Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information.
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Energy Supply
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Winter 2022/23
is Coming
Dual-Use Act in Trialog
Nuclear Power Plants:
2019 atw Compact Statistics
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atw Vol. 65 (2020) | Issue 1 ı January
USA: 80 Years Actually
3
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
6
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.
G
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
7
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
8
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
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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.
11
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
12
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.
13
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
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
| 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]
17
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
25
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
(5
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
β
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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
(c) β=45°
| 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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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 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
R
C
(%)
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
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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RESEARCH AND INNOVATION 42
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
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[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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[3] Z.S. Xie. Physical Analysis of Nuclear Reactor [M]. Xi’an: Xi’an
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[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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[7] J. Jang, H. Arastoopour. CFD simulation of a pharmaceutical
bubbling bed drying process at three different scales [J].
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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
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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
45
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
46
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
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mit Preisverleihung
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Programme Overview
atw Vol. 65 (2020) | Issue 1 ı January
Inside
47
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
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F: +49 30 498555-51
E-Mail:
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ktg.org
www.ktg.org
KTG Inside
atw Vol. 65 (2020) | Issue 1 ı January
48
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
49
STATISTICS
Statistics
Nuclear Power Plants: 2019 atw Compact Statistics
atw Vol. 65 (2020) | Issue 1 ı January
50
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)
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
51
STATISTICS
Statistics
Nuclear Power Plants: 2019 atw Compact Statistics
atw Vol. 65 (2020) | Issue 1 ı January
52
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
53
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
54
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
55
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)
1)
Refueling
2)
Inspection
3)
Repair
4)
Stretch-out-operation
5)
Stretch-in-operation
6)
Hereof traction supply
7)
Incl. steam supply
8)
New nominal
capacity since
January 2016
9)
Data for the Leibstadt
(CH) NPP will
be published in a
further issue of atw
BWR: Boiling
Water Reactor
PWR: Pressurised
Water Reactor
Source: VGB
News
atw Vol. 65 (2020) | Issue 1 ı January
56
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
140.00
) 1
120.00
120.00
100.00
100.00
80.00
80.00
60.00
60.00
40.00
40.00
Yearly average prices in real USD, base: US prices (1982 to1984) *
20.00
20.00
0.00
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
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.
)1
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. 2008
Jan. 2009
Jan. 2009
Jan. 2010
Jan. 2010
Jan. 2011
Jan. 2011
Jan. 2012
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. 2013
Jan. 2014
Jan. 2014
Jan. 2015
Jan. 2015
Jan. 2016
Jan. 2016
Jan. 2017
Jan. 2017
Jan. 2018
Jan. 2018
Jan. 2019
Jan. 2019
Jan. 2020
Jan. 2020
57
NEWS
News
atw Vol. 65 (2020) | Issue 1 ı January
58
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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