundlage für die Gründung des gemeinsamen

Tochterunternehmens RISKAUDIT und nicht zuletzt

der EUROSAFE Initiative, die kürzlich, 20 Jahre nach

ihrer Gründung, im Europäischen Netzwerk Technischer

Sicherheitsorganisationen (ETSON) aufgegangen ist.

Nach dem Reaktorunfall von Tschernobyl hat er sich

mit Nachdruck dafür eingesetzt, durch eine enge

und partnerschaftliche Zusammenarbeit mit den damaligen

Institutionen die Sicherheit von Kernkraftwerken

russischer Bauart in den Staaten des Ostblocks zu erhöhen.

Die daraus erwachsenen vertrauensvollen Beziehungen

der GRS zu den dortigen Aufsichtsbehörden und Fachorganisationen

haben auch heute noch Bestand.

Für sein großes fachliches Engagement wurden Adolf

Birkhofer zahlreiche Ehrungen zuteil, darunter das Große

Bundesverdienstkreuz, der Bayerische Maximiliansorden

für Wissenschaft und Kunst und die Auszeichnung als

Ritter der französischen Ehrenlegion. Die Universität

Karlsruhe und das Kurtschatow-Institut in Moskau

verliehen ihm die Ehrendoktorwürde.

Wie sehr ihm Forschung und Lehre am Herzen lagen,

zeigte sich zuletzt auch darin, dass er nach seiner

Emeritierung im Jahr 2003 das Institute for Safety and

Reliability (ISaR) an der TU München gründete und dort

auch als Geschäftsführer wirkte.

Adolf Birkhofer hat uns Vieles hinterlassen, für das wir

dankbar sein müssen – nicht nur mit seinen Verdiensten

um die Erhöhung der Reaktorsicherheit und die GRS

als führendes technisch-wissenschaftliches Kompetenzzentrum

auf diesem Gebiet, die uns Ansporn und

Verpflichtung sind, sondern auch mit den Erinnerungen

an eine engagierte, freundliche, weltoffene und humorvolle

Persönlichkeit. Wir werden ihm stets ein ehrendes

Andenken bewahren.

Autoren

Uwe Stoll

Technisch-wissenschaftlicher Geschäftsführer der GRS

Hans Steinhauer

Kaufmännisch-juristischer Geschäftsführer der GRS

N A C H R U F

Nachruf

atw Vol. 65 (2020) | Issue 1 ı January

NEWS

Top

IAEA's Grossi at COP 25:

More nuclear power needed

for clean energy transition

(iaea) IAEA Director General Rafael

Mariano Grossi, speaking at the

United Nations Climate Change

Conference (COP 25) in Madrid,

December 2019, said greater use of

low-carbon nuclear power is needed

to ensure the global transition to clean

energy, including to back up variable

renewables such as solar and wind.

The world is currently well off the

mark from reaching the climate goals

of the Paris Agreement. With around

two-thirds of the world’s electricity

still generated through burning fossil

fuels, and despite growing investment

in renewable energy sources, global

emissions of greenhouse gases

reached a record high last year.

Mr Grossi said greater deployment

of a diverse mix of low-carbon sources

such as hydro, wind and solar, as well

as nuclear power and battery storage,

will be needed to reverse that trend

and set the world on track to meet

climate goals.

“We should not see nuclear energy

and renewables as being in competition

with one another,” he said in

Madrid at a side event on Sustainable

Development Goal 7 (SDG 7) – to

ensure access to affordable and reliable

energy. “We need to make use of

all available sources of clean energy.”

Nuclear power plants produce

virtually no greenhouse gas emissions

or air pollutants during their operation.

They are also able to operate

around the clock at near full capacity,

while variable renewables require back

up power during their output gaps.

“Nuclear power offers a steady,

reliable supply of electricity,”

Mr Grossi stated. “It can provide

continuous, low-carbon power to back

up increasing use of renewables. It can

be the key that unlocks their potential

by providing flexible support – day or

night, rain or shine.”

He also spoke of the role of nuclear

applications that help countries adapt

to the consequences of climate change

which are already apparent. “Our

scientists help countries to develop

new varieties of rice and barley that

are tolerant of drought, extreme temperatures

and salinity,” he said. “We

support the use of nuclear techniques

to identify and manage limited water

resources.”

The UN side event, entitled “Accelerating

the energy transformation in

support of sustainable development

and the Paris Agreement”, focused on

initiatives that could have a significant

impact toward achieving SDG 7 goals,

helping to close the energy access gap

in a sustainable way and promoting

climate action by transitioning toward

zero-carbon energy solutions.

The event was opened by remarks

from Liu Zhenmin, Under-Secretary-

General of the United Nations Department

of Economic and Social Affairs

(UN DESA), Damilola Ogunbiyi, Chief

Executive Officer of Sustainable

Energy for All and Li Yong, Director

General of the United Nations Industrial

Organization (UNIDO).

Mr Grossi said nuclear power needs

a place at the table where the world’s

energy future is decided, and that he

was encouraged by his talks with other

international organizations and their

willingness to work with the IAEA towards

a cleaner climate.

He underscored the symbolism of

coming to COP 25 just one week after

taking office.

“This reflects the importance of the

issue and my firm belief that nuclear

science and technology have an

important role to play in helping the

world to address the climate emergency,”

he said. “That view is shared

by many of the IAEA’s 171 Member

States.”

| (193471203); www.iaea.org

Europe

FORATOM welcomes Commission’s

Green Deal ambitions

(foratom) FORATOM welcomes the

European Commission’s goal of

becoming more ambitious in reducing

its CO 2 emissions whilst at the same

time ensuring that no EU citizen is left

behind in the transition.

If the EU is to achieve its zero- carbon

target 2050, then its current 2030 CO 2

reduction targets may not be enough.

We therefore support the Commission’s

goal of raising this target, so long as it

leaves Member States free to choose

their own low-carbon energy mix.

Expecting them to reduce their GHG

emissions, whilst at the same time

preventing them from investing in specific

low-carbon technologies such as

nuclear, would be counter-productive.

As indicated by Fatih Birol upon the

publication of the 2019 edition of the

IEA’s World Energy Outlook “There is

no single or simple way to transform

global energy systems. Many technologies

& fuels have a part to play across

all sectors of the economy.”

FORATOM furthermore supports

the goal of designing and implementing

a strong industrial strategy. Not

only is nuclear key in providing the

baseload electricity which other

industries depend on at a reasonable

cost, it is also an important European

industry in itself.

“The European nuclear industry

currently sustains more than 1.1 million

jobs in the EU and generates more

than half a trillion euros in GDP”

states FORATOM Director General

Yves Desbazeille. “This is important

when we bear in mind the potential

impact of the energy transition on

citizens. For example, those currently

employed in the coal industry could

be retrained in order to fill the skills

gap in the nuclear industry”.

Both the IPCC (Global Warming

of 1.5°C) and the IEA (Nuclear Power

in a Clean Energy System) have made

it very clear that decarbonisation

goals cannot be achieved without

nuclear energy. The European

Commission (A Clean Planet for all)

has confirmed that nuclear will

form the backbone of a carbon-free

European power system, together

with renewables.

| (193471322); www.foratom.org

Myrrha: First contract

signed for Belgian

nuclear research facility

(nccnet/sck) A € 7.6 m contract has

been signed for the design of buildings

and utilities for the first phase of the

Myrrha nuclear research facility, the

Belgian Nuclear Research Centre

SCK•CEN has confirmed.

SCK•CEN said the contract, the

first for the project, includes water and

electricity for the facility. It was signed

with Belgian company Tractebel

and Spanish company Empresarios

Agrupados.

Twelve months ago the Belgian

government announced financing of

€558m towards the development of

the Myrrha facility.

SCK•CEN said the funding would

be used for construction of the first

phase of the facility at SCK•CEN’s

premises in Mol.

The facility, scheduled to begin

operation in 2026, will produce radioisotopes

and promote fundamental

and applied research on materials,

SCK•CEN said. Myrrha will contribute

to producing new radioisotopes and

to developing less invasive therapies to

fight against cancer.

The facility will also be used to develop

solutions for managing nuclear

waste and for researching methods of

deep geological disposal.

The Myrrha project, supported by

the European Union, is to design

and build a multifunctional research

installation.

Myrrha will be the first prototype

of a nuclear reactor driven by a

News

atw Vol. 65 (2020) | Issue 1 ı January

Operating Results September 2019

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto BWR FI 910 880 720 658 684 5 766 073 267 421 281 100.00 96.63 99.97 95.71 99.44 95.67

OL2 Olkiluoto BWR FI 910 880 720 656 944 5 426 464 257 323 006 100.00 90.88 99.94 90.35 99.18 90.04

KCB Borssele PWR NL 512 484 720 356 232 5 138 418 166 860 106 98.88 83.81 98.87 83.73 96.62 80.39

KKB 1 Beznau 7) PWR CH 380 365 720 271 481 2 127 323 129 461 433 100.00 86.29 100.00 86.10 99.23 85.34

KKB 2 Beznau 1,2,7) PWR CH 380 365 255 91 519 2 103 198 136 453 605 35.42 85.19 33.87 85.00 32.89 84.35

KKG Gösgen 7) PWR CH 1060 1010 720 755 968 5 894 880 319 770 408 100.00 85.88 99.99 85.30 99.05 84.89

KKM Mühleberg BWR CH 390 373 720 272 160 2 494 730 129 899 045 100.00 100.00 99.66 99.74 96.92 97.65

CNT-I Trillo PWR ES 1066 1003 720 758 289 6 149 267 253 440 935 100.00 89.13 99.94 88.72 98.00 87.46

Dukovany B1 1) PWR CZ 500 473 0 0 2 662 535 114 892 028 0 83.33 0 83.07 0 81.29

Dukovany B2 2) PWR CZ 500 473 227 105 461 1 715 933 109 950 104 31.53 54.07 29.49 53.43 29.30 52.39

Dukovany B3 PWR CZ 500 473 720 350 232 2 672 518 109 170 559 100.00 83.90 99.65 83.50 97.29 81.59

Dukovany B4 PWR CZ 500 473 720 356 514 3 244 818 109 688 087 100.00 99.83 100.00 99.66 99.03 99.06

Temelin B1 PWR CZ 1080 1030 720 775 772 5 496 200 119 857 242 100.00 78.48 99.96 78.21 99.58 77.54

Temelin B2 PWR CZ 1080 1030 720 781 223 5 795 810 115 068 327 100.00 81.59 99.99 81.33 100.28 81.77

Doel 1 2) PWR BE 454 433 720 327 462 2 250 940 137 695 402 100.00 74.44 99.96 74.15 97.58 74.08

Doel 2 PWR BE 454 433 643 287 390 2 533 531 136 335 470 89.33 86.32 88.89 84.86 87.50 84.80

Doel 3 PWR BE 1056 1006 720 759 431 5 598 794 260 731 278 100.00 80.96 99.52 80.29 99.49 80.43

Doel 4 PWR BE 1084 1033 720 725 939 6 850 498 267 223 908 100.00 100.00 92.96 96.21 91.25 94.91

Tihange 1 PWR BE 1009 962 720 707 386 6 553 065 305 383 923 100.00 100.00 99.95 99.99 97.37 99.25

Tihange 2 PWR BE 1055 1008 720 738 405 2 154 428 256 806 358 100.00 32.92 99.99 32.25 98.08 31.40

Tihange 3 PWR BE 1089 1038 720 763 539 6 946 670 278 173 943 100.00 99.97 99.99 99.24 97.85 97.86

NEWS

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Since Month Year Month Year Month Year

commissioning

KBR Brokdorf DWR 1480 1410 720 959 402 7 391 648 357 959 458 100.00 85.33 94.27 80.10 89.60 75.93

KKE Emsland DWR 1406 1335 720 989 749 7 717 806 354 536 775 100.00 85.56 100.00 85.45 97.76 83.78

KWG Grohnde DWR 1430 1360 720 968 134 7 676 462 385 250 676 100.00 86.70 99.95 86.40 93.42 81.41

KRB C Gundremmingen SWR 1344 1288 720 960 604 7 420 665 338 362 419 100.00 85.49 100.00 84.87 98.78 83.84

KKI-2 Isar DWR 1485 1410 720 1 014 325 8 834 840 362 560 653 100.00 94.58 100.00 94.22 94.41 90.44

GKN-II Neckarwestheim 1,2) DWR 1400 1310 197 265 700 7 347 710 337 174 544 27.39 92.02 26.39 82.86 26.41 80.22

KKP-2 Philippsburg DWR 1468 1402 720 961 047 7 837 362 373 998 517 100.00 86.20 100.00 85.95 89.44 80.24

particle accelerator. The system

consists of a proton accelerator that

delivers a beam to a spallation target,

which in turn couples to a subcritical

lead-bismuth cooled fast reactor.

| (193471323); www.sckcen.be,

www.myrrha.be/

Reactors

China: Construction begins

of two Hualong One reactors

at Changjiang

(nucnet) China has begun construction

of two new reactor units at the

Changjiang nuclear station in the

island province of Hainan off the

country’s southeast coast, state media

reported on Monday following a

signing ceremony.

The China Nuclear Energy Association

also confirmed the news, saying

each unit will take around 60 months

to complete.

The company in charge, the

Huaneng Nuclear Development

Corporation, has chosen China’s indigenous

Generation III HPR1000

reactor technology, also known as the

Hualong One, for the two units, the

official China News Service reported.

Other press reports said the total

investment for the two new units

will reach 39.45 billion yuan

($5.64bn), and they are scheduled to

begin commercial operation in 2025

and 2026.

There are already two units in commercial

operation at Changjiang. The

new units will be Changjiang-3 and -4.

Changjiang-1 and -2 are both

CNP600 units developed by China

National Nuclear Corporation and

have a net capacity of 601 MW. They

began commercial operation in 2015

and 2016.

According to the International

Atomic Energy Agency, China has

10 commercial nuclear units under

construction, not including the new

Changjiang units, and 48 in operation.

China has ambitious nuclear plans

with an official target of 58 GW of

installed nuclear capacity by 2020,

up from around 36 GW today.

According to Shanghai-based

energy consultancy Nicobar, China’s

goal is to have 110 nuclear units in

commercial operation by 2030, but

this target is likely to be adjusted.

A recent forecast by the China

Electricity Council said the country

will fall short of its nuclear power

generation capacity target for 2020.

| (193471344); en.cnnc.com.cn/

Hungary: Licence documentation

for Paks 2 reactors ‘will

be submitted next summer’

(mvm/nucnet) Documentation for the

licence application for two new

reactors units at the Paks nuclear

station in Hungary is on schedule to be

ready in spring 2020 in time to be

submitted to the regulator for approval

in the summer, Rosatom chief executive

officer Alexey Likhachev said.

Paks 2, the company overseeing

the project, told NucNet that earlier

this week János Süli, the minister

Net-based values

(Czech and Swiss

nuclear power

plants gross-based)

Refueling

Inspection

Repair

Stretch-out-operation

Stretch-in-operation

Hereof traction supply

Incl. steam supply

New nominal

capacity since

January 2016

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

News

atw Vol. 65 (2020) | Issue 1 ı January

NEWS

responsible for the planning, construction

and commissioning of the

Paks II project, reviewed progress

with Mr Likhachev.

Mr Likhachev said more than 400

licences are needed for the two units

and the next major milestone will

be completion of all the design documentation,

which will be submitted to

the regulator, the Hungarian Atomic

Energy Authority.

Mr Süli said that without the new

Paks units, Hungary would not be

able to reach its climate goals. He said

Paks 2 will avoid 17 million tonnes of

carbon dioxide emissions a year –

compared to total emissions of the

transport sector of 12 million tonnes

of C0 2 a year.

An agreement signed in 2014 will

see Russia supply two VVER-1200

pressurised water reactors for Paks 2,

and a loan of up to €10bn to finance

80% of the €12bn project.

| (193471332); www.mvm.hu

Iran: Bushehr – Start of

construction for second

nuclear plant confirmed

(nucnet) Iran has officially started

construction of a second Russiasupplied

nuclear power plant at

the Bushehr nuclear station on the

Persian Gulf coast.

The Atomic Energy Organisation of

Iran said on Sunday a ceremony had

been held to mark the pouring if first

concrete for the Bushehr-2 VVER-1000

plant.

Ali Akbar Salehi, head of the

AEOI, and deputy chief of Russia’s

state nuclear corporation Rosatom,

Alexander Lokshin, launched construction

at the ceremony where

concrete was poured for the reactor

base.

The AEOI also confirmed it had

“long-term” plans to build a third

Russian plant at the site, about 750 km

south of Teheran.

Iran and Russia signed an agreement

to build two new units at

Bushehr in November 2014. This was

followed in June 2019 by the signing

of a final contract for construction.

In 2014 the country’s official

Islamic Republic News Agency (IRNA)

said Russia and Iran had signed an

agreement for the construction of up

to eight new nuclear reactor units in

the Middle Eastern republic.

According to the International

Atomic Energy Agency, Bushehr-1,

Iran’s first commercial nuclear plant,

officially began commercial operation

in September 2013. It had begun operating

at full capacity in 2012 and now

supplies about 2 % of the country’s

electricity.

A 2015 nuclear deal Iran signed

with six major powers, including

Russia, placed restrictions on the sort

of nuclear technology Tehran could

develop and its production of nuclear

fuel, but it did not require Iran to halt

its use of nuclear energy for power

generation.

“In a long-term vision until 2027-

2028, when these projects are

finished, we will have 3,000 megawatts

of nuclear plant-generated

electricity,” Mr Salehi said at the

ceremony.

The Islamic republic has been

seeking to reduce its reliance on oil

and gas through the development of

nuclear power.

As part of the 2015 agreement,

Moscow provides Tehran with the fuel

it needs for its electricity-generating

nuclear reactors.

Bushehr is fuelled by uranium

produced in Russia and is monitored

by the International Atomic Energy

Agency.

| (193471334)

Science & Research

Managing ageing research

reactors to ensure safe,

effective operations

(iaea) As over two thirds of the world’s

operating research reactors are now

over 30 years old, operators and regulators

are focusing on refurbishing

and modernizing reactors to ensure

they can continue to perform in a safe

and efficient manner. This is also one

of many topics have been discussed

from 25 to 29 November at the IAEA's

International Conference on Research

Reactors: Addressing Challenges and

Opportunities to Ensure Effectiveness

and Sustainability in Buenos Aires,

Argentina.

“The lifetime of research reactors is

normally determined by the need for

their use and their conformance with

up-to-date safety requirements, since

most of their systems and components

can be replaced, refurbished or modernized

without major difficulty,” said

Amgad Shokr, Head of the IAEA’s

Research Reactor Safety Section.

“ Refurbishment and modernization

should not be limited to just systems

and components; operators should also

review safety procedures against IAEA

safety standards to prevent the interruption

of research reactor ervices.”

For more than 60 years, research

reactors have been centres of innovation

and development for nuclear

science and technology programmes

around the world. These small nuclear

reactors primarily generate neutrons

– rather than power – for research,

education and training purposes, as

well as for applications in areas such

as industry, medicine and agriculture.

There are two kinds of ageing

related to research reactors: physical

ageing, which is the degradation

of the physical condition of the

reactor’s systems and components,

and obsolescence, which is when

the technology used for computers,

instrumentation and control systems

or safety regulations becomes outdated.

The ageing of facilities was one of

the concerns that led to the IAEA

initiating its Research Reactor Safety

Enhancement Plan in 2001. This plan

aims to help countries ensure a high

level of research reactor safety. It

includes the Code of Conduct on the

Safety of Research Reactors, which

provides guidance to countries on

the development and harmonization

of policies, laws and regulations regarding

the safety of research reactors.

As part of this plan, countries work

with the IAEA to implement systematic

ageing management programmes

that, among others, use

good practices to minimize the performance

degradation of systems and

components, to continuously monitor

and assess reactor performance and to

implement practical safety upgrades.

These ageing programmes can also

benefit from operating programmes in

other areas, such as maintenance,

periodic testing, inspections and

periodic safety reviews.

“While the number of operating

research reactors is decreasing, the

average age is increasing,” said Ram

Sharma, a nuclear engineer on

research reactor operation and maintenance

at the IAEA. “So, it is of

paramount importance to establish,

implement and continuously improve

plans for management, refurbishment

and modernization to ensure costeffective

operation and utilization to

get the most out of existing research

reactors. IAEA support, such as peer

review missions, can play a key part in

achieving that goal.”

Comprehensive support

Countries can draw on a range of IAEA

support to address ageing at their

research reactors. This includes

assistance with developing safety

standards and optimizing reactor

availability, as well as adopting

recommended practices based on

IAEA-published collections on safety

and using information disseminated

by the IAEA on developing and implementing

modernization and refurbishment

projects. This assistance

News

atw Vol. 65 (2020) | Issue 1 ı January

extends to new research reactor programmes

and to assessing plans to

proactively address ageing throughout

all phases of the research reactor’s

lifetime, from the design and selection

of materials to the construction and

operation of the facilities.

Review missions are initiated upon

the request of a country and are

supported by the IAEA and teams of

international experts who carry out

assessments and provide recommendations

for further improvements. In

November 2017, the first ageing management

peer review mission for a

research reactor was completed at

Belgian Reactor 2 (BR2), which is one

of three operating research reactors at

the Belgian Nuclear Research Centre

(SCK•CEN). The mission was based

on the methodology of Safety Aspects

of Long Term Operation (SALTO) missions

for nuclear power plants and

adapted to suit research reactors.

“The mission identified a number

of items that were overlooked, such as

ageing management of radioisotope

production facilities and experimental

devices,” said Frank Joppen, a

nuclear safety engineer at SCK•CEN.

“As a result, the classification systems

of components are being updated, and

feedback from maintenance, inspection

and surveillance is being used to

further improve ageing management

programmes.”

In operation since 1963, BR2 is one

of the oldest research reactors in

Western Europe. It produces around

one quarter of the global supply

of radioisotopes for medical and

industrial purposes, including for

cancer therapy and medical imaging.

It also produces a type of silicon

that is used as a semiconductor

material in electronic components.

BR2 is now permitted to operate

until its next periodic safety review

in 2026, when a decision on extending

its operation for another ten years

may be taken.

“The ageing management programme

of BR2 will be further developed,

which means taking into

account the remarks made during the

IAEA mission,” Joppen said. “The

efficiency of the programme will be

reviewed and will be the subject of the

next safety review.”

The next IAEA ageing managem

ent missions for research reactors

have been requested by the Netherlands

and Uzbekistan and are planned

for 2020. “The BR2 mission showed

that the SALTO methodology could

be effectively applied to research

reactors. We will continue to improve

the efficiency and effectiveness of this

mission, as well as other services, to

Uranium

Uranium prize range: Spot market [USD*/lb(US) U

Prize range: Spot market [USD*/lb(US) U 3O 8]

3O 8]

) 1

140.00

) 1

120.00

100.00

80.00

60.00

40.00

Yearly average prices in real USD, base: US prices (1982 to1984) *

20.00

0.00

Year

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

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019

* Actual nominal USD prices, not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019

| Uranium spot market prices from 1980 to 2019 and from 2008 to 2019. The price range is shown.

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

Separative work: Spot market price range [USD*/kg UTA]

Conversion: Spot conversion price range [USD*/kgU]

180.00

22.00

) 1 20.00

160.00

) 1

18.00

140.00

16.00

120.00

14.00

100.00

12.00

10.00

80.00

8.00

60.00

6.00

40.00

4.00

20.00

2.00

0.00

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

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019

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

Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2019

1980

Jan. 2008

Jan. 2009

1985

Jan. 2010

1990

Jan. 2011

Jan. 2012

maximize the benefits from research

reactors,” Shokr said.

| (193471308); www.iaea.org

Market data

(All information is supplied without

guarantee.)

Nuclear Fuel Supply

Market Data

Information in current (nominal)

U.S.-$. No inflation adjustment of

prices on a base year. Separative work

data for the formerly “secondary

market”. Uranium prices [US-$/lb

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

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

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

(Separative work unit)].

2017

p Uranium: 19.25–26.50

p Conversion: 4.50–6.75

p Separative work: 39.00–50.00

2018

p Uranium: 21.75–29.20

p Conversion: 6.00–14.50

p Separative work: 34.00–42.00

2019

January 2019

p Uranium: 28.70–29.10

p Conversion: 13.50–14.50

p Separative work: 41.00–44.00

February 2019

p Uranium: 27.50–29.25

1995

Jan. 2013

Jan. 2014

2000

Jan. 2015

2005

Jan. 2016

Jan. 2017

2010

Jan. 2018

2015

Jan. 2019

2019

Jan. 2020

| Separative work and conversion market price ranges from 2008 to 2019. The price range is shown.

In December 2009 Energy Intelligence changed the method of calculation for spot market prices. The change results in virtual price leaps.

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

Sources: Energy Intelligence, Nukem; Bilder/Figures: atw 2019

Jan. 2008

Jan. 2009

Jan. 2010

Jan. 2011

Jan. 2012

p Conversion: 13.50–14.50

p Separative work: 42.00–45.00

March 2019

p Uranium: 24.85–28.25

p Conversion: 13.50–14.50

p Separative work: 43.00–46.00

April 2019

p Uranium: 25.50–25.88

p Conversion: 15.00–17.00

p Separative work: 44.00–46.00

May 2019

p Uranium: 23.90–25.25

p Conversion: 17.00–18.00

p Separative work: 46.00–48.00

June 2019

p Uranium: 24.30–25.00

p Conversion: 17.00–18.00

p Separative work: 47.00–49.00

July 2019

p Uranium: 24.50–25.60

p Conversion: 18.00–19.00

p Separative work: 47.00–49.00

August 2019

p Uranium: 24.90–25.60

p Conversion: 19.00–20.00

p Separative work: 47.00–49.00

September 2019

p Uranium: 24.80–26.00

p Conversion: 20.00–21.00

p Separative work: 47.00–50.00

October 2019

p Uranium: 23.75–25.50

p Conversion: 21.00–22.00

p Separative work: 47.00–50.00

| Source: Energy Intelligence

www.energyintel.com

Jan. 2013

Jan. 2014

Jan. 2015

Jan. 2016

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020

NEWS

News

atw Vol. 65 (2020) | Issue 1 ı January

NUCLEAR TODAY

John Shepherd is a

freelance journalist

and communications

consultant.

Sources:

Green/EFA statement

https://bit.ly/

2sYvUoQ

Foratom comments

on taxonomy

https://bit.ly/

352Pf6A

Dr Angela Wilkinson

opinion

https://bit.ly/

2RvC6Pm

New Year Brings a Fresh Political Challenge

for a Champion of Climate Change

As you read this article, the Christmas decorations will have been packed away for another 12 months and some of us

will be wondering how we can keep to those new year resolutions we set for ourselves – perhaps rashly – in the dying

hours of 2019.

Resolutions are a great way to start the new year and even

though 2020 is already well under way, it’s not too late to

set goals for the year ahead. In fact, it’s imperative the

nuclear energy community resolves to take action in

defence of the industry in Europe in the face of a strident

push from its opponents.

The European Parliament elections of 2019 witnessed a

surge in support for groups including the Greens and that

new-found political muscle in Brussels is now being

harnessed to attack nuclear under the guise of environmental

concern.

As I sent this article off to the publisher, the Greens/

European Free Alliance grouping in the European

Parliament were celebrating a compromise on a proposed

framework to facilitate sustainable investment, known as

‘taxonomy’.

The taxonomy follows the European Commission’s

setting up in 2018 of a technical expert group on sustainable

finance. Tasks set for the group included helping the

Commission to develop an EU classification system

(dubbed the taxonomy) to identify environmentallysustainable

economic activities. This is in line with the

bloc’s goal of decarbonising its energy sector.

The EU said the guidelines released by the expert group

in June 2019 formed part of moves to ensure that the

financial sector “can play a critical role in transitioning to a

climate-neutral economy and in funding investments at

the scale required”.

According to the findings, nuclear does have the

credentials to help tackle climate change, but the guidance

questioned the technology’s suitability because of the

storage of nuclear waste.

Foratom, the voice of the European nuclear industry,

indicated that, in the case of nuclear, the focus on the

waste issue had been deliberately used to exclude nuclear

from the taxonomy. Foratom said the waste criteria did not

appear to have been applied in the same way for other

technologies and hoped future talks on the taxonomy “will

remain open and transparent, include real experts on the

various issues and focus on a fact-based, rather than an

ideological, debate”.

France has, sensibly, lobbied to keep nuclear in the

taxonomy. However, the Greens now say their compromise

will mean “any investment in coal cannot be considered

sustainable”. They say a “strengthened ‘no-harm’ test will

help avoid nuclear energy from being considered an

environmentally sustainable investment”.

According to the EU, the taxonomy is not set to be

implemented until the end of 2022, one year later than

initially proposed, which means battle lines have been

drawn in this latest twist in the fight for the future of

European energy policy.

Ironically, Germany, which had already joined fellow

EU states Austria and Luxembourg in opposing a role for

nuclear in the taxonomy, could yet save the day if nuclear’s

proponents box clever.

This is because natural gas could find itself excluded

from the taxonomy as being too emitting. This would hit

countries such as Germany, where power utilities and

others have invested in natural gas. Going forward, this

would make it harder to achieve emissions reductions. And

there is speculation that Germany could now have a vested

interest in seeing the taxonomy proposals scrapped for

that reason – using nuclear as a scapegoat.

Germany takes over the presidency of the EU’s Council

of Ministers for six months in July 2020, giving it a key role

in driving forward the Council’s work on EU legislation, so

the year ahead promises to be an interesting one!

But why is energy policy always driven by political

dogma rather than common sense?

According to the Paris-based International Energy

Agency, a range of technologies, including nuclear power,

will be needed for clean energy transitions around the

world. The IEA said “the key to making energy systems

clean is to turn the electricity sector from the largest

producer of CO 2 emissions into a low-carbon source that

reduces fossil fuel emissions in areas like transport, heating

and industry”.

And while the IEA said renewables are expected to

continue to lead, “nuclear power can also play an important

part along with fossil fuels using carbon capture, utilisation

and storage”.

Those who seek to remove nuclear from the array of

technologies the world needs to rely on would also do well

to heed the words of the newly-appointed secretary- general

and chief executive of the World Energy Council,

Dr Angela Wilkinson. She said recently: “Don’t let perfection

( ideology) become the enemy of the faster, deeper and

social affordability decarbonisation. There is no need to

reinvent the wheel – leverage technology and policy

innovation by encouraging countries to learn with and

from each other and increase the pace of learning by

doing.”

Wilkinson has also correctly pointed out that energy

transition is not a single issue and there is a need to “ manage

the connected challenges of energy security, energy equity

and affordability and environmental sustainability”.

Europe’s leaders should look across the Atlantic, to

where the heads of three provincial Canadian governments

have agreed to work together “to explore new, cutting-edge

technology in nuclear power generation to provide carbonfree,

affordable, reliable, and safe energy, while helping

unlock economic potential across Canada”.

The provinces of Ontario, New Brunswick and

Saskatchewan said in December 2019 that they were

committed to collaborating on the development and

deployment of “innovative, versatile and scalable” small

modular reactors in Canada.

Meanwhile, here in Europe, there is no time to lose in

preventing the Greens and their supporters in the

European Parliament from what can only be described as

an incredible act of self harm.

Nuclear is not a sacred cow to be protected at all costs.

But nuclear energy is, by any sensible measure, a key element

in the shield against climate change. Policies that

pander to political correctness over practical solutions to

tackling climate change deserve to be stopped in their

tracks.

Nuclear Today

New Year Brings a Fresh Political Challenge for a Champion of Climate Change ı John Shepherd

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