atw - International Journal for Nuclear Power | 06/07.2020

nucmag.com
2020
6/7
ISSN · 1431-5254
23.55 €
Deep Geological Radioactive
and Chemical
Waste Disposal:
Where We Stand and
Where We Go
How Final Disposal Can Work
What has Happened
to the U.S. Nuclear Waste
Disposal Program?

Bewege mit uns
die Welt
Auf der Suche nach
neuen Lösungen?
Kommen Sie in unser Team!
BEWERBEN SIE SICH JETZT!
www.kraftanlagen.com/karriere
Aktuell suchen wir:
Stellvertretender Leiter Strahlenschutz (m/w/d)
Ingenieur für nukleare Abfallbehandlungsprozesse und -systeme (m/w/d)
Als Unternehmen von Bouygues Construction sind wir ein vielseitiger Partner für Industrie und Energiewirtschaft.
In ganz Europa entwickeln und bauen wir in der Kraftanlagen Gruppe modernste Anlagentechnik. Wir bieten
individuelle, zukunftsfähige Lösungen an und bedienen die gesamte Wertschöpfungskette.
Kraftanlagen Heidelberg bietet Dienstleistungen über den ganzen Lebenszyklus kerntechnischer Anlagen. Dazu
zählen der Nachbetrieb und Rückbau kerntechnischer Anlagen sowie die Planung und Lieferung von Anlagen
zur Konditionierung und Entsorgung von radioaktiven Abfällen. www.kraftanlagen.com/loesungen/nukleartechnik

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Cyber Security and Nuclear Power
Malware is as old as the development of technical systems themselves. While malware was still “physically tangible”
in “real” tangible systems, such as gears or engines, like “sugar in the gas tank” or “cutting in the gearbox”, it has
developed further with digital technology in the computer sector, rather invisibly, virtually, and very rapidly. Around
the almost all-encompassing topic of digitalization, malware is ultimately a digital code that influences computer
systems or digital systems themselves, directly or indirectly modifying them and mostly damaging. Only in the early
days was the creation of viruses a more or less sporting challenge for the sake of honour in the “Hall of Fame”.
The beginnings of malware go back to the year 1949.
Only 11 years earlier Konrad Zuse had completed his Zuse
Z1, a freely programmable mechanical calculator. Due to
manufacturing problems, however, it was never functional.
In 1941 Zuse succeeded in putting the first computer in the
world into operation with his Z3. The decisive factor was
the Z3’s ability to execute arbitrary algorithms, i.e. the
possibility of free programming was given. However, the
malware of 1949 was initially only a theoretical consideration
for self-reproducing computer programs. In the following
years, the subject of computer viruses accompanied
computer developments more or less theoretically.
In 1971 the first self-replicating experimental program,
the first virus, called “Creeper” was activated on a DEC
PDP-10 under the TENEX operating system. Creeper was
able to replicate itself on other DEC machines via the
ARPANET, a predecessor of today’s “Internet”, but did not
damage them, but indicated its infection with the words
“I’m the creeper, catch me if you can”. An answer to
Creeper was also given by the same programmer: Reaper
also moved independently in the ARPANET and destroyed
Creeper.
Today the number of computer viruses and other
malware is estimated to be about 900 million types, a considerable
number when compared to the approximately
5.3 billion Internet users worldwide, who also use more
than one end device on average. It is not possible to provide
an exact or even approximately reliable estimate of the
damage caused by malware, the cost of defending against
it, or the payment of “ransoms” for data encrypted by
malware, as the grey area of unreported incidents is said to
be extensive. A survey of Internet users carried out for
Germany by the industry association BITKOM in February
2020 shows a frightening picture for malware: 46 % of
the 1004 respondents have had experience with malware
in the previous 12 months, mostly via e-mails or via
appropriately prepared websites. Another statistic shows
that on average 13 out of 1000 e-mails are prepared
with malware. It is pleasing that the majority of the
respondents, 78 %, see the responsibility in terms of data
security primarily as being with the individual user
himself. Thus, with 85 % coverage for virus protection and
70 % for firewalls, the precautions taken are already
considerable. Unfortunately, the remaining proportion of
computers on the Internet is sufficient to pass on computer
virus infections. In addition to the risk of malware attacks
from private or professional use of computers, there is also
a risk of malware attacks for industrial digital systems.
One of the more publicly known malware is the
Stuxnet computer worm. This was developed specifically
to sabotage Iran’s nuclear program. Stuxnet was able
to manipulate programmable logic controllers (PLCs) for
uranium enrichment ultracentrifuges and to damage
them mechanically through faulty control systems. The
basis for the spread was the initial infection of a computer
with the Windows® operating system via a USB stick – the
developers assumed that the digital technology in focus
works as an isolated system – and the subsequent spread to
the PLCs in the local network. Stuxnet has thus made it
clear how sensitive it is to act even with isolated networks
or even singular systems. Any “contact” with the outside
world carries potential risks.
Due to the special and particularly high importance of
the topic of safety in the nuclear industry, the topic of cyber
security is also one of high priority and early measures are
taken. In principle, the safety-critical systems and safety
systems in nuclear facilities are “island facilities”. They
neither have a direct connection to the Internet nor are they
connected to other internal systems or networks in order to
exclude possible backdoors from the outset – they are
“ air-gapped” computers or networks that, if possible, even
have no hardware network interfaces in order to exclude
entry points for malware at this level. The hardwired
instrumentation and control and security control
technology still present in many nuclear facilities does not
have such vulnerabilities. This experience is currently
being used to drive forward projects for Small and Medium
Sized Reactors (SMR) based on Field Programable Gate
Arrays. This technology dispenses with software-based and
thus malware-critical micro processors.
Furthermore, the staggered security concept for
nuclear power plants, designed and implemented for a
large number of possible or postulated cases of impact on
plant and plant security, also guarantees the security and
protection of people and the environment in conceivable
cyber attack scenarios.
In addition, as mentioned at the beginning, the human
factor is an important factor in cyber protection. The
sensitive handling of all types of data carriers, i.e. potential
carriers of malware, precise instructions for handling
hardware and software, intensive and careful training and
constant sensitization to the topic of cyber security and
measures to avoid risks are just as much a part of this as
continuous programs for testing and optimizing the
robustness of all measures per se – both on the hardware
and software side and the soft skills of employees.
Cyber security is an issue for nuclear energy, but it is also
protected against cyber attacks by the many measures
taken.
Christopher Weßelmann
– Editor in Chief –
303
EDITORIAL
Editorial
Cyber Security and Nuclear Power

atw Vol. 65 (2020) | Issue 6/7 ı June/July
304
EDITORIAL
Cyber-Security in der Kernenergie
Malware ist so alt wie die Entwicklung von technischen Systemen an sich. War bei „echten“ greifbaren Systemen, wie
zum Beispiel Getrieben oder Motoren, die Malware noch „physisch greifbar“, so „Zucker im Benzintank“ oder „Späne
im Getriebe“, entwickelte sich diese mit der Digitaltechnik im Computersektor eher unsichtbar, virtuell weiter, und dies
sehr rasant. Rund um das fast allumfassende Thema der Digitalisierung ist Malware letztendlich ein digitaler Code,
der Computersysteme oder auch digitale Systeme an sich beeinflusst, direkt oder indirekt verändernd und meist
schädigend. Nur in den Frühzeiten war das Kreieren von Viren eine mehr oder minder sportliche Herausforderung um
der Ehre in der „Hall of Fame“ willen.
Die Anfänge der Malware reichen dabei in das Jahr 1949
zurück. Erst 11 Jahre vorher hatte Konrad Zuse seinen Zuse
Z1 fertiggestellt, einen frei programmierbaren mechanischen
Rechner. Aufgrund von Fertigungsproblemen war dieser
allerdings nie funktionstüchtig. 1941 gelang es Zuse dann,
mit seinem Z3 den ersten Computer der Welt in Betrieb zu
nehmen. Entscheidend war die Eigenschaft des Z3, beliebige
Algorithmen auszuführen, also die Möglichkeit einer freien
Programmierung war gegeben. Die Malware des Jahres
1949 war aber vorerst nur eine theoretische Überlegung
für sich selbst reproduzierende Computer programme. Das
Thema Computervirus begleitete in den Folgejahren die
Computerentwicklungen ebenso theoretisch.
Im Jahr 1971 wurde das erste sich selbst replizierende
experimentelle Programm, der erste Virus, mit dem
Namen „Creeper“ auf einer DEC PDP-10 unter dem TENEX
Betriebssystem aktiviert. Creeper konnte sich über das
ARPANET, einem Vorgänger des heutigen „Internet“, auf
weiteren DEC-Maschinen replizieren, schädigte diese
allerdings nicht, sondern zeigte seine Infizierung mit den
Worten „I´m the creeper, catch me if you can“ an. Eine
Antwort auf Creeper gab es vom selben Programmierer
auch: Reaper bewegte sich ebenfalls selbstständig im
ARPANET und zerstörte Creeper.
Heute wird die Zahl von Computerviren und anderer
Malware auf etwa 900 Millionen Typen geschätzt, eine
beachtliche Zahl, stellt man dieser die rund 5,3 Milliarden
Internetnutzer weltweit gegenüber, die zudem im Schnitt
mehr als ein Endgerät nutzen. Eine genaue oder auch nur
annähernd verlässliche Bezifferung für die durch Malware
verursachten Schäden, den Aufwand für deren Abwehr
oder Zahlungen von „Lösegeldern“ für durch Malware
ver schlüsselte Daten, ist nicht möglich, da die Grauzone
nicht gemeldeter Vorfälle beträchtlich sein soll. Eine
für Deutschland durchgeführte Umfrage des Branchenverbandes
BITKOM aus dem Februar 2020 unter Internetnutzern
zeigt für Schadware ein erschreckendes Bild: 46 %
der 1004 Befragten haben in den vorangegangenen
12 Monaten Erfahrungen mit Schadprogrammen gemacht,
meist über E-Mails oder über entsprechend präparierte
Webseiten. Eine weitere Statistik weist aus, dass im Schnitt
13 von 1000 E-Mails mit Malware präpariert sind.
Erfreulich ist, dass der überwiegende Teil der Befragten,
78 %, die Verantwortung in Sachen Datensicherheit
originär beim einzelnen User selbst sehen. So wird mit
85 % Abdeckung beim Virenschutz und 70 % bei der
Firewall eine schon beachtliche Vorsorge getroffen. Leider
ist der verbleibende Anteil der Rechner im Internet ausreichend,
um Computervirus-Infizierungen weiter zu
geben. Über den privaten oder beruflichen Umgang mit
Computern hinaus, besteht aber auch für industrielle
digitale Systeme das Risiko von Malwareangriffen.
Eine öffentlich bekanntere Malware ist der Stuxnet
Computer-Wurm. Dieser wurde gezielt entwickelt, um das
Atomprogramm des Iran zu sabotieren. Stuxnet war in der
Lage, speicherprogrammierbare Steuerungen (SPS) für
Urananreicherungs-Ultrazentrifugen zu manipulieren und
diese durch Fehlsteuerungen mechanisch zu schädigen.
Grundlage für die Verbreitung war die Erstinfektion eines
Computers mit dem Windows®-Betriebssystem über einen
USB-Stick – die Entwickler sind davon ausgegangen, dass
die im Fokus stehende Digitaltechnik als isoliertes System
arbeitet – und die folgende Verbreitung bis hin zu den SPS
im lokalen Netz. Stuxnet hat damit deutlich gemacht, wie
sensibel auch mit isolierten Netzwerken oder sogar
singulären Systemen zu agieren ist. Jeglicher „Kontakt“
mit der Außenwelt birgt potenzielle Risiken.
Aufgrund des besonderen und besonders hohen Stellenwerts
des Themas Sicherheit in der kerntechnischen
Industrie, ist auch das Thema Cybersicherheit eines mit
hohem Stellenwert und schon frühzeitigen Maßnahmen.
Grundsätzlich sind die sicherheitskritischen Systeme und
Sicherheitssysteme in kerntechnischen Anlagen „Inseleinrichtungen“.
Sie besitzen weder eine direkte Verbindung
mit dem Internet, noch sind sie mit anderen internen
Systemen bzw. Netzwerken verbunden, um mögliche
Hintertüren von vornherein auszuschließen – es sind
„ air-gapped“ Computer oder Netzwerke, die möglichst
sogar keine Hardware-Netzwerkschnittstellen besitzen, um
Einfallstore für Malware auf dieser Ebene auszuschließen.
Die noch in vielen kerntechnischen Anlagen vorhandene
fest verdrahtete Leit- und Regel- sowie Sicherheitsleittechnik
besitzt solche Anfälligkeiten nicht. Mit dieser
Erfahrung werden aktuell unter anderem Projekte für
Small and Medium Sized Reactors (SMR) vorangetrieben,
die auf Field Programable Gate Arrays aufbauen. Diese
Technologie verzichtet auf die Software-basierten und
damit Malware-kritischen Mikroprozessoren.
Des Weiteren gewährleistet bei Kernkraftwerken das
gestaffelte Sicherheitskonzept, ausgelegt und umgesetzt
für eine Vielzahl möglicher bzw. postulierter Fälle von
Einwirkungen auf Anlage und Anlagensicherheit, auch die
Sicherheit und den Schutz von Mensch und Umwelt bei
denkbaren Cyberangriffsszenarien.
Darüber hinaus ist, wie eingangs angemerkt, der Faktor
Mensch ein wichtiger beim Cyberschutz. Der sensible
Umgang mit jeglicher Art von Datenträgern, also potenziellen
Trägern von Malware, exakte Handlungs anweisungen
zum Umgang mit Hard- und Software, eine intensive und
sorgfältige Ausbildung und ständige Sensibilisierung für das
Thema Cybersicherheit sowie Maßnahmen zur Vermeidung
von Risiken gehören ebenso dazu wie kontinuierliche
Programme zur Prüfung und Optimierung der Robustheit
aller Maßnahmen an sich – sowohl auf der Seite von Hardware,
Software als auch der Soft-Skills der Beschäftigten.
Cybersicherheit ist ein Thema für die Kernenergie, sie ist
aber auch mit den vielfältigen Maßnahmen geschützt vor
Cyberattacken.
Christopher Weßelmann
– Chefredakteur –
Editorial
Cyber Security and Nuclear Power

Kommunikation und
Training für Kerntechnik
Suchen Sie die passende Weiter bildungs maßnahme im Bereich Kerntechnik?
Wählen Sie aus folgenden Themen: Dozent/in Termin/e Ort
3 Atom-, Vertrags- und Exportrecht
Atomrecht – Das Recht der radioaktiven Reststoffe und Abfälle RA Dr. Christian Raetzke 20.10.2020 Berlin
Export kerntechnischer Produkte und Dienstleistungen –
Chanchen und Regularien
Atomrecht – Was Sie wissen müssen
Atomrecht – Ihr Weg durch Genehmigungs- und
Aufsichtsverfahren
RA Kay Höft M.A. (BWL) 04.11.2020 Berlin
RA Dr. Christian Raetzke
Akos Frank LL. M.
11.11.2020 Berlin
RA Dr. Christian Raetzke 20.01.2021 Berlin
3 Kommunikation und Politik
Public Hearing Workshop –
Öffentliche Anhörungen erfolgreich meistern
Dr. Nikolai A. Behr 10.11. - 11.11.2020 Berlin
3 Rückbau und Strahlenschutz
In Kooperation mit dem TÜV SÜD Energietechnik GmbH Baden-Württemberg:
3 Nuclear English
Stilllegung und Rückbau in Recht und Praxis
Das Strahlenschutzrecht und
seine praktische Umsetzung
Dr. Stefan Kirsch
RA Dr. Christian Raetzke
Dr. Maria Poetsch
RA Dr. Christian Raetzke
23.09. - 24.09.2020 Berlin
29.10. - 30.10.2020 Berlin
English for the Nuclear Industry Angela Lloyd 07.10. - 08.10.2020 Berlin
3 Wissenstransfer und Veränderungsmanagement
Erfolgreicher Wissenstransfer in der Kerntechnik –
Methoden und praktische Anwendung
Veränderungsprozesse gestalten –
Herausforderungen meistern, Beteiligte gewinnen
Dr. Tanja-Vera Herking
Dr. Christien Zedler
Dr. Tanja-Vera Herking
Dr. Christien Zedler
05.10. - 06.10.2020 Berlin
24.11. - 25.11.2020 Berlin
Haben wir Ihr Interesse geweckt? 3 Rufen Sie uns an: +49 30 498555-30
Kontakt
INFORUM Verlags- und Verwaltungs gesellschaft mbH ı Robert-Koch-Platz 4 ı 10115 Berlin
Petra Dinter-Tumtzak ı Fon +49 30 498555-30 ı Fax +49 30 498555-18 ı Seminare@KernD.de
Die INFORUM-Seminare können je nach
Inhalt ggf. als Beitrag zur Aktualisierung
der Fachkunde geeignet sein.

atw Vol. 65 (2020) | Issue 6/7 ı June/July
306
Issue 6/7 | 2020
June/July
CONTENTS
Contents
Editorial
Cyber Security and Nuclear Power E/G 303
Inside Nuclear with NucNet
William Magwood – NEA Head Says Cost is
Driving Nuclear Industry Towards SMRs 308
Feature | Environment and Safety
Deep Geological Radioactive and Chemical Waste Disposal:
Where We Stand and Where We Go 311
Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . .317
Spotlight on Nuclear Law
No “Standstill in the Administration of Justice”
in Corona Times G 318
Environment and Safety
How Final Disposal Can Work 320
What has Happened
to the U.S. Nuclear Waste Disposal Program? 325
Safely Stored for All Eternity
How the Bundesgesellschaft für Endlagerung is Conducting
its Search for a Repository for High-level Radioactive Waste 331
Research and Innovation
Off-site Consequence Analysis During Severe Accidents
in a Nuclear Power Plant 334
Code and Data Enhancements of the MURE C++ Environment
for Monte-Carlo Simulation and Depletion 337
Modelling Thermal-hydraulic Effects of Zinc Borate Deposits
in the PWR Core After LOCA – Experimental Strategies
and Test Facilities 341
Investigation on PWR Neutron Noise Patterns 346
Operation and New Build
Reactor Core Control Based on Artificial Intelligence 350
Decommissioning and Waste Management
On the Potential to Increase the Accuracy
of Source Term Calculations for Spent Nuclear Fuel
from an Industry Perspective 353
Experimental Investigations into Flow Conditions
of Konrad Exhaust Air Channel 362
KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .366
Cover:
Picture alliance | Lehtikuva | Emmi Korhonen
G
E/G
= German
= English/German
News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .368
Nuclear Today
‘Green Energy’ Plans Will never Ripen
without Nuclear in the Mix 374
Imprint 309
Contents

atw Vol. 65 (2020) | Issue 6/7 ı June/July
307
Feature
Environment and Safety
311 Deep Geological Radioactive and
Chemical Waste Disposal:
Where We Stand and Where We Go
CONTENTS
Marcos Buser, André Lambert and Walter Wildi
Environment and Safety
320 How Final Disposal Can Work
Nicole Koch
325 What has Happened to the U.S. Nuclear Waste Disposal Program?
James Conca
331 Safely Stored for All Eternity
How the Bundesgesellschaft für Endlagerung is Conducting
its Search for a Repository for High-level Radioactive Waste
Steffen Kanitz
Operation and New Build
350 Reactor Core Control Based on Artificial Intelligence
Victor Morokhovskyi
Decommissioning and Waste Management
353 On the Potential to Increase the Accuracy of Source Term Calculations
for Spent Nuclear Fuel from an Industry Perspective
Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman
Contents

atw Vol. 65 (2020) | Issue 6/7 ı June/July
308
INSIDE NUCLEAR WITH NUCNET
William Magwood – NEA Head Says Cost
is Driving Nuclear Industry Towards SMRs
NuScale reactor could be on market by end of year ‘as a real product’
Competition in the nuclear industry –
including from China and Russia – is
leading to more choice in terms of reactor technology, but
financing and contract terms are often the determining
issue for many customers, with small modular reactors
attracting attention because of their affordability, NEA
Director-General William Magwood told NucNet.
Mr Magwood, who has been Director-General of the
Paris-based agency since 2014, said building large nuclear
plants can be expensive and customers need to find a way
to finance projects. “And that’s something that all suppliers
have to take into account,” he said.
He said the industry’s calls for market reforms that
would reward the security of baseload nuclear energy are
legitimate, but warned that the nuclear sector needs to
evolve to reflect the market. “The market is not going to
change overnight,” he said. “It’s going to take a long time
for it to be reformed.
“It would be to the nuclear sector’s advantage to have
products that fit the budgets of current customers under
current circumstance.”
Cost is one of the issues driving the market to consider
smaller reactors. This is because the initial capital needed
is so much less than for traditional large light-water
reactors (LWRs) of the kind that have been under
construction and faced delays and cost overruns at Vogtle
in the US, Flamanville-3 in France and Olkiluoto-3 in
Finland. Instead of talking about an investment of $10bn
or more, small modular reactors, or SMRs, might make it
to market for around $1bn billion, Mr Magwood said. That
is much more “in the affordability range” for a lot of
customers and has inevitably created a lot of interest.
However, Mr Magwood said he does not agree with the
notion that the industry has seen the last of the large
reactors. If SMRs are as successful as a lot of people hope,
the first examples could begin construction by the
mid-2020s and replace some large LWRs in the future.
“But until they [SMRs] are on the market, until they are
real, it’s hard to say,” Mr Magwood said.
Last month the author of a think-tank report said the
Vogtle-3 and -4 nuclear power plants under construction in
Georgia could become the last large-capacity reactors to be
built in the US, with SMRs and other Generation IV
advanced reactors taking over as key technologies.
Jane Nakano, a senior fellow specialising in energy
security and climate change at the Washington-based
Center for Strategic and International Studies, said she
“would not be surprised” if the two Westinghouse AP1000
units were the last large commercial units in the US.
But Mr Magwood said technologies like the AP1000
being used at Vogtle is excellent technology and “probably
the safest large reactor technology that’s been built”.
However, the Vogtle project has shown that for any
first-of-a-kind project you are going to run into some
issues, Mr Magwood said. “The good news on all of this is
that Vogtle-3 and Vogtle-4 plants are almost complete and
we will see this technology in operation.”
AP1000s have already been built in China and are
operating extremely well. “I have talked to Chinese
officials about the AP1000 and they are operating extraordinarily
well, they are very pleased with the plants. The
question is, what’s the market for the future?”
A lot depends on what demand looks like. In some
countries, particularly emerging economies like China and
India, there is very large growth in electricity demand.
More people are moving from rural areas to urban areas.
Factories are being built. The need for electricity increases.
In Western Europe and North America, electricity demand
is flat, increasing by about one percent or less a year.
In contrast to the large LWRs like the AP1000s at Vogtle,
SMRs fit into systems where electricity demand is not as
large, Mr Magwood said.
“In countries like the US, the question is not really of
meeting growing demand, but more of switching to
modern technologies to replace old coal plants that are
going offline,” he said. The issue is mostly one of replacing
existing capacity, not meeting increased demand.
“And that kind of market doesn’t lend itself to very large
investments in plant equipment like it used to. Which
makes SMRs more attractive than in markets where large
reactors are needed to meet growing demand.
“This is still somewhere where I think the large reactors
play a role,” Mr Magwood said. “They play a role where
there is large demand growth, but they also play a role in
situations where you need to retire very large facilities.
“If there are large coal plants that have to go offline
because they are too old, or even old nuclear plants, that
presents an opportunity to replace that capacity with new
large capacity.
“And in those cases, the traditional large plants might
fit. But that‘s something that has to decided on a case- bycase
basis.”
SMRs are still in the design stage, but construction and
operation are coming. In the US, NuScale’s SMR – a fully
factory-fabricated module capable of generating 60 MW of
electricity using a scaled-down version of PWR technology
– is the first to be going through the regulatory approval
process in the US and could be on the market by the end of
the year “as a real product”. That will be the first step to see
what the small reactor revolution might look like.
Mr Magwood also addressed the issue of financing new
nuclear, saying that if the market was completely open and
took into account the full system costs of all technologies,
nuclear would probably be better off. But he pointed out
that “it’s always important to recognise that it’s not exactly
a free market anywhere”.
He said many electricity markets are “heavily distorted
and dysfunctional” because of selective subsidies. “These
market imbalances make investing in nuclear power very
unfavourable in many countries,” he said.
According to Mr Magwood, business models for utilities
have changed and selling electricity is generating little
profit, or even a loss, in many countries.
“A situation has been created where a mechanism,
which had been so successful for so many years, where
revenue is generated through the sale of electricity to
enable investment into future plants and equipment, is
breaking down. That’s not a sustainable situation.”
Inside Nuclear with NucNet
William Magwood – NEA Head Says Cost is Driving Nuclear Industry Towards SMRs

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Utilities today, for example, are expecting nuclear
equipment vendors to come up with ready designs for
plants, but are unwilling or unable to pay for this element
of new-build projects. At the same time, Mr Magwood said,
vendors often do not have the financial resources to cover
the cost of getting a design ready for deployment.
Mr Magwood said there are some policy approaches,
including the regulated asset-base (RAB) model being
considered in the UK, that could be more favourable to
large capital projects and be an incentive to nuclear. Large
wind farms have large capital costs and could also benefit
from “reforms to the market”.
“On the other hand, the nuclear sector needs to evolve
to reflect the market,” he said. The market is not going to
change overnight. It’s going to take a long time for it to be
reformed.
“It would be to the nuclear sector’s advantage to have
products that fit the budgets of current customers under
current circumstance.”
And so we come back to SMRs and to microreactors that
can be built quicker and more easily than the large LWRs.
“And I think that’s one of the things that’s really driving
this interest in small reactors – the idea that instead of
investing in 2,000 MW you can build 300 MW now and add
another 300 MW when it’s needed, until you get to the
2,000 you’re looking for,” Mr Magwood said.
“And that way, every time you install a 300-MW system
and put it on the grid, you are making money back and
starting to recover your costs, while you start constructing
the next module.
“I mean, that‘s a model that is very attractive to a lot of
people. And I think that’s something that the nuclear
sector is going to have to do if it‘s going to survive over the
next decade or so.”
Author
NucNet – The Independent Global Nuclear News Agency
Editor responsible for this story: Kamen Kraev
Avenue des Arts 56 2/C
1000 Bruxelles
www.nucnet.org
Imprint
Official Journal of Kerntechnische Gesellschaft e. V. (KTG)
Publisher
INFORUM Verlags- und Verwaltungsgesellschaft mbH
Robert-Koch-Platz 4, 10115 Berlin, Germany
Phone: +49 30 498555-33
Fax: +49 30 498555-18
www.nucmag.com
General Manager
Dr. Thomas Behringer
Editor in Chief
Christopher Weßelmann
+49 2324 4397723
editorial@nucmag.com
Managing Editor
Nicolas Wendler
+49 30 498 555-21
nicolas.wendler@nucmag.com
Editorial Advisory Board
Frank Apel
Erik Baumann
Dr. Erwin Fischer
Carsten George
Eckehard Göring
Dr. Florian Gremme
Dr. Ralf Güldner
Carsten Haferkamp
Christian Jurianz
Dr. Anton Kastenmüller
Prof. Dr. Marco K. Koch
Ulf Kutscher
Herbert Lenz
Jan-Christan Lewitz
Andreas Loeb
Dr. Thomas Mull
Dr. Joachim Ohnemus
Olaf Oldiges
Dr. Tatiana Salnikova
Dr. Andreas Schaffrath
Dr. Jens Schröder
Norbert Schröder
Prof. Dr. Jörg Starflinger
Dr. Brigitte Trolldenier
Dr. Walter Tromm
Dr. Hans-Georg Willschütz
Dr. Hannes Wimmer
Layout
zi.zero Kommunikation
Antje Zimmermann
Berlin, Germany
Editor
Nicole Koch
+49 176 841 846 04
nicole.koch@nucmag.com
Advertising and Subscription
Petra Dinter-Tumtzak
+49 30 498555-30
petra.dinter@nucmag.com
Price List for Advertisement
Valid as of 1 July 2020
Published monthly, 9 issues per year
Germany:
Per issue/copy (incl. VAT, excl. postage) 23.55 €
Annual subscription (incl. VAT and postage) 183.50 €
All EU member states without VAT number:
Per issue/copy (incl. VAT, excl. postage) 23.55 €
Annual subscription (incl. VAT, excl. postage)183.50 €
EU member states with VAT number
and all other countries:
Per issues/copy (no VAT, excl. postage) 22.43 €
Annual subscription (no VAT, excl. postage) 174.77 €
Copyright
The journal and all papers and photos contained in it
are protected by copyright. Any use made thereof
outside the Copyright Act without the consent of the
publisher, INFORUM Verlags- und Verwaltungsgesellschaft
mbH, is prohibited. This applies to reproductions,
translations, micro filming and the input and incorporation
into electronic systems. The individual author is
held responsible for the contents of the respective
paper. Please address letters and manuscripts only
to the Editorial Staff and not to individual persons
of the association's staff. We do not assume any
responsibility for unrequested contributions.
Signed articles do not necessarily represent the views
of the editorial.
INSIDE NUCLEAR WITH NUCNET 309
Printing
inpuncto:asmuth
druck + medien gmbh
Buschstraße 81, 53113 Bonn, Germany
ISSN 1431-5254

atw Vol. 65 (2020) | Issue 6/7 ı June/July
CALENDAR 310
Calendar
2020
Virtual 13.07. – 16.07.2020
46 th NITSL Virtual Conference – Fusing Power
& People. Virtually Hosted, www.nitsl.org
Virtual 03.08. – 06.08.2020
ICONE 28 – 28 th International Conference on
Nuclear Engineering. Virtually Hosted,
www.event.asme.org/ICONE
30.08. – 04.09.2020
IGORR – Standard Cooperation Event in the
International Group on Research Reactors
Conference. Kazan, Russian Federation, IAEA,
www.igorr2020.org
09.09. – 11.09.2020
World Nuclear Association Symposium 2020.
London, United Kingdom, WNA World Nuclear
Association, www.wna-symposium.org
Postponed to 13.09. – 17.09.2020
Jahrestagung 2020 – Fachverband Strahlenschutz
und Entsorgung. Aachen, Germany, Fachverband
für Strahlenschutz, www.fs-ev.org
21.09.-25.09.2020
64 th IAEA General Conference. Vienna, Austria, International
Atomic Energy Agency IAEA,
www.iaea.org
30.09. – 03.10.2020
Nuclear Energy: Challenges and Prospects. Sochi,
Russia, Pocatom, www.nsconf2020.ru
06.10. – 08.10.2020
HTR2020 – 10 th International Conference
on High Temperature Reactor Technology.
Yogyakarta, Indonesia, Indonesian Nuclear Society,
www.htr2020.org
11.10. – 15.10.2020
RRFM – European Research Reactor Conference.
Helsinki, Finland, European Nuclear Society,
www.euronuclear.org
11.10. – 17.10.2020
BEPU2020– Best Estimate Plus Uncertainty International
Conference, Giardini Naxos. Sicily, Italy,
NINE, www.nineeng.com
09.11. – 13.11.2020
International Conference on Radiation Safety:
Improving Radiation Protection in Practice.
Vienna, Austria, IAEA, www.iaea.org
15.11. – 19.11.2020
ANS Winter Meeting and Nuclear Technology
Expo. Chicago, Illinois, US, American Nuclear Society,
www.ans.org
18.11. – 19.11.2020
INSC — International Nuclear Supply Chain
Symposium. Munich, Germany, TÜV SÜD,
www.tuvsud.com
23.11. – 25.11.2020
KELI 2020 – Conference for Electrical Engineering,
I&C and IT in generation plants. Bremen, Germany,
VGB PowerTech e.V., www.vgb.org
24.11. – 26.11.2020
ICOND 2020 – 9 th International Conference on
Nuclear Decommissioning. Aachen, Germany,
AiNT, www.icond.de
Postponed to 30.11. – 02.12.2020
Enlit (former European Utility Week and
POWERGEN Europe). Milano, Italy,
www.powergeneurope.com
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
08.12. – 10.12.2020
World Nuclear Exhibition 2020. Paris Nord
Villepinte, France, Gifen,
www.world-nuclear-exhibition.com
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. Dates are subject to change.
Please check the listed websites for updates.
Postponed to 08.09. – 10.09.2021
3 rd International Conference on Concrete
Sustainability. Prague, Czech Republic, fib,
www.fibiccs.org
27.09. – 01.10.2021
NPC 2021 International Conference on Nuclear
Plant Chemistry. Antibes, France, SFEN Société
Française d’Energie Nucléaire,
www.sfen-npc2021.org
Postponed to June 2021
International Forum on Enhancing a Sustainable
Nuclear Supply Chain. Helsinki, Finland, Foratom,
https://events.foratom.org/mstf2020/
Postponed to 2021
The Frédéric Joliot/Otto Hahn Summer School
on Nuclear Reactors “Physics, Fuels and Systems”.
Aix-en-Provence, France, CEA & KIT, www.fjohss.eu
Postponed to 2021
International Conference on Operational Safety
of Nuclear Power Plants. Beijing, China, IAEA,
www.iaea.org
Postponed to 2021
INDEX 2020: International Nuclear Digital
Experience. Paris, France, SFEN,
www.sfen-index2020.org
Cancelled
NuclearEurope 2020 – Nuclear for a sustainable
future. Paris, France, Foratom,
events.foratom.org/nuclear-europe-2020
2022
19.10. – 23.10.2020
International Conference on the Management
of Naturally Occurring Radioactive Materials
(NORM) in Industry. Vienna, Austria, IAEA,
www.iaea.org
2021
KERNTECHNIK 2022.
Germany, KernD and KTG,
www.kerntechnik.com
20.10. – 23.10.2020
ATH'2020 – International Topical Meeting
on Advances in Thermal Hydraulics.
Paris, France, SFEN, www.sfen-ath2020.org
26.10. – 30.10.2020
NuMat 2020 – 6 th Nuclear Materials Conference.
Gent, Belgium, IAEA, www.iaea.org
04.11. – 05.11.2020
The Power & Electricity World Africa 2020.
Johannesburg, South Africa, Terrapinn,
www.terrapinn.com
Postponed to 10.05. – 15.05.2021
FEC 2020 – 28 th IAEA Fusion Energy Conference.
Nice, France, IAEA, www.iaea.org
Postponed to 31.05. – 04.06.2021
20 th WCNDT – World Conference on
Non-Destructive Testing. Incheon, Korea,
The Korean Society of Nondestructive Testing,
www.wcndt2020.com
Calendar

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Deep Geological Radioactive and
Chemical Waste Disposal:
Where We Stand and Where We Go
Marcos Buser, André Lambert and Walter Wildi
Introduction A recognized waste disposal concept and its troubles
For about 40 years, deep geological disposal of radioactive
and chemical waste has become the most widely recognized
strategy for eliminating waste. However, this pole position
in the ranking of concepts contrasts with the daily lived
situation in the field, as exposed here.
In 1976, the International Atomic Energy Agency
published a brochure entitled “Radioactive Waste – Where
from – Where to”; its cover picture showed a schematic
cross-section of the Asse II repository for low and intermediate
level waste in Wolfenbüttel (Germany). The
contents of the brochure revealed that the nuclear industry
and international organisations were confident about the
feasibility and long-term safety of repositories for radioactive
waste. This confidence persisted until after the
turn of the millennium, despite all the difficulties and
problems that were persistent and became apparent in the
selection of sites for deep disposal infrastructures or the
implemen tation of concrete projects. In 2002, a fire broke
out in the Stocamine (Alsace, France) underground storage
facility for chemo-toxic waste, which signalled the end
of the project, and for the first questioned the long-term
safety of geological repositories 1 . If this event could be
attributed to the lack of safety culture in the final disposal
of non-radioactive waste, this could not explain the water
inflow from the overlying strata into the former Asse II
experimental repository mine, which became known by
the public in 2008. This was when the responsible operators
publicly admitted for the first time that there was an inflow
of water into the repositories and also the existence of
potential hydrogeological hazards. This is a fact that was
known by the monitoring staff since 1988 (or even before) 2 .
Another German repository for radioactive waste in
Morsleben (ERAM) showed similar stability problems and
indications of leachate intrusion. These needed extensive
stabilisation measures which cost billions of Euros 3 .
Finally, between 2014 and 2017, various incidents and
accidents occurred at the Waste Isolation Pilot Plant (WIPP,
New Mexico), the repository for trans-uranium radioactive
waste, which above all put into question the safety culture
and governance of the facility 4 . The conditions for a safe
implementation of a repository in the WIPP model project
seemed to be particularly favorable, as the framework
conditions for comprehensive, safety- oriented management
of the project were clearly set. “ Fifteen years of smooth,
uneventful operations had lulled these sites into routines
and practices inconsistent with the discipline and order that
is in the centre of a ‘nuclear culture’” 5 , as described by an
insider about the loss of safety culture. Another observer
regretted that the inves tigating authorities failed to identify
the real causes of the event 6 . Lessons were, of course,
learned from these incidents. Also, numerous investigations
have been carried out on the incidents and accidents,
and several reports have been published. However, the
question regarding the effectiveness and sustainability of
this learning process remains open.
“Lessons Learned”
As a Basis to Establish a New Safety Culture
At least since the publication of Charles Perrow’s book on
“Normal Accidents” in 1984 7 , planners, builders and
operators of high-risk technologies and facilities have
increasingly perceived the need to protect their large-scale
technological projects and facilities from avoidable errors
and from crashes that are very costly and can damage their
image. This led to the development of methodological
instruments in a wide variety of government and economy
sectors, which were designed to detect and correct sources
of errors at an early stage of a technological development
and production process. A number of these methods are
briefly mentioned below.
“Lessons learned” is the most frequently used term
when it comes to evaluating running or future projects and
programs. The term originally comes from the Anglo-
Saxon industrial world and has subsequently spread and
established itself in project and knowledge management 8 .
What makes “lessons learned” so attractive as a term is a
fact that it can be used in any field and it conveys a fundamentally
positive message. Errors do not necessarily have
to be understood in every detail; what is more important is
how to eliminate them. With “lessons learned” one wants
to show that a certain project and program is under control
and that one is able and willing to learn and thus to correct
errors. However, the term has weaknesses in the universal
claims to accomplish projects and in its applicability. As a
rule, “lessons learned” do not lay claim to standardization,
and does not guarantee a more comprehensive quality
assurance process; particularly it does not promise that a
process can be reflected and reviewed in its entirety.
Over the last decades, a large number of different
methods have been developed and used to evaluate and
311
FEATURE | ENVIRONMENT AND SAFETY
1 Copil, 2011, Expert report, Steering committee, June 2011;
2 Ibsen, D., Kost, S., Weichler, H., 2010, analysis of the usage history and the forms of planning and participation of the Asse II mine, final report AEP, University of Kassel; Möller, D., 2009,
Final disposal of radioactive waste in the Federal Republic, Peter Lang.
Blum, P., Goldscheider, N., Göppert, N., Kaufmann-Knoke, R. et al., 2016, groundwater – humans - ecosystems, 25 th conference of the FH-DGGV, Karlsruhe, 13.-16. April 2016,
KIT Scientific Publishing, p. 152;
3 Beyer, F. 2005, The (GDR) history of the Morsleben nuclear waste repository. “Contributions in kind”, No. 36, Magdeburg 2005..
4 Augustine N., Mies R. et al, 2014, A New Foundation for the Nuclear Enterprise, Report of the Congressional Advisory Panel on the Governance of the Nuclear Security Enterprise, November
2014; Klaus, D. 2019, What really went wrong at WIPP: An insider’s view on two accidents at the only underground nuclear waste repository, Bulletin of the Atomic Scientists, 75(4), pp. 197-204.
5 Klaus, D. 2019, What really went wrong at WIPP: An insider’s view on two accidents at the only underground nuclear waste repository, Bulletin of the Atomic Scientists, 75(4), pp. 197-204.
6 Ialenti, Vincent, 2018, Waste makes haste. How a campaign to speed up nuclear waste shipments shut down the WIPP long-term repository, in: Bulletin of the Atomic Scientists, 74.
7 Perrow, Charles, 1984, Normal Accidents: Living with High Risk Technologies, Princeton University Press.
8 Milton, N., 2010, The Lessons Learned Handbook: Practical approaches to learning from experience, Elsevier.
Feature
Deep Geological Radioactive and Chemical Waste Disposal: Where We Stand and Where We Go ı Marcos Buser, André Lambert and Walter Wildi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
FEATURE | ENVIRONMENT AND SAFETY 312
optimise processes, all of which follow the so-called
“ top-down” approach, i.e. the hierarchically prescribed
decision paths. The range of methods developed is broad
and extends from benchmarking in the field of economic
comparability of processes and projects 9 , through best
practice in business administration 10 , auditing and quality
assurance programmes in the monitoring of companies and
industrial processes 11 , to risk management in the application
of risky projects or risk technologies. The latter, in
particular, is characterised by a strong standardisation of
process sequences and contents, whereby this also includes
organisational references. As a rule, the method of risk
management differs fundamentally from that of “lessons
learned” in terms of stringency and quality level of its
procedure. As for other quality assessment processes, risk
management is also defined by guidelines of the International
Organization for Standardization (ISO), and in
particular by (ISO 31000).
A method specially adapted to risk issues is the so-called
safety culture, which is applied in high-risk areas such as
nuclear energy, and also in medical fields 12 . The safety
culture focuses not only on standardised procedures for
determining risks (e.g. event and fault tree analyses, safety
analysis) but also on the safety management of an organisation
and therefore strongly addresses questions of the
organisation of a company and the relationship between
the company and its employees. This also includes the
processes of supervision and control, the documentation
of process sequences and establishment of chains of errors,
the management of processes and conflict management,
and the methods used for their correction. What makes
safety culture fundamentally different from other processes
is the emphasis on the term “culture”, which implies
that the people involved in a system actively shape a
process. In this way, safety culture transcends the purely
technical-scientific level and elevates to issues of organisational
structures and the behaviour and behavioural
interplay of organisations, their staff and collaborators.
The safety culture in the field of nuclear energy was introduced
after the Chernobyl reactor accident 13 .
Of all these methods the one to be used to improve processes
in a particular project depends on the preferences of
the institutions and organisations doing the project. In our
context, we will mainly apply terms that are characterised
by standardised and well-defined methods.
A Review of Concepts and Failures
in Nuclear Waste Management
A review of nuclear waste management over the past
75 years can be focussed on both the concepts proposed
and the success of the strategies and projects implemented
to date. The concepts of nuclear waste management
developed over decades can be found in a large number of
publications. It is worth remembering the writings of
Bürgisser et al. (1979) 14 , Milnes et al. (1980) 15 , Milnes
(1986) 16 , the Swiss expert group EKRA (2000) 17 , or the
recently published research reports in the German Entria-
Project (Appel et al. 2014/2015) 18 . They describe most of
the concepts that have been put forward or implemented
by different authors and institutions since the late 1940s
(see Table 1). If we examine the maturity of these
concepts, it is striking that most of the ideas for dealing
with radioactive waste were not technically mature, were
not considered, or could not be considered with respect to
risk considerations. Also, most of these concepts were
based on ideas that originated from university institutions
or military agencies and whose technical implementation
had not been tested adequately and deeply. An example of
how quickly ideas are caught up by reality can be seen in
the concept of final storage in polar ice shields, an idea that
was widely discussed by scientists in the 1950s and that
was then considered as completely obsolete a few decades
later.
The situation was quite different, however, for the two
concepts of dilution and containment, which emerged in
the late 1940s. Dilution was implemented in the early days
of nuclear energy use, mainly for cost reasons. It was done
by sea dumping, dilution in rivers or dumping of solid,
liquid or slurry materials in landfills or percolation ponds,
as is also explained in many early publications 19 . At the
military plutonium factory in Hanford (Washington), for
example, the cooling water for the plutonium-breeding
reactors was fed directly into the Columbia River via a
settling basin. Other large research laboratories, such as
the Oak Ridge National Laboratory (Tennessee), similarly
handled their liquid waste. At the Windscale/Sellafield
reprocessing plant, the conviction prevailed until well into
the 1960s, when there were serious discussions about
diluting the entire global inventory of highly active fission
products in the oceans 20 . It was not until the end of the
1950s that the concerns of the radiation protection authorities
became increasingly widespread and led to
the gradual reduction and abandonment of the dilution
principle. However, sea dumping of L / ILW waste
continued into the 1980s 21 . In the 1970s, the increasing
social discussion and questioning of the dilution and
dumping strategies finally led to the specification of a
strategy for the containment of radioactive substances,
which is essentially covered by the multiple-barrier
concept still valid today. The idea of containment, which
can be traced back to the late 1940s 22 and early 1950s 23
received decisive impetus in the 1970s from the American
programmes (ERDA/DOE), the “sub-seabed-disposal”
project and the Swedish disposal programme (SKB) 24 . The
concept of various barriers connected in series according
to the principle of the Russian doll (“Multi-barriers”) has
9 Zairi, M., Leonard, P., 1996, Practical Benchmarking: The Complete Guide, Springer Science+Business Media Dordrecht.
10 Bardach, E., 2011, A Practical Guide for Policy Analysis, Sage Publications; Bretschneider, S., Marc-Aurele, F.J., Wu, J., 2005,
“Best Practices” Research: A methodological guide for the perplexed, Journal of Public Administration Research and Theory (15)2:307-323.
11 Matthews, D., 2006, History of Auditing, Routledge.
12 International Organization for Standardization, ISO 9’000 and ISO 14’000. Guldenmund, F. W., 2000, The nature of safety culture: a review of theory and research, Safety Science, 34, 215-257
13 NSAG, 1991, Safety Culture, Safety Series No 75-INSAG-4, International Nuclear Safety Advisory Group, IAEA.
14 Bürgisser, H., et al., 1979, Geological aspects of radioactive waste disposal in Switzerland, Switzerland. Energy foundation.
15 Milnes, A.G., Buser, M. & Wildi, W. 1980: Overview of final disposal concepts for radioactive waste. - Z. dtsch. Geol. Ges. 131, 359-385.
16 Milnes, A.G.,1985, Geology and Radwaste, Academic Press.
17 EKRA, 2000, Disposal Concepts for Radioactive Waste, Final Report, 31st January 2000.
18 Appel. D., Kreusch, J., Neumann, W., o.J., presentation of disposal options, ENTRIA report 01 (first published 2014/2015)
19 Scott, K., 1950, Radioactive Waste Disposal - How Will It Affect Man’s Economy, Nucleonics, Vol. 6/1, p. 15-25.
20 Glückauf, E., 1955, The long-term problem of the disposal of radioactive waste, Proceedings of the international conference on the peaceful uses of atomic energy,
held in Geneva from 8 to 20 August 1955, volume IX, IAEA, 1956
21 IAEA TECDOC-1105 “Inventory of radioactive waste disposals at sea” August 1999 retrieved 2011-12-4.
22 Western, Forrest, 1948, Problems of Radioactive Waste Disposal, Nucleonics 3/2, August 1949, p. 43-49.
23 Hatch, L. P., 1953, Ultimate Disposal of Radioactive Waste, American Scientist Vol. 41/3, p. 410-412.
24 Hollister, C.D., 1977, The Seabed Option, Oceanus 20, p. 18-25; KBS, 1978a, Handling of spent fuel and final storage of vitrified high-level reprocessing waste, Kärnbränslesäkerhet; KBS, 1978b,
Handling and final storage of unreprocessed spent nuclear fuel, Kärnbränslesäkerhet.
Feature
Deep Geological Radioactive and Chemical Waste Disposal: Where We Stand and Where We Go ı Marcos Buser, André Lambert and Walter Wildi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Waste Management
Concepts
HLW: immobilization
in clay / ceramics
HLW: vitrification &
ceramics
HLW & LILW: disposal
in near-surface strata
LILW (& HLW?):
dilution & seepage
Specification Comment Author and
Year
smectites
(montmorillonites)
borosilicate glasses
and ceramics
dump or land burial
ventilation of gases /
drainage of fluids
Hatch 1953;
Ginell et al. 1954,
proposed since 1951 Herrington et al. 1953;
Rodger 1954
as part of the
nuclear fuel chain
Publication
Amer. Scientist 41/3
Nucleonics 12/12
Nucleonics 11/9 Nucl.
Engineering 50/
Status of
Implementation
no direct disposal
current application
(vitrification)
Result and
Success
laboratory-tested
laboratory-tested
Goodman 1949 Nucleonics 4/2 widely implemented basically failed, wide
pollutions
Beers 1949 ;
Browder 1951,
de Laguna et al. 1958
LILW: injection in boreholes or wells Herrington
et al. 1953
LILW (& HLW):
sea dumping
HLW: subsea bed
disposal
LILW & HLW:
geological disposal
HLW: disposal
in subduction zones
HLW: disposal
in fault zones
dumping / dilution
in sea water
final disposal
in marine sediments
diverse host-rocks
in mines
submarine repository
in subducting plate
regulated after 1972
by London Convent.
from 1977 as
“sub-seabed”-project
mostly
in disused mines
Nucleonics 4/4 & 6/1
Nucleonics 6/1
Nucleonics 11/9
widely implemented
widely implemented
(UdSSR, USA)
basically failed, wide
pollutions
effects not known,
DSP-principle
Claus 1955 IAEA 1955 P/848 widely implemented basically failed,
DSP-principle
Evans 1952 NSA 8, 1954: 4929 project abandoned not achieved
Theis 1955
NAS 1957
deep-sea trenches Renn 1955;
Bogorov et al. 1959
IAEA 1955 P/564
Report
widely implemented
mostly damaged or
under observation
Bostrom et al. 1979 Nature 1970, 228 idea abandoned not developed
IAEA P/569
IAEA 1958, P/2058
HLW: disposal in ice Antarctic repository meltdown in ice Philbert 1959 Atomkernenergie
4/3
HLW: meltdown in the
deep underground
deep underground
melting
melting in atomically
generated cavern
idea abandoned
idea abandoned
not developed
not developed
Gilmore 1977 NDC-Publication idea abandoned not developed
HLW: Disposal in space Hollocher 1975 MIT Press idea actually
abandoned
HLW: partitioning and
transmutation
Cost-based implementation
of disp. practices
| Tab. 1.
Historical management concepts.
long-lived species
conversion
reduction
of disposal time
Cecille et al. 1977
Hage, W., 1978
IAEA 1977 36/366
EUR-5897
research
still in progress
not feasible
(costs, risks)
uncertain (costs,
success, risks)
reduction of costs Scott 1950 Nucleonics 6/1 still central cost-related practice
has consistently failed
FEATURE | ENVIRONMENT AND SAFETY 313
remained more or less unchanged even after several
decades; it speaks for the great acceptance and the almost
unchallenged conceptual stringency of this approach.
However, the concrete success of this concept can only be
“proven” more or less reliably after its implementation, the
emplacement of the waste in the storage media and the
longer-term monitoring of the repositories in the deep
geological underground.
Two conclusions can be drawn at that stage from the
compilation of the concepts for nuclear disposal:
p On the one hand, all relevant ideas and concepts of
nuclear disposal were already formulated at a time
when industrial use by nuclear power plants was
beginning to emerge. Indeed, important scientific
representatives of the nuclear community – first and
foremost Enrico Fermi and James Conant – had pointed
out the challenges and risks of radioactive residues and
their disposal 25 . But the implementation of nuclear
waste management was considered feasible a priori by
the majority of involved institutions and scientists. This
way of thinking has remained unchanged until today.
p On the other hand, it became clear from the very
beginning which concepts of disposal were based solely
on ideas that – published in scientific journals – were
noticed by the scientific community and caused discussions
at congresses and conferences. With the
exception of the Sub-Seabed Disposal Project, which
was led by the Woods-Hole Oceanographic Institute,
Massachusetts, and Sandia Laboratories, Albuquerque, 26
none of the numerous ideas outside of continental
disposal reached a conceptual technical and economic
maturity that would have given reasons to trust and use
them for a successful implementation of a project.
As early work on the topic shows, the implementation of
long-term safe disposal was strongly influenced by the cost
pressure on the various national reactor programmes 27 .
A large part of the difficulties that arose in the actual
disposal process is due to the lack of finance and
implementation of better programmes. The idea of the
chairman of the American Atomic Energy Commission,
Lewis Strauss, that nuclear energy is “too cheap to meter” 28 ,
reflected the prevailing opinion that nuclear disposal
was not only feasible but also practically at zero cost.
This misconception that economic criteria should take
precedence over safety considerations is probably the main
reason for the misguided developments in waste management
policy to date. And so, it is not surprising that under
such conditions, one waste management project after the
other ran into difficulties and the list of initiated but failed
projects is constantly growing (Table 2). Contrary to
the requirements of a comprehensive safety culture, the
required practices have not been dealt systematically,
which led to serious reservations in the acceptance of
disposal programmes to this day, as we shall see later 29 .
Trouble Shooting in Waste Management
and Improving of Geological Waste Disposal
Projects
The lessons learned by repository planners worldwide
from past failures consisted primarily in adapting the
concept for geological repositories. This adaptation was
nothing more than a further development of the old
25 Buser, M., 2019, Where to go with nuclear waste, Rotpunkt Verlag Zürich, p. 38, 53-54.
26 Hollister, Ch., Anderson, D. R., Health, G. R., 1981, Subseabed Disposal of Nuclear Wastes, Science, Vol. 213, 18 Sep 1981.
27 Scott 1950, S. 18–25; Herrington et al. 1953, S. 34–37; Ford 1982, 208-210.
28 Strauss, Lewis, 1954, Remarks For The Delivery At The Founder’s Day Dinner, National Association of Science Writers, New York, 16. September 1954, Atomic Energy Commission, p. 9
29 The cases of Asse and WIPP may be exceptions.
Feature
Deep Geological Radioactive and Chemical Waste Disposal: Where We Stand and Where We Go ı Marcos Buser, André Lambert and Walter Wildi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
FEATURE | ENVIRONMENT AND SAFETY 314
mining concept with one major difference: Disused mines
should no longer be converted into repositories. New
facilities were now planned which were to serve the sole
purpose of final disposal. The first country to present a
detailed concept for such a geological repository was
Sweden. As mentioned in chapter 3, almost all newer
nuclear waste disposal projects around the world followed
this KBS – multi-barrier concept developed by the Swedish
company SKB (Svensk Kärnbränslehantering AB) in the
1970ties. After that, many countries developed their
specific design variants with regard to the importance of
the individual barriers, to the access structures (ramp/
shaft) or the positioning of the canisters in the disposal
galleries. But these minor changes lastly did not deviate
from the original concept, which still assumes a geological
repository at depths of several hundred meters in a system
of galleries. With this adaptation, the main conceptual
flaw seemed to be resolved and the requirement to identify
and correct the main planning flaw was satisfied. Further
analyses, which sought answers to possible risk or breakpoints
in the concepts and the procedure for implementing
the programmes, were not required. The responsible
institutions were satisfied with the results achieved and
no longer questioned the emerging developments. Even
before the turn of the millennium, it became clear that
there was a need for action, as can be briefly illustrated by
three aspects:
Public implication and responsibility
On the one hand, the official institutions entrusted with
the project development have underestimated for a long
time the problems concerning the social acceptance of
repositories for long-lived highly toxic waste. If waste management
projects are ever to be realized, they must
be supported by the public opinion and the affected
population. After decades of debate, this insight seems to
be more or less accepted by all stakeholders. But the degree
of involvement of concerned regions and people is still
disputed. A fundamental question in this context is,
how far can the rights and responsibilities of affected
communities go? Is it a simple participation right, that
makes discussions possible but does not go beyond them
or that leaves decisions in the hands of the repository
designers and authorities? Or do these latter want to leave
some of the key decisions to those affected? If yes, how
many? How much can and should be decided jointly? Is
the blockage caused by “NIMBY” due to these questions?
One can answer them partly from experience, but only
partly. Today’s projects are planned still exclusively based
on scientific and technical expert knowledge. In contrast,
the ethical, political, but also technical concerns of the
public on questions of nuclear safety, public health and
environmental impact are still treated negligently, as
the Swiss case of the “sectoral plan for deep geological
repositories” shows very clearly. These projects institutionalize
“participation” and even public forums – socalled
“regional conferences” – and claim to remedy these
deficiencies. They do not, however, give the concerned
population any real responsibility, i.e. no voice for codecision,
which ultimately strengthens the resistance
against such projects. “Safety is not negotiable”, as the
Federal Office of Energy (SFOE) repeatedly stated.
From the Office’s point of view, the so-called “licence
holder” or “operator” and his experts are responsible for
safety, which is monitored by the authorities. However,
how can it be explained that with the continuous
Repository,
Owner
Hutchinson-Mine,
Kansas (USA), ORNL
Lyons Kansas (USA),
ORNL
Asse II Mine (FRG),
(Test Disposal Site),
several owners
ERAM Morsleben GDR/
FRG, several owners
WIPP
DOE
Waste-
Type
Host
Rock
Operation
Period
HLW salt test-phase
1959 - 1961
HLW salt test-phase 1965 - 1968
Project 1970 - 1972
LILW with
TRUwastes,
CTW
salt 1967 - 1978
from 2008 onwards
Status of
Implementation
tests with non-radioactive
liquids and heaters
tests with fuel elements
Site selection
in operation, remediation
project
LILW, CTW salt 1971 -1998 in operation, remediation
project
TRU-
Wastes
salt 1999 - 2014,
from 2017 onwards
in operation, remediation
project completed
Result and
Success
“encouraging but not
conclusive”
site selection, abandoned
(vulnerable site)
site abandoned
(vulnerable site),
remediation in planning
site abandoned (vulnerable
site), remediation under way
site still in operation, although
seriously questioned
Olkiluoto (FI), Posiva LILW crystalline since 1992 in operation site in operation, long-term
safety questioned
Forsmark (SE), SKB LILW crystalline since 1988 in operation site in operation, long-term
safety questioned
Examples of shallow
subsurface mines
Hostim (HU),
several owners
Mina Beta (ES)
JEN/CIEMAT
Research
wastes
limestone 1959 - 1964,
closure 1997
final repository
vulnerable site (limestone),
long-term safety questioned
LILW crystalline 1961 - 1980 remediated remediation successfully
achieved
Bratrstvi (CZ), Súrao LILW (MIR) pegmatites since 1974 in operation vulnerable site (uranium
mine), long-term safety open
Alcazar (CZ) LILW, CTW limestone 1959 - 1964,
1991
Richard II (CZ), Súrao
ORNL (USA), injection
in boreholes or wells
Russian sites, injection
in boreholes or wells
(USSR)
LILW
((MIR)
LILW
(HLW)
final repository, reopened
1991, higher toxic radwaste
and CTW removed
limestone since 1964 in operation, refurbishment
2005-2007
LILW/TRUwastes
LILW/TRUwastes
LILW
(HLW)
potentially vulnerable site,
long-term safety open
potentially vulnerable site,
long-term safety open
1950 - 1980ies completed monitoring data show
remobilisation, results only
partially available
since 1957
(Tomsk-7, Krasnoyarsk-26,
Dimitrovgrad etc.)
| Tab. 2.
Implementation, result and success of geologic repositories for nuclear wastes (sources in bibliography).
Author and
Year
Walker, S. jr., 2006
Boffey 1975, Walker, S.jr.
2009, Alley et al. 2013
Möller 2008
BGE 2020
Documentation of BGE
DOE 2014a, 2014b, 2015,
Ialenti 2018, Klaus 2019
Buser 2019, WNWR 2019
Buser 2019, WNWR 2019
WNWR 2019
Lopez Perez et al 1976,
Estratos 1987
Woller 2008,
WNWR 2019
Woller 2008
Woller 2008, WNWR
2019
ERDA 1977; ORNL,
1985; Stow et al. 1986
completed? unknown Spytsin et al. 1975; NDC
1977; Schneider et al.
2011
Feature
Deep Geological Radioactive and Chemical Waste Disposal: Where We Stand and Where We Go ı Marcos Buser, André Lambert and Walter Wildi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
occurrences of serious problems and accidents none –
and really none – of the deep geological repository
projects implemented to date have been able to meet the
required quality standards (Table 2). This is because
failure to plan waste disposal as well as project to date
puts the quality of expertise and control into question,
which is a heavy burden on the acceptance of new projects.
And this leads to a second fundamental weakness of
nuclear waste management: Organization and safety
culture.
Safety culture: “Desiring to promote an effective
nuclear safety culture worldwide”
It is at the top of the list of objectives, as can be seen from
the preamble (V) of the IAEA Joint Convention 30 . But if
you then look for the concrete regulations, you will hardly
find anything regarding safety culture in the field of
geological waste repository planning processes. The
conception and planning seem to have escaped the
attention of a comprehensive supervisory process. Yet it is
precisely the concepts that are the fundamental guard rails
for safety, as the entire history of waste management of
highly toxic waste shows. The fact that not a single formal
overall review of the planning and implementation of
repositories to date has been carried out (Table 2) clearly
shows this deficit.
Industrial maturity
In this context, the questions relating to the long-term
safety of deep geological repositories can be asked in a far
more stringent manner. The statements made to date on
the long-term safety of these planned repositories over
periods of up to one million years are based exclusively on
calculations from a safety analysis known as a safety case 31 .
However, industrial experience and feasibility are rarely
included in these considerations. The reason is understandable,
as the IAEA correctly states in a publication
from 2012: “While the maturity criterion can be applied to
disposal facilities for radioactive waste, it has to be
recognized that data on the actual long term performance
of disposal facilities are not available” 32 . However, the
question of the industrial maturity of a plant is the
determining factor for the assessment of long-term safety.
This maturity process can only be achieved by a step- bystep
procedure and by knowledge and approach, based on
experiments and experience. As with any industrial
process, the development of a deep geological repository
requires a step-by-step approach that is divided into clear
stages and characterized by experimental validation. The
success of the planning process is therefore largely
determined by the quality and time dedicated to the
implementation of this process, which has a decisive
influence not only on the design of a deep geological
repository itself but also on the possibilities for corrective
action, as demonstrated, for example, by the current
difficulties encountered in retrieving the emplaced waste
from the Asse II experimental mine. It goes without saying
that such a planning process, until industrial maturity is
reached, also has an impact on the duration of interim
waste storage.
| Fig. 1.
EKRA-Concept.
An Inclusive Planning Approach
As seen above, the strategy for deep geological disposal of
radioactive waste is considered to be largely uncontested.
However, it is also undisputed that solutions for a deep
geological repository must be implemented at the highest
possible quality level and on a socially acceptable basis over
a long term. The first planning group to give these basic principles
the necessary comprehensive consideration was the
Expert Group on Disposal Concepts for Radioactive Waste
(EKRA), which was set up by the competent Swiss ministry.
In their first report published in 2000, they proposed a procedure
that not only followed this step-by-step philosophy
but also provided for the appropriate facilities to systematically
monitor the planning and implementation process 33 .
For this purpose, a phase of intensive experimental verification
of the site is planned as well as the construction of a socalled
pilot plant (Figure 1); the entire emplacement and
storage process is to be implemented and monitored with a
representative waste quantity, as long as there is a social consensus
on it. In a second report, EKRA later defined the
guidelines for the structural monitoring and governance
of the project 34 . EKRA was celebrated as a model of an
acceptance-building approach and was more or less fully
anchored in Switzerland’s new nuclear energy legislation.
The developments observed since then, with a
steady stream of new accidents, show that the current
planning for deep geological repositories does not meet
the requirements for a long-term safe planning process
and needs to be fundamentally improved. If one wants to
avoid similar developments as in the past, an inclusive
planning approach is required that considers the findings
from previous errors and problems:
p Without any doubt, the first improvement that is
needed is a safety culture that deserves this name, as
mentioned above, and which has to be a key element
during the most important phase of the process – the
conceptual design and planning phase.
p One has to recognize, that a top-down approach, as it
has been followed in all previous planning processes
for deep geological repositories, must be supplemented
by a bottom-up approach, which ensures that the
concerns of the regions and people directly affected are
considered. A simple right of co-determination in the
sense of con sultation processes, as practiced in the Swiss
sectoral plan procedure, is by no means sufficient to
ensure the necessary acceptance by the population. Trust
must also be established by subjecting security issues
to an assessment process by the population directly
FEATURE | ENVIRONMENT AND SAFETY 315
30 IAEA 1997: Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. Int. Atomic Energy Agency, Vienna.
31 for the development of the Safety Case: Pescatore, C., 2004, The Safety case, Concept, History and Purpose, Nuclear Energy Agency (OECD).
32 IAEA, 2012, The Safety Case and Safety Assessment for the Disposal of Radioactive Waste, Specific Safety Guide SSG-23.
33 EKRA, 2000, Disposal Concepts for Radioactive Waste, Final Report, 31st January 2000
34 EKRA, 2002, Contribution to the disposal strategy for radioactive waste in Switzerland, October 2000.
Feature
Deep Geological Radioactive and Chemical Waste Disposal: Where We Stand and Where We Go ı Marcos Buser, André Lambert and Walter Wildi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
FEATURE | ENVIRONMENT AND SAFETY 316
affected. This is also a central element in ensuring the
contemporary governance of such a long-term risk
project.
p The site selection and implementation process must
be carried out in clearly defined steps and must be
completed to industrial maturity. Even the best
project ideas are not sufficient and have to be
com plemented by an experiment based process that
can be implemented on an industrial scale. This applies,
for example, not only to the disposal of radioactive
waste at depth but also to industrial retrieval in the case
of undesirable developments, incidents or accidents. Of
course, the safety culture in these phases is again a key
process variable, as the recent example of the aviation
industry (Boing 737 MAX 8) impressively shows.
p The last of the central elements of the process is the
possible step back option: this is an essential condition
in this process of site selection and in the realization of
a deep geological repository. Corrections and returns
must always be possible in a process that promises
safety over 1 million years. The project must be
managed in a way that it can actually maintain this
extraordinarily high long-term safety benchmark.
References
ı Alley, W., Alley R. (2013) ‘Too Hot To Touch. The Problem of High-Level Nuclear Waste’, Cambridge
University Press.
ı Appel. D., Kreusch, J., Neumann, W., O.J., ‘Presentation of disposal options’, ENTRIA report 01
(first published 2014/2015).
ı Augustine, N. R., R. W. Mies, M. R. Anastasio, K. H. Donald, T. J. Glauthier, D. L. Hobson, G. B. Jaczko et al.
(2014) “A New Foundation for the Nuclear Enterprise: Report of the Congressional Advisory Panel on the
Governance of the Nuclear Security Enterprise.” November.
ı Beers, N. (1949) ‘Stack Meteorology and Atmospheric Disposal of Radioactive Waste’, Nucleonics 4/4,
April 1949, p. 28-38.
ı Beyer, F. (2005) Die (DDR-) History of the Morsleben Nuclear Waste Repository. “Contributions in Kind”,
No. 36, Magdeburg 2005.
ı Blum, P., Goldscheider, N., Göppert, N., Kaufmann-Knoke, R. et al., (2016) Groundwater – human ecosystems,
25th conference of the FH-DGGV, Karlsruhe 13th-16th April 2016, KIT Scientific Publishing, p. 152.
ı Boffey, P. (1975) The Bain Bank of America, McGraw Hill.
ı Bogorov, V. G., Tareev, B. A., Fedorov, K.N., (1959) ‘The Depths of the Ocean and the Question of
Radioactive Waste Disposal in them’, in Conference Proceedings Monaco 16.-21. November 1959, Disposal
of Radioactive Wastes, Vol. 2.
ı Bostrom, R.C., Sherif, M. A., (1970) ‘Disposal of Waste Material in Tectonic Sinks’, Nature 228, p. 154-156.
ı Browder, F. N., (1951) ‘Liquid Waste Disposal at Oak Ridge National Laboratory’, Industrial and
Engineering Chemistry, July 1951.
ı Bürgisser, H., et al., (1979) ‘Geological aspects of radioactive waste disposal in Switzerland’, Switzerland.
Energy foundation.
ı Buser, M. (2019) ‘Where to go with Nuclear Waste’, Rotpunkt Verlag Zurich.
ı Buser, M. (2019) ‘Repositories for low and intermediate level radioactive waste in Sweden and Finland’, a
travel report, Nuclearwaste.info.
ı Cecille, L., Hage, W., Hettinger H., Mannone, F. (1977) ‘Nuclear Transmutation of Actinides Other than
Fuel as a Radioactive Waste Management Scheme, IAEA-CN-36/366.
ı Claus, W. (1955) Fundamental Considerations on the Disposal of Large Amounts of Radioactive Waste in
the Land and Sea, United Nations, Proceedings of the International Conference on the Peaceful Use of
Atomic Energy, Held in Geneva from 8 to 20 August 1955, Volume LX, 1956
ı Copil, (2011) ‘ Expert report’, Steering committee, June 2011.
ı de Laguna, W., Cowser, K.E., Parker, F.L. (1958) ‘Disposal of High-Level Liquid Waste in Seepage ponds:
new data’, Proceedings of the Second United Nations International Conference on the Peaceful Use of
Atomic Energy , Held in Geneva from September 1 to 13, 1958.
ı DOE (2014a) “Accident Investigation Report, Underground Salt Haul Truck Fire at the Waste Isolation Pilot
Plant, February 5, 2014.” Office of Environmental Management, Department of Energy.
ı DOE (2014b) “Accident Investigation Report, Phase 1: Radiological Release Event at the Waste Isolation
Pilot Plant, February 14, 2014.” Office of Environmental Management, Department of Energy.
ı DOE (2015) “Accident Investigation Report, Phase 2: Radiological Release Event at the Waste Isolation
Pilot Plant, February 14, 2014.” Office of Environmental Management, Department of Energy.
ı EKRA (2000) ‘Disposal Concepts for Radioactive Waste’, Final Report, 31st January 2000.
ı EKRA (2002) ‘Contribution to the Disposal Strategy for Radioactive Waste in Switzerland’, October 2000.
ı ERDA (1977) ‘Management of Intermediate Level Radioactive Waste’, US Energy Research and
Development Administration.
ı Strata (1987) Emptying the Beta mine, V, p. 8-11.
ı Evans, J. E. (1952), Disposal of Active Wastes at Sea, NSA, Vol. 8 1954.
ı Ford, D (1982), ‘The Cult of the Atom. The Secret Papers of the Atomic Energy Commission’, Simon &
Schuster.
ı Gilmore, W.R. (1977) ‘Radioactive Waste Disposal. Low and High Level’, Noyes Data Corporation.
ı Ginell, W. S., Martin, J.J., Hatch, L. P. (1954) ‘Ultimate Disposal of Radioactive Wastes’, Nucleonics 12/12,
December 1954.
ı Glückauf, E. (1955) ‘The long-term Problem of Radioactive Waste Disposal’, Proceedings of the
International Conference on the Peaceful Use of Atomic Energy, Held in Geneva from August 8 to 20,
1955, volume IX, IAEA, 1956.
ı Goodmann, C. (1949) ‘Future Developments in Nuclear Energy’, Nucleonics 4/2, February 1949, p. 2-16.
ı Hatch, L. P. (1953) ‘Ultimate Disposal of Radioactive Waste’, American Scientist Vol. 41/3, p. 410-412.
ı Herrington, A.C., Shaver, R.G., Sorenson, C.W. (1953) ‘Permanent Disposal of Radioactive Wastes:
Economic Evaluation’, Nucleonics 11/9, September 1953, p. 34-37.
ı Hollister, C.D. (1977), ‘The Seabed Option’, Oceanus 20, p. 18-25;
ı Hollister, Ch., Anderson, D. R., Health, G. R. (1981) ‘Subseabed Disposal of Nuclear Wastes’, Science,
Vol. 213, 18 Sep 1981.
ı Hollocher, Thomas C. (1975) ‘Storage and Disposal of High-Level Radioactive Waste, in: Union of
Concerned Scientists, The nuclear Fuel Cycle’, The MIT Press, Cambridge, Massachusetts.
ı Ialenti, V. (2018) ‘Waste Makes Haste: How a Campaign to Speed up Nuclear Waste Shipments Shut
down the WIPP Long-Term Repository.’ Bulletin of the Atomic Scientists 74 (4): 262–275. doi:10.1080/009
63402.2018.1486616.
ı Ibsen, D., Kost, S., Weichler, H. (2010) ‘Analysis of the usage history and the planning and participation
forms of the Asse II mine’, final report AEP, University of Kassel.
ı IAEA (1977) Radioactive Waste - Where From, Where To?, International Atomic Energy Agency.
ı IAEA (1979) ‘Geologic Disposal of Nuclear Waste’, Conf. papers CONF-790304 - DE82 902335 ,March 15
ans 16, 1979, https://inis.iaea.org/collection/NCLCollectionStore/_Public/13/684/
13684877.pdf?r=1&r=1.
ı IAEA (2012) ‘The Safety Case and Safety Assessment for the Disposal of Radioactive Waste’, Specific
Safety Guide SSG-23.
ı INSAG (1991), Safety Culture, Safety Series No 75-INSAG-4, International Nuclear Safety Advisory Group,
IAEA.
ı KASAM (1999) ‘Retrievability of high-level waste and spent nuclear fuel’, Proceedings of an international
seminar organized by the Swedish National Council for Nuclear Waste in co-operation with the
International Atomic Energy Agency And held in Saltsjöbaden, Sweden, 24–27 October 1999,
IAEA-TECDOC-1187, https://www-pub.iaea.org/MTCD/publications/PDF/te_1187_prn.pdf.
ı Klaus, D.M. (2019) ‘What really went wrong at WIPP: An insider’s view of two accidents at the only US
underground nuclear waste repository’, Bulletin of the Atomic Scientists, 75:4, 28 June 2019.
ı KBS (1978a) Handling of spent fuel and final storage of vitrified high-level reprocessing waste,
Kärnbränslesäkerhet.
ı KBS (1978b) Handling and Final Storage of Unreprocessed Spent Nuclear Fuel, Kärnbränslesäkerhet.
ı López Pérez, B., Martínez Martínez, A. (1976) ‘Experience Gained and Technology Developed at the JEN
in the management of radioactive waste’, IAEA SM-207/90.
ı Milnes, A.G., Buser, M., Wildi, W. (1980) ‘Overview of Repository Concepts for Radioactive Waste’ –
Z. dtsch. Geol. Ges. 131, 359-385.
ı Milnes, A.G. (1985) Geology and Radwaste, Academic Press.
ı Möller, D (2009) ‘Final disposal of radioactive waste in the Federal Republic’. Peter Lang.
ı NAS, National Academy of Sciences (1957a) ‘The Disposal of Radioactive Wastes on Land. Report of the
Committee on Waste Disposal of the Division of the Earth Sciences’, National Research Council.
ı NDC (1977) ‘Radioactive Waste Disposal, Low- and High-Level’, Pollution Technology Review 38.
ı ORNL (1985) ‘The Management of Radioactive Waste at the Oak Ridge National Laboratory’, National
Academy Press, p. 124-125.
ı Pescatore, C. (2004) ‘The Safety case, Concept, History and Purpose’, Nuclear Energy Agency (OECD).
ı Philbert, B. (1959) ‘Removal of radioactive waste substances in the ice caps of the earth’, Nuclear Energy 4/3.
ı Renn, Ch., (1955) ‘The Dumping of Radioactive Waste’, Proceedings of the International Conference
on the Peaceful uses of Atomic Energy, Held in Geneva from August 8 to 20, 1955, Volume LX, 1956.
ı Rodgers W. A. (1954) ‘Radioactive Wastes - Treatment, Use, Disposal’, Nuclear Engineering, 50/5,
January 1954, p. 263-266.
ı Schneider, L., Herzog, Ch., (2011) ‘Sites and projects for the disposal of radioactive waste and repositories
in Russia and other states of the former USSR’, Stoller Engineering 2011.
ı Scott, K., (1950) ‘Radioactive Waste Disposal - How Will It Affect Man’s Economy’, Nucleonics, Vol. 6/1,
p. 15-25.
ı Spitsyn, V. I., Balukoda, V.D. (1978) The Scientific Basis For, and Experience With, Underground Storage of
Liquid Radioactive Wastes in the USSR, in Scientific Basis for Nuclear Waste Management, Springer,
pp. 237-248.
ı Stow, S. H., Haase, S. C. (1986) ‘Subsurface disposal of liquid low-level radioactive wastes at Oak Ridge’,
Tennessee, Oak Ridge National Laboratory P.0.Box, Oak Ridge, Tennessee 37831.
ı Strauss, Lewis, (1954) ‘Remarks for the Delivery at The Founder’s Day Dinner’, National Association of
Science Writers, New York, 16. September 1954, Atomic Energy Commission.
ı Theis, Ch. (1955) ‘Problems relating to the burial of nuclear waste’, Proceedings of the International
Conference on the Peaceful Use of Atomic Energy, Held in Geneva from August 8 to 20, 1955, Volume IX,
1956.
ı Walker, Samuel jr. (2006) ‘An ‘Atomic Garbage Dump’ for Kansas, Kansas History’: A Journal of the Central
Plains 27 (Winter 2006-2007), p. 266-285.
ı Walker, S.J. (2009) ‘The Road to Yucca Mountain’, University of California Press London.
ı Western F (1948) ‘Problems of Radioactive Waste Disposal’, Nucleonics 3/2, August 1949, p. 43-49.
ı WNA, (2020) ‘Storage and Disposal of Radioactive Waste’ World Nuclear Association, November 2020.
ı WNWR (2019) ‘The World Nuclear Waste Report’, Focus Europe, Heinrich Böll Foundation and Partners.
ı Woller, F. (2008) ‘Disposal of radioactive waste in rock-caverns: Current situation in Czech Republic’, in
Rempe, N. T., 2008, Deep Geological Repositories, The Geological Society of America.
Authors
Marcos Buser
marcos.buser@bluewin.ch
Funkackerstrasse 19
8050 Zürich, Switzerland
André Lambert
Ziegelhaustrasse 19
5400 Baden, Switzerland
Walter Wildi
Département F.A. Forel des sciences de l’environnement
et de l’eau
University of Geneva
Chemin des Marais 23
1218 Le Grand Saconnex, Switzerland
Feature
Deep Geological Radioactive and Chemical Waste Disposal: Where We Stand and Where We Go ı Marcos Buser, André Lambert and Walter Wildi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Did you know...?
Report of Bundesnetzagentur (BNetzA) on Reserve Power Plant
Requirements Winter 2020/21 and Years 2024/25
The Bundesnetzagentur (Federal Network Agency for Electricity, Gas,
Telecommunications, Post and Railway) regulating the electrical, the
gas, the railway and telecommunication grids and supervising
among other German TSOs released its current report “Reserve
Power Plant Requirements for the winter 2020/21 and the years
2024/25” in May 2020. This type of reporting started after the
political decision to accelerate the nuclear phase-out in 2011 starting
with the immediate permanent shut-down of eight NPPs with
approximately 8 GW of capacity. Before 2011 no systematic assessment
of this kind was performed for Germany because the conventional
generation capacity was dimensioned rather generously.
The report includes the preview for the upcoming winter, a T+xanalysis
for 2024/25 and the notification of actually contracted
reserve capacity in previous periods. Graph 1 combines these
backward and forward looking data from 2011 to 2025 and shows a
very clear link between the steps of nuclear phase-out and reserve
power plant requirements. The escalating requirements till 2018
lead to the abolition of the common bidding zone of Germany and
Austria in the electricity market in October 2018, which lead to a
significant preliminary reduction in requirements afterwards.
The cost of major system stability measures (graph 2) shows the link
to nuclear phase-out too, but here the annual amount of the volatile
renewable energies wind and solar, the measure of their volatility
and the implementation of grid extension measures such as the
opening of a new AC power line from the north to the region of the
NPP Grafenrheinfeld in two steps at the end of 2015 and in the
summer of 2016 also play a major role. The grid extension necessary
to reasonably accommodate nuclear phase-out and current levels of
renewable generation is planned to be completed by the end of
2025, nuclear phase-out will be completed by the end of 2022. Of
the current reserve power plant requirements estimated by the
Federal Network Agency for 2022/23 of 10,647 MW and 8,042 MW
for 2024/25 only 6,930 MW and 5,970 MW respectively are currently
judged to be potentially disposable domestically. The generation
capacity of conventional, adjustable power plants in the German
electricity market is supposed to shrink from 91.4 GW in 2020/21 to
71.1 GW in 2024/25 due mostly to nuclear phase-out and the first
steps of coal phase-out currently being legislated.
For further details
please contact:
Nicolas Wendler
KernD
Robert-Koch-Platz 4
10115 Berlin
Germany
E-mail: presse@
KernD.de
www.KernD.de
317
DID YOU KNOW...?
Totalized Capacity of Domestic and International Grid Reserve Power Plants and
Identified Requirements for the Winters/Years (in MW)
(Contracted Capacity, Winter Reports, Reports T+x)
12,000 p Domestic p International p Sum
10,000
8,000
6,000
7,660
8,383
11,430
6,598 6,598 6,596
10,647
8,042
4,000
2,000
1,472
2,559
2,945 3,024
0
2011/12 2012/13 2013/14 2014/15 2015/16 2016/17 2017/18 2018/19 2019/20 2020/21 2022/23 2024/25
Start of analysis of Reserve
Power Plant Requirements
following accelerated
Phase-out of Nuclear Power
Shut-down
of NPP
Grafenrheinfeld,
27.06.2015
Shut-down
of NPP
Gundremmingen B,
31.12.2017
Abolition of
the common
bidding zone
Germany-
Austria
Shut-down
of NPP
Philippsburg 2,
31.12.2019
Scheduled Shut-down of NPPs
Brokdorf, Grohnde, Gundremmingen C,
31.12.2021 and
Emsland, Isar 2, Neckarwestheim 2,
31.12.2022
800
600
400
200
0
Preliminary Costs of major System Stability Measures in million Euro
1,600 p Redispatch (TSO)
p Countertrading (TSO)
1,400 p Feed-in Management (DSO und TSO)
p Grid Reserve (domestic)
1,200 p Grid Reserve (international)
p Sum
1,000
179.1
223.7 214.7
436.1
1,141.4
2011 2012 2013 2014 2015 2016 2017 2018
893.0
1,513.8
1,436.6
Source:
Bundesnetzagentur,
Bericht Feststellung
des Bedarfs an Netzreserve
für den Winter
2020/21 sowie das
Jahr 2024/25; Bericht
Feststellung des
Bedarfs an Netzreserve
für den Winter
2019/2020 sowie das
Jahr 2022/2023
Did you know...?

atw Vol. 65 (2020) | Issue 6/7 ı June/July
318
SPOTLIGHT ON NUCLEAR LAW
Kein „Stillstand der Rechtspflege“ in Coronazeiten:
Beginn des digitalen Zeitalters im Verwaltungsverfahrensrecht
mit dem neuen Planungssicherstellungsgesetz?
Ulrike Feldmann
A Einleitung In Zeiten der hohen gesundheitlichen Gefährdung durch die COVID-19-Pandemie und den zur Eindämmung dieser Pandemie
erforderlichen Beschränkungen ergeben sich vielfältige praktische und rechtliche Probleme, wie allein die vom Deutschen Richterbund geschätzten
rund 1000 gerichtlich anhängig gemachten Eilverfahren zu den Beschränkungen sowie die Gründung einer eigens den Rechtsfragen zur Corona-
Krise gewidmeten Zeitschrift „COVuR“ belegen. Unter Anderem führen die geltenden Veranstaltungs- und Kontaktbeschränkungen zu Umsetzungsproblemen
bei der Durchführung von Verwaltungsverfahren. Als problematisch erweisen sich insbesondere die öffentliche Aus legung von
Antragsunterlagen und die Durchführung von – verpflichtend vorgeschriebenen – Erörterungs terminen (z. B. nach UVPG). Viele Gemeindeverwaltungen,
in denn die öffentliche Auslegung stattfinden müsste, sind aufgrund der Kontaktbeschränkungen gesperrt worden. Die Bekanntgabe
von Zulassungsentscheidungen, für die eine öffentliche Auslegung vorge schrieben ist, ist nicht mehr möglich. Personalverknappung z. B. durch die
Zugehörigkeit von Personal zu Risikogruppen oder aufgrund notwendiger Kinder betreuung kann zu einem zusätzlichen Problem werden.
Um gleichwohl wichtige Planungs- und Genehmigungsverfahren (insbesondere
im Wohnungsbau sowie auf dem Energie-, Verkehrs- und
Klimaschutzsektor) nicht auf unbestimmte Zeit verschieben zu müssen
und Vorhabenträgern Planungs- und Verfahrenssicherheit zu geben,
stimmte der Deutsche Bundestag am 14.05.2020 dem Entwurf der Koalitionsfraktionen
für ein „Planungssicherstellungsgesetz“ (BT-Drucksache
19/18965) in der Fassung der Beschlussempfehlung des Bundestagsausschusses
für Inneres und Heimat (BT-Drucksache 19/19214) mit den
Stimmen der Koalitionsfraktionen zu. Bereits einen Tag später erteilte
der Bundesrat ohne Aussprache dem Gesetz ebenfalls seine Zustimmung.
Manche Verbände hätten zu dem Gesetzentwurf wohl gerne etwas
mehr gesagt, hatten dazu aber – wenn überhaupt – lediglich ein äußerst
knapp bemessenes Arbeitszeitfenster zwischen Freitagmittag (27.04.)
und dem folgenden Montagmittag zur Verfügung.
B
Zum Inhalt des Gesetzes
I Anwendungsbereich
Das Planungssicherstellungsgesetz (PlanSiG), das in seiner Langform
den etwas sperrigen Titel „Gesetz zur Sicher stellung ordnungs gemäßer
Planungs- und Genehmigungsverfahren während der COVID-19-
Pandemie“ trägt, gilt einheitlich für Verwaltungsverfahren nach den in
§ 1 PlanSiG abschließend genannten 23 Fachgesetze, u. a. auch für das
Atomgesetz (§ 1 Nr. 7 PlanSiG) und das Strahlenschutzgesetz (§ 1 Nr. 8
PlanSiG).
Das PlanSiG soll gewährleisten, dass Planungs- und Genehmigungsverfahren
sowie besondere Entscheidungsverfahren mit Öffentlichkeitsbeteiligung
auch unter den erschwerten Bedingungen während
der COVID-19- Pandemie ordnungsgemäß durchgeführt werden
können. Das Gesetz sieht für diese Verfahren, für die nach bis herigem
Recht Verfahrensberechtigte zur Wahrnehmung ihrer Beteiligungs -
rechte physisch präsent sein müssen und bei denen häufig diese Rechte
von einer Vielzahl von Verfahrensberechtigten ausgeübt werden, formwahrende
Alternativen vor.
II Ortsübliche und öffentliche Bekannt machungen
(§ 2 PlanSiG)
Für die ortsübliche oder öffentliche Bekanntmachung von Unter lagen
und anderen Informationen (z. B. durch Aus legung zur Einsichtnahme)
sieht das Gesetz alternativ die Veröffentlichung der Unterlagen und
Informationen im Internet vor (§ 2 Abs. 1 S. 1 PlanSiG). Für befristete
Bekanntmachungen gilt dies nur, wenn die Auslegungsfrist spätestens
mit Ablauf des 31. März 2021 endet, da zu diesem Zeitpunkt gem. § 7
Abs. 2 S.1 PlanSiG die §§ 1-5 PlanSiG außer Kraft treten. Um auch
potentielle Verfahrensberechtigte zu erreichen, die keinen Internetzugang
besitzen, lässt das Gesetz die bisherige – analoge – Bekanntmachung
in einem amtlichen Veröffentlichungsblatt oder einer
örtlichen Tages zeitung unberührt (§ 2 Abs. 1 S. 2 PlanSiG).
III Auslegung von Unterlagen oder Entscheidungen
(§ 3 PlanSiG)
1. Ähnlich verhält es sich bei Verfahren, in denen eine Auslegung von
Unterlagen oder Entscheidungen angeordnet ist. Kann auf die
Aus legung nicht verzichtet werden, können die Unterlagen im
Internet veröffentlicht werden, sofern die jeweilige Auslegungsfrist
spätestens mit Ablauf des 31.03.2021 endet (§ 3 Abs. 1 S. 1 PlanSiG).
In der Bekanntmachung ist auf die Veröffentlichung im Internet
genau hinzuweisen. Neben der Veröffentlichung im Internet soll
gemäß § 3 Abs. 2 S. 1 PlanSiG auch die ursprünglich angeordnete
analoge Auslegung erfolgen, soweit zuvor die Behörde festgestellt
hat, dass dies den Umständen nach möglich ist. Kommt die Behörde
zu dem Ergebnis, dass die – analoge – Auslegung nicht möglich ist,
hat die Behörde im Interesse der Bürger ohne Internetzugang additiv
zur digitalen Veröffent lichung eine andere Möglichkeit zur Verfügung
zu stellen.
2. In Bezug auf den gemeinsamen Referentenentwurf von BMU und
BMI war industrieseitig zu Recht besonders die fehlende
Berücksichtigung des geltenden Schutzes von Betriebs- und
Geschäftsgeheimnissen kritisiert worden, zumal mit der Veröffentlichung
im Internet die Informationen einem potentiell weltweit
unbegrenzten Personenkreis frei zugänglich gemacht würden,
Sicherheitsrisiken damit nicht ausgeschlossen seien und außerdem
fraglich sei, wie nach Ablauf der Anhörungsfrist die Löschung der
Informationen aus dem Internet erfolgen solle. Diesen Bedenken
trägt die o. g. und in das verabschiedete Gesetz eingeflossene
Beschlussempfehlung des Bundestagsinnenausschusses Rechnung:
Der Vorhabenträger hat nunmehr einen Anspruch darauf, dass seine
Betriebs- und Geschäftsgeheimnisse von der Behörde nicht unbefugt
offenbart werden. Er kann der Veröffentlichung im Internet widersprechen.
Macht er jedoch von seinem Widerspruchsrecht Gebrauch,
hat dies zur Konsequenz, dass die Behörde das Verfahren bis zur Auslegung
der Unterlagen aussetzen muss (§ 3 Abs. 1 Sätze 5-7 PlanSiG).
Ein beschleunigtes Verfahren ist demnach nur möglich, wenn
Transparenz gewährt wird.
3. Hält eine Behörde eine „analoge“ Aus legung neben der digitalen
Veröffentlichung nicht für möglich (s. o.), hat die Behörde zusätzlich
zur digitalen Veröffentlichung andere leicht zu erreichende Zugangsmöglichkeiten
zu schaffen. § 3 Abs. 2 S. 2 PlanSiG nennt als Beispiele
öffentlich zugängliche Lesegeräte oder „in begründeten Fällen“ die
Versendung per Post. Allerdings können erfahrungsgemäß nicht nur
in atomrecht lichen Genehmigungsverfahren der Aktenumfang und
der Kopieraufwand recht gewaltig werden. Dessen eingedenk ist in
die vom Bundes kabinett beschlossene „Formulierungshilfe“ für den
Gesetzentwurf der Fraktionen der CDU/CSU und SPD in der Begründung
zu § 3 Abs. 2 S. 2 PlanSiG der Satz angefügt worden:“ Eine
Versendung von Unterlagen mit der Post kann sich z. B. bei einem
kleinen Adressatenkreis anbieten“.
IV Erklärungen zur Niederschrift
§ 4 PlanSiG erfasst die Fallkonstellationen, in denen die Abgabe von
Erklärungen zur Niederschrift vorgesehen sind. Derartige Abgaben
kann die zuständige Behörde ausschließen, sofern die Erklärungsfrist
spätestens mit Ablauf des 31. März 2021 endet. Stattdessen muss die
Behörde einen Zugang für die Abgabe elektronischer Erklärungen (z. B.
per E-Mail) vorsehen und muss in der üblichen Bekanntmachung auf
den Zugang hinweisen.
Spotlight on Nuclear Law
No “Standstill in the Administration of Justice” in Corona Times ı Ulrike Feldmann

atw Vol. 65 (2020) | Issue 6/7 ı June/July
V Erörterungstermin, mündliche Verhandlungen und
Antragskonferenzen
Neben § 3 PlanSiG dürfte insbesondere § 5 PlanSiG von großem
Interesse für die anstehenden atomrechtlichen Stilllegungsverfahren
sein.
1. Ist gesetzlich ein behördliches Ermessen darüber eingeräumt, ob ein
Erörterungstermin durchgeführt wird, so kann die Behörde bei ihrer
Ermessensausübung auch mit der aufgrund der COVID-19-Pandemie
verbundene Probleme (z. B. geltende Kontakt beschränkungen und
Abstandsregeln) berücksichtigen und damit auf einen Erörterungstermin
oder die mündliche Verhandlung verzichten (s. § 5 Abs. 1
PlanSiG).
2. Ist dagegen gesetzlich die Durchführung eines Erörterungstermins
oder eine mündliche Verhandlung angeordnet und ein Verzicht nicht
möglich, genügt eine Online-Konsultation (§ 5 Abs. 2 PlanSiG).
3. Die Bekanntmachung der behördlichen Entscheidung zur Durchführung
einer Online-Konsultation ist unter Hinweis auf § 73 Abs. 6
Sätze 2-4 VwVfG in § 5 Abs. 3 PlanSiG geregelt. Die Online-
Konsultation dient als Ersatz für die sonst üblichen mündlichen
Stellung nahmen und Gegenstellungnahmen.
4. Der nicht-öffentliche Charakter von Erörterungs terminen oder
mündlichen Verhandlungen spiegelt sich in der Online- Konsultation
dadurch wider, dass nur den zur Online-Teilnahme Berechtigten die
entsprechenden Informationen zur Verfügung zu stellen sind und
nur ihnen wie sonst auch innerhalb einer festgesetzten Frist Gelegenheit
zur schriftlichen (auch elektro nischen) Stellungnahme zu geben
ist. Klar gestellt wird, dass damit jedoch keine neuen, zusätz lichen
Einwendungsmöglich keiten geschaffen werden (§ 5 Abs. 4 PlanSiG).
5. § 5 Abs. 5 PlanSiG sieht die Möglichkeit vor, die Online-Konsultation
auch als Telefon-oder Videokonferenz durchzuführen. Diese
Möglichkeit war in dem ursprünglichen BMU/BMI-Referentenentwurf
nicht enthalten. Da sie vom Einverständnis der zur Teilnahme
Berechtigten abhängt, ist offen, wie oft und ob überhaupt
diese Möglichkeit zur Anwendung kommen wird.
6. Im Falle einer gesetzlich vorgesehenen Antrags konferenz sieht § 5
Abs. 6 PlanSiG ein vereinfachtes Verfahren gegenüber einer
Online-Konferenz nach § 5 Abs. IV PlanSiG vor.
7. In der Entstehungsgeschichte des Gesetzes gibt es keinen Hinweis
darauf, dass der Gesetzgeber sich über die Frage vertieft Gedanken
gemacht hat, ob der Vorhabenträger bei Vorliegen von COVID-19
bedingten Pro blemen bei der Durchführung von Verwaltungsverfahren
ebenfalls einen Anspruch auf Anwendung der im PlanSiG
geregelten formwahrenden Alter nativen haben soll. In den §§ 2-5
PlanSiG finden sich eine Reihe von „Kann“-Bestimmungen, die
„klassischer weise“ für eine Ermessensentscheidung sprechen. Dabei
muss für jede Vorschrift individuell geprüft werden, ob sie einen
Ermessensspielraum eröffnet. Dies liegt dann nahe, wenn wie z. B. in
den §§ 2 bis 4 und 5 Abs. 1 PlanSiG ausdrücklich „Kann“-Bestimmungen
gewählt wurden, und eher fern, wenn darauf wie z. B. in § 5
Abs. 2 PlanSiG verzichtet wurde. In jedem Fall ist zu bedenken, dass
der Zweck des Gesetzes (Gewähr leistung der ordnungsgemäßen
Durchführung von Planungs- und Genehmigungsverfahren sowie
besonderen Entscheidungsver fahren mit Öffentlichkeits beteiligung
auch unter den er schwerten Bedingungen während der COVID-
19-Pandemie; keine Ver schiebung dieser Verfahren auf unbestimmte
Zeit; Planungs- und Verfahrenssicherheit für Vorhabenträger) ins
Leere liefe, wenn die zuständigen Behörden ihr Ermessen
dahingehend ausübten, dass sie von den alternativen Verfahrensmöglichkeiten,
die das PlanSiG bereit stellt, keinen Gebrauch
machten. Deshalb jedoch grund sätzlich bereits von einer „Ermessensreduzierung
auf Null“ auszugehen, würde der ungewissen
Entwicklung der Pandemie-Situation und einer entsprechend
erforderlichen Anpassung an den Umfang der Beschränkungen
einerseits und den unterschiedlichen Gegebenheiten in den mit den
jeweiligen Verfahren befassten Kommunen andererseits nicht
genügend Rechnung tragen. Soweit Ermessensentscheidungen
eingeräumt sind, muss allerdings berücksichtigt werden, dass das
PlanSiG zumindest auch im drittschützenden Interesse des Vorhabenträgers
besteht. Ein Außerachtlassen der Möglichkeiten des
PlanSiG und einer daraus folgenden Verfahrensverzögerung wäre
daher in jedem Fall besonders rechtfertigungs bedürftig.
VI Übergangsregelung
1. Beachtenswert ist ebenfalls die in § 6 PlanSiG vorge sehene Übergangsregelung.
Um den Wirkungsbereich des Gesetzes aus reichend
praxistauglich zu gestalten, sollen auch begonnene Planungs- und
Genehmigungsverfahren an den Verfahrenserleichterungen des
PlanSiG teilhaben können. Entscheidet sich die Behörde, einen
bereits „analog“, d. h. nach bisher geltendem Recht begonnenen
Verfahrensschritt nach dem PlanSiG durch zuführen, ist dieser
betreffende Verfahrensschritt nach diesem Gesetz zu wiederholen
(§ 6 Abs. 1 S. 2 PlanSiG). Diese Regelung erlaubt es der Behörde,
jeden nach „analogem“ Verfahrensrecht begonnenen Ver fahrensschritt
je nach Sachlage individuell im Hinblick auf eine Wiederholung
nach PlanSiG zu betrachten. Eine Ausnahme von dem Grundsatz
in § 6 Abs. 1 S. 2 PlanSiG findet sich in § 6 Abs. 1 S. 3 PlanSiG.
2. Für unter diesem Gesetz durchgeführte Verfahren, die mit Ablauf des
31.03.2021, dem Datum des Außerkrafttretens der §§ 1-5 PlanSiG,
noch nicht beendet sind, fingiert § 6 Abs. 2 PlanSiG die Fortgeltung
der Bestimmungen bis zum Ende des jeweiligen Ver fahrensschrittes.
3. Im Hinblick auf Fehlerfolgen gelten gem. § 6 Abs. 3 PlanSiG die
einschlägigen Regelungen in den jeweiligen Fachgesetzen. Diese
Regelungen sind daneben auch auf Verstöße gegen die §§ 2-5 PlanSiG
anzuwenden.
C Fazit und Ausblick
Am 28.05.2020 ist das PlanSiG im Bundesgesetzblatt verkündet worden
und am 29.05.2020 in Kraft getreten. Auch wenn das PlanSiG unter
dem Druck der COVID-19-Pandemiesituation entstanden ist, um die
zügige Umsetzung wichtiger, auch im öffentlichen Interesse liegender
Vorhaben unter den herrschenden Pandemie bedingungen rechtssicher
zu gewährleisten, ist den Urhebern des PlanSiG anzukreiden, dass es
durch das Gesetzgebungsverfahren mit größter Eile durchgetrieben
wurde, Verbänden kaum Zeit für Stellungnahmen blieb und unklar
bleibt, ob die jeweilige Behörde im Einzelfall von den digitalen Möglichkeiten,
die das PlanSiG bietet, Gebrauch machen wird, und ob in allen
Verfahren und unter allen 23 im PlanSiG erfassten Fachgebieten
dieselben Maßstäbe angelegt werden. Der Anspruch des Gesetzes,
Planungs- und Verfahrens sicherheit zu geben, wird damit nicht recht
erfüllt.
Mit dem richtigen Augenmaß angewandt und den Gesetzeszweck
im Auge behaltend gibt das PlanSiG aber immerhin Anlass für die
Hoffnung auf Praxistauglichkeit, ohne dass dem Gesetz zum Vorwurf
gemacht werden kann, dass es die Wahrnehmung von Verfahrensrechten
ungebührlich beschneidet. Dies wird allerdings unter anderem
von Bürgerinitiativen gegen die geplanten Stromtrassen bestritten, die
bundesweit für den 24.05.2020 zu einer Protestaktion gegen das
PlanSiG aufgerufen hatten. Mit weiteren Protestaktionen dürfte zu
rechnen sein.
Ohne die COVID-19-Pandemiekrise hätte sicherlich der verwaltungsverfahrensrechtliche
Sprung in das digitale Zeitalter noch
eine Reihe von Jahren auf sich warten lassen. Wie den Ausführungen
von Philipp Amthor und Konstantin Kuhle in der Bundestagsdebatte zur
2. und 3. Lesung zum PlanSiG zu entnehmen ist (Plenarprotokoll v.
14.05.2020), soll das Gesetz nach seinem Ablauf evaluiert werden, um
die Digitalisierung des Verwaltungsverfahrensrechts möglichst auch
nach dem Ende der Corona-Krise fortsetzen zu können. Versuch macht
bekanntlich klug, und man wird zum Zeitpunkt des Außerkrafttretens
des Gesetzes wissen, ob mit dem PlanSiG ein „ Löwe“ sprang, der auch
als „Löwe“ und nicht als „Bettvorleger“ landete.
Autor
Ulrike Feldmann
Justitiarin
Kerntechnik Deutschland e.V. (KernD)
Robert-Koch-Platz 4
10115 Berlin
SPOTLIGHT ON NUCLEAR LAW 319
Spotlight on Nuclear Law
No “Standstill in the Administration of Justice” in Corona Times ı Ulrike Feldmann

atw Vol. 65 (2020) | Issue 6/7 ı June/July
320
How Final Disposal Can Work
Nicole Koch
ENVIRONMENT AND SAFETY
A huge tunnel system is emerging in Finland that could solve one of the greatest problems facing mankind:
the disposal of high level nuclear waste.
A look Back. The first preparations
for final disposal already began in
the 1980s. Teollisuuden Voima carried
out some research related to
the final disposal in the 1980s and
early 1990s, but Imatran Voima (currently,
Fortum) transported its spent
nuclear fuel to the Soviet Union or
Russia. In 1994, the Nuclear Energy
Act entered into force, according
to which all nuclear waste must be
treated, stored and disposed of in
Finland, and no nuclear waste from
other countries shall be imported into
Finland. After this, Imatran Voima
and Teollisuuden Voima established
Posiva Oy to take care of the implementation
of the final disposal of
spent nuclear fuel and the associated
research.
Research associated with the final
disposal proceeded as follows:
p 1983 to 1985: Screening study of
the entire area of Finland
p 1986 to 1992: Preliminary site
investigations
p 1993 to 2000: Detailed site investigations
and an environmental
impact assessment procedure
was carried out for four sites:
Romuvaara in Kuhmo, Kivetty in
Äänekoski, Olkiluoto in Eurajoki,
and Hästholmen in Loviisa.
According to the site investigations
and safety analyses, as well as the
environmental impact assessment
procedure, all the investigated sites
would have been suitable for the final
disposal of spent nuclear fuel. The
local consent was highest in Eurajoki
and Loviisa. Of these two, the
Olkiluoto island in Eurajoki had a
larger area reserved for the repository.
Furthermore, the larger portion of the
spent nuclear fuel was already on the
island. In 2000, the Olkiluoto island
in Eurajoki was selected as the site for
final disposal.
The construction licence application
for the repository was submitted
in 2012. Construction licence was
granted in November 2015 and the
operation licence application will be
submitted in 2020. The final disposal
is scheduled to start in the 2020’s.
According to current plans the repository
would be sealed up by the
2120’s.
Facts & Figures STUK
STUK was established in 1958 and operated under the Medical
Administration as the Department of Radiation Physics with the task
of inspecting radiation sources used in hospitals
p In 2018, STUK’s operating expenses were 39 million EUR
p 12.3 million of the funding came from the taxpayers
via the government budget
p 21 million of the operating expenses was collected by STUK
as regulatory oversight fees
p …of which 17.7 million came from the regulatory oversight
of nuclear energy use
p At the end of 2019 STUK had 353 employees
Jussi Heinonen is the director of the
nuclear waste and material regulation
department at STUK (the Finnish
Radiation and Nuclear Safety Authority).
He studied material sciences at
Helsinki Uni versity of Technology,
has 18 years experience in STUK
and has been closely involved in the
licensing process and construction
of the Olkiluoto spent nuclear fuel
encapsulation and disposal facility
called ONKALO. Mr. Heinonen is also
actively involved in international cooperation
through IAEA, NEA and the
European Commission.
atw: How does final storage work
in Finland?
Heinonen: First of all I would like to
say that we are using the term
permanent disposal or repository.
Storage, even being final,
still refers more to something
that you store and
then take back. The
objective of disposal is
to perma nently dispose
radioactive waste in a
manner that is safe for the public, the
environment and future generations
to come.
Spent fuel is defined in Finland as
radioactive waste. Our national policy
is that the safe solution is disposal
Our national policy
is that the safe
solution is disposal
in Finnish bedrock.
in Finnish bedrock. The spent fuel
disposal in Finland is based on a
design where spent nuclear fuel
is packed in long-lasting disposal canister
and placed deep in our bedrock.
Canisters are surrounded with clay
material that protects canisters from
groundwater and small rock movements.
The basic principle is
to contain spent fuel canisters and
bedrock provides protection against
human activities and changes happening
at the surface.
atw: As science will evolve, was a
possible interest of retrieval considered,
at any point in time?
Heinonen: Posiva’s spent fuel disposal
has been designed so that it can
be retrieved. During operational time
disposal can be reversed if some type
of defect is identified or there is a
safety concern. Retrievability was
set as a requirement in decision-inprinciple
step. The Finnish bedrock is
stable and disposal canister has long
lifetime and this forms the technical
basis to make retrievability possible.
Technic has been demonstrated in
Swedish underground research
laboratory in Äspö. Retrievability
after closure is also
a possibility, however,
disposal is planned to be
permanent and retrieval
is not easy or cheap.
atw: Which framework
conditions can
promote a successful implementation
besides a stable environment?
Heinonen: From a technical point of
view a stable and predictable bedrock
environment is important. Then the
other parts of the disposal concept
Environment and Safety
How Final Disposal Can Work ı Nicole Koch

atw Vol. 65 (2020) | Issue 6/7 ı June/July
have to be designed so that the overall
system has multiple barriers and will
be safe even if some parts of that would
not function as expected. So framework
conditions or elements needed
would be a multibarrier system, simple
enough materials and com ponents
the behaviour of which we have
good understanding and the already
mentioned stable bedrock environment.
Stable environment can also refer
to society and the political environment.
This is also needed for the
successful implementation and is
further addressed in other questions.
atw: Finland has a leading position
in the final storage of high-level
waste. What are the differences to
other approaches?
Heinonen: A key thing, I would say,
is that our government made a strong
decision early on showing the will
to have a solution for spent fuel. We
have also had political will to support
progress in spent fuel
dis posal. Our stakeholders,
especially implementer
and regulator,
have committed
to follow the roadmap
established by the
govern ment and we have also
had courage to make the decisions
needed. These are important elements
of framework in the background.
This has meant in practice that in
1978 the government decided about
waste management principles including
implementation responsi bilities,
financing and R&D steering. The
govern ment made a principle decision
in 1983 that set the main steps and
time schedule for disposal development.
Posiva as implementer and
STUK as regulator have been following
these steps very well. Our government
and parliament have made decisions
when needed and they have understood
that using of nuclear energy also
requires that we need to have a safe
disposal solution. This has made it
possible to have concrete progress.
Finland, however, did not plan to
be in a leading position. The former
Posiva vice-president Timo Äikäs
sometimes joked that we have failed
our strategy miserably as we are now
in a leading position. The strategy
established in the 80’s was that
we will let the bigger countries have
disposal solved first and we will merit
from their experience. Other countries
like USA, UK and also Germany
have then had obstacles and we have
kept going and following our time
schedule. We failed to follow strategy,
We have also had
political will to
support progress in
spent fuel dis posal.
| Fig. 1.
Illustration of Encapsulation Plant (©Posiva Oy).
but we are having concrete progress
in disposal.
atw: What is needed
to gain public acceptance?
Heinonen: This is a
difficult question and
strongly related to national
culture. Therefore
something that has worked in
Finland might not be so useful in other
countries. Some fundamentals of
course exist. The public needs to have
trust in res pon sible stakeholders and
it needs to have possibilities to be
involved.
To gain trust the implementer,
re gulator and ministries have to be
open and transparent.
We need to be open for
dialogue with the public.
Also the roles of different
stakeholders need to be
clear. For example in
Finland private companies
are responsible for disposal
development, STUK is res ponsible to
evaluate safety and the Ministry of
Economic Affairs and Employment
of Finland ( MEAE) is responsible
for licensing and steering from the
government side. One element of
trusts is also com petence that for
example the public trust STUK to be
competent to evaluate safety and also
to be able to provide arguments why
we came to some conclusion.
In Finland Posiva as implementing
company has had the main responsibility
to provide information and
communicate with the public. Posiva
was nationally active during site
selection process and decision- inprinciple
1 . On local level Posiva has
The public needs
to have trust
in responsible
stakeholders.
had a communication group constituting
of local stakeholders for long
time. STUK has also been communicating
from its role. Our policy
has been that we are there for local
municipalities if they want our
service. During decision-in-principle
phase STUK had its own tour to have
communications with local public,
decisions makers and NGOs. One
principle has been that we have our
own events for communication. We
don’t have joint events with Posiva
except formal hearing events organized
by MEAE. Communication with
local public is one element that we still
continue. An other important element
is to provide information through
media and social media.
STUK’s policy is to serve
media and journalists. We
want our experts to be
present in media and to
provide interviews when
requested. We also want
to be actively involved in social media.
This is an area that is more under
development and there is still a lot
that can be improved. Our task is to
serve the public and in that role we
want to provide neutral and fact-based
information. And, it is important that
roles of different organisations are
clear.
atw: How does the cooperation
between authorities and corresponding
institutions work exactly?
Are the executing companies stateowned
companies?
Heinonen: In the Finnish system
nuclear power companies are directly
responsible to plan, implement and
finance nuclear waste management.
They have decided to establish Posiva
1) A decision-inprinciple
taken by
the Government
means a decision to
the effect that something
is in the overall
interest of society.
A decision-inprinciple
is applied
for by submitting an
application to the
Government. For the
discussion on the
application for the
decision-in-principle,
the Ministry for
Employment and the
Economy requests
statements from
the council of the
municipality in which
the planned repository
is to be located,
from the neighbouring
municipalities,
and from other
institutions such as
the Ministry of
Environment. In
addition, the Ministry
acquires a preliminary
safety assessment
on the project
from the Radiation
and Nuclear Safety
Authority.
ENVIRONMENT AND SAFETY 321
Environment and Safety
How Final Disposal Can Work ı Nicole Koch

atw Vol. 65 (2020) | Issue 6/7 ı June/July
ENVIRONMENT AND SAFETY 322
Facts & Figures Posiva
p Posiva Oy is a joint venture: owned by the two nuclear operators.
„TVO“ that holds 60 % and „Fortum Power and Heat“ with 40 %
p For a decade and a half, they have been working on the long-term
solution for the radiant legacy of nuclear power production
p In 2019, 86 people worked for the company
p 20.5 EUR were spent on R&D, which corresponds to 25.1 % of sales
p The profit was 1.88 billion EUR
p The underground research facility ONKALO® was registered as a
trademark within the EU area.
for co-operation in spent fuel disposal.
Executing or implementing companies
are private. Fortum, one of
Finnish nuclear companies, is partly
owned by State, but functions as
private company.
STUK is the safety authority. We
have a fully independent role in evaluating
safety and having oversight of
the use of nuclear energy. We have an
active dialogue with executing
companies and with the ministry. So
we are independent but not isolated.
This has been important especially in
the development of spent fuel disposal,
which is a first-of-a-kind project.
We need to follow close-enough
Posiva’s activities so that we are
capable to develop our own understanding
and safety regulation.
atw: Final disposal is a government
contract. What are the legal framework
and boundary conditions?
Heinonen: I understand the question
so that government has supreme responsibility
and in the end disposal will
come to state respon sibility. In general
waste producers are respon sible for
waste management, government and
partly parliament are making the
main licensing decisions, safety criteria
are provided in legis lation and
legislation also explains when waste
producers have ful-filled their task
and the closed disposal facility is
transferred as state respon sibility.
atw: A key aspect is long-term
safety. How has this been demonstrated?
Heinonen: Long term safety argumentation
is compiled in the safety
case which is a comprehensive collection
of data, models, reports and
argumen tation. The safety case is
compiled by Posiva and submitted to
STUK for review and assessment. Key
elements of the safety case and long
term safety are: understanding the
disposal site and engineered barriers
evolution, FEPs (features, events and
processes) that can have effect on the
disposal system, assessment of barrier
performance, analysis of possible
future scenarios and analysis of
radionuclide release through those
scena rios. The development of the
safety case requires a large amount of
research, characterization, modelling
and analysis work. Long term safety is
assessed broadly from different viewpoints
and over extremely long timescales.
atw: Another essential feature is
the corrosion-resistant copper
canister. Why did you end up with
copper?
Heinonen: Copper was proposed
and selected for canister material in
the early 80’s when SKB in Sweden
published the KBS-3 concept. Copper
is a material that exists also in nature,
it has good and passive
properties in anoxic
groundwater and it
has been evaluated to
last for long time.
Compared to some
other metals copper’s
corrosion resis tance is not based on an
oxide layer, which might be more
vulnerable in anoxic con ditions.
atw: Looking at the construction
process, did unforeseen difficulties
occur? If so, how have you dealt
with it?
Heinonen: In such a new and long
project some challenges or difficulties
are bound to occur. One challenge has
One challenge has
been to move from
the research-oriented
phase to construction.
been to move from the researchoriented
phase to construction. This
has been a challenge for Posiva and
also for STUK. In the con struction
phase all requirements and specifications
have to be exact and understandable
for workers coming outside
of the nuclear waste management
community. As ONKALO is a first-of- a-
kind construction it provides new
information which needs to be
adapted to safety eva luation. In the
first phase of Onkalo construction
Posiva had challenges in integrating
of research and construction. The
challenge was more contractual and
project management related than
technical. One challenge has been to
make decisions when the disposal
design is still partly developing and
uncertainties exist.
Overall there have been challenges
and difficulties. Many of them such
that can or could have been foreseen.
And most importantly we have been
able to overcome difficulties and have
progress in disposal.
atw: Is your expertise shared with
other countries looking for final
disposal and if so, what do you
think are common
challenges?
Heinonen: We share
our expertise in several
international groups
in IAEA, OECD/NEA
and in the European
Com mission. We also have frequent
visits from other countries that are
inte rested in Finnish disposal. The
situation that countries have differ.
Common challenges are in public
acceptance and in political clarity and
support. Of course safety and technical
development is also needed, but
this is seldom the reason not to have
progress in disposal.
| Fig. 2.
Finnish Disposal Container with an inner part made of steel and an outer shell made of copper
(©Posiva Oy).
Environment and Safety
How Final Disposal Can Work ı Nicole Koch

atw Vol. 65 (2020) | Issue 6/7 ı June/July
| Fig. 3.
Release Barriers (©Posiva Oy).
The disposal project
in general
In its construction licence application,
Posiva proposes the disposal of spent
nuclear fuel for a maximum of
6,500 tonnes of uranium (tU). This
corresponds to the accumulation of
spent nuclear fuel generated during
the operation of Teollisuuden Voima
Oyj’s (TVO) operating plant units
Olkiluoto 1 and 2, the plant unit
Olkiluoto 3 under construction, as
well as the operating Loviisa 1 and 2
plant units of Fortum Power and Heat
Oy (Fortum). The volume does not
include the spent nuclear fuel
delivered from the Loviisa plant units
to the reprocessing facility in Mayak,
Russia, in accordance with the agreement
that remained in force until
1996.
The spent nuclear fuel is stored in
interim storages at the nuclear power
plants, from which it will be transported
to the encapsulation plant for
disposal. The encapsulation plant has
not been designed for extensive
storage of nuclear fuel; instead, only
the amount of nuclear fuel intended
for disposal will be transported there
each time.
The project is based on the KBS-3
concept in accordance with the multibarrier
principle, in which the spent
nuclear fuel is packed into canisters
made out of copper and iron after a
minimum of 20 years of interim
storage and then disposed of in a
repository to be built in bedrock.
Posiva’s nuclear waste facility consists
of an encapsulation plant located on
top of the disposal facility above
ground as well as a disposal facility
reaching down to a depth of approximately
450 metres.
At the encapsulation plant, the
spent nuclear fuel is placed into a
disposal canister and the canister’s
copper lid is welded. The finished
disposal canisters are transferred
from the encapsulation plant into the
underground disposal facility via a
shaft. The construction of the encapsulation
plant will be completed
before the operation of the nuclear
waste facility begins.
In the disposal facility, the disposal
canisters are transferred into the
deposition tunnels and emplaced into
disposal holes lined with bentonite
clay. After the canisters have been
emplaced, the tunnels are backfilled
with clay material as the planned
number of canisters is emplaced in
them. More deposition tunnels are
constructed in the disposal facility as
the disposal progresses during the
operating period.
A repository will also be constructed
as part of the disposal facility
for the disposal of waste containing
radioactive substances generated
during the operation of the encapsulation
plant and disposal facility
and in connection with its decommissioning.
After all the spent nuclear fuel and
the nuclear waste produced during
use and decommissioning have been
disposed of, the operating period of
the nuclear waste facility will end
with the decommissioning of the
encapsulation plant located above
ground and backfilling as well as
sealing the rooms in the disposal
facility underground. Close to the
surface, the underground rooms are
filled in with structures that make
intrusion into the repositories difficult.
The planned disposal of spent
nuclear fuel will be passively safe after
closure. Ensuring the safety of disposal
will not require monitoring the
disposal site or other maintenance
activities after the disposal facility has
been closed.
Multibarrier principle
The multibarrier principle is a principle
guiding the design of the dis posal
of nuclear waste, which corresponds to
the defence-in-depth safety principle
required by Section 7 b of the Nuclear
Energy Act of Finland. In disposal in
the bedrock, the bedrock surrounding
the reposi tory acts as a natural barrier.
The characteristics of the bedrock must
be stable and maintain favourable
con ditions for the performance of the
engineered barriers. The bedrock must
also retard the migration of radioactive
material into the biosphere above the
bedrock. In designing the disposal
system, the waste matrix, waste
package, buffer surrounding the packages,
backfill of the emplacement
rooms and structures closing off the
entire disposal facility must be taken
into account as engineered barriers.
The activity of the spent nuclear fuel,
along with the risk caused by the radioactive
substances, shall decrease by
several orders of magnitude during the
first few thousands of years. For this
reason, the safety requirements
separately state that the engineered
barriers must effectively prevent the
release of radioactive substances into
the surrounding bedrock for several
thousands of years. The activity
concentration of the low- and intermediate-
level waste gene rated during
the use of the encapsulation plant is
significantly lower than the activity
concentration of the spent nuclear
fuel, and the half-life of the radioactive
materials is typically shorter; for this
reason, the engineered barriers are
required to contain the radio nuclides
for several hundreds of years for this
type of waste.
ENVIRONMENT AND SAFETY 323
Environment and Safety
How Final Disposal Can Work ı Nicole Koch

atw Vol. 65 (2020) | Issue 6/7 ı June/July
ENVIRONMENT AND SAFETY 324
State of Matter of the Fuel: The ceramic state of the fuel forms the
first release barrier in itself. The uranium within the gas-tight metal
rods is solid and dissolves in water only slowly, which slows down the
rate of release of radioactive substances.
Final Disposal Canister: The fuel is packed in a gas-tight, corrosionresistant
canister made of copper and cast iron. The canister protects
the fuel assemblies from the mecha nical stress occurring deep inside
the bedrock.
Bentonite Barrier: The final disposal canister is surrounded with
bentonite clay that protects the canister from any potential jolts
in the bedrock and slows down the movement of water in the
proxi mity of the canister.
Bedrock: The bedrock provides the canister and bentonite with
conditions where changes are slow and predictable. Deep in the
bedrock, the canisters are pro tected from any changes occurring
above ground, such as future Ice Ages, and kept away from people’s
normal living environment.
The spent fuel disposal solution is
primarily based on containment of
the radioactive materials from the
bedrock and the living environment.
The containment is primarily based
on maintaining the leak-tightness of
the disposal canister. The performance
of the canister is ensured
by the bentonite buffer that surrounds
it as well as the closure structures
of the disposal facility and bedrock
that surrounds the disposal facility,
which creates favourable and foreseeable
conditions for the disposal
system. As the radionuclides are
released from the disposal canister,
the second objective of the disposal
system is to isolate and retard the
migration of radionuclides into
organic nature.
Safety functions for the com ponents
of the spent fuel disposal
system:
p The safety function of the disposal
canister is
P to ensure a prolonged period of
containment of spent fuel
within the protective structures.
This safety function rests first
and foremost on the mechanical
strength of the canister’s cast
iron insert and the corrosion
resistance of the copper surrounding
it.
P to ensure the subcriticality of
the spent nuclear fuel in the
long term.
p The safety functions of the buffer
are intended to:
P contribute to mechanical, geochemical
and hydrogeological
conditions that are favourable
for the canister.
P protect canisters from external
processes that could compromise
the safety function of
complete containment of the
spent fuel and associated radionuclides.
P limit and retard radionuclide
releases in the event of canister
failure.
p The safety functions of backfilling
the deposition tunnels are intended
to:
P contribute to favourable and
predictable mechanical, geochemical
and hydrogeological
conditions for the buffer and
canisters.
P limit and retard radionuclide
releases in the event of canister
failure.
P contribute to the mechanical
stability of the rock adjacent to
the deposition tunnels.
p The safety functions of the closure
are intended to:
P prevent the underground
openings from compromising
the long- term isolation of the
re pository from the surface
environment and normal habitats
for humans, plants and
animals.
P contribute to favourable and
predictable geochemical and
hydrogeological conditions for
the other engineered barriers
by preventing the formation of
significant water conductive
flow paths through the openings.
P limit and retard inflow to and
release of harmful substances
from the repository.
p The bedrock acts as a natural
barrier, and its safety functions are
intended to:
P isolate the spent fuel repository
from the surface environment
and normal habitats for humans,
plants and animals and limit the
possibilities of human intrusion
and isolate the repository from
the changing conditions at the
ground surface.
P provide favourable and predictable
mechanical, geochemical
and hydrogeological conditions
for the engineered barriers.
P limit the transport and retard
the migration of harmful substances
that could be released
from the repository.
References
ı Posiva Oy Olkiluoto www.posiva.fi
ı STUK – Radiation and Nuclear Safety Authority www.stuk.fi
ı Safety assessment by the Radiation and Nuclear Safety Authority
of Posiva’s construction licence application,
February 11, 2015
Author
Nicole Koch
Editor
atw – International Journal
for Nuclear Power
nicole.koch@nucmag.com
Environment and Safety
How Final Disposal Can Work ı Nicole Koch

atw Vol. 65 (2020) | Issue 6/7 ı June/July
What has Happened to the U.S. Nuclear
Waste Disposal Program?
James Conca
Introduction One of science’s strongest abilities is to be able to reduce uncertainties in a problem.
If left to itself, science usually does
this very well. But it’s rarely left to
itself. Science exists within the larger
framework of society and has to deal
with the realities of politics, economics,
history and even religion.
Nowhere is this more obvious then
with nuclear waste disposal. For this
problem, the question we want to
know with a fair degree of certainty is:
If we put nuclear waste in this spot,
what’s likely to happen to it in 10,000
or 100,000 years? Will it contaminate
the environment before it decays away?
What are the risks to humans and the
ecosphere?
Unfortunately, even though we
in the scientific community have
answered these questions pretty
well, our nuclear waste program is
presently in shambles.
The nuclear waste program in
America began during WWII and the
making of the Bomb. The production
and reprocessing of fuel from weapons
reactors to make Pu resulted in the
first significant amount of nuclear
waste beginning in 1944.
With the increase in weapons
production, and the advent of commercial
power reactors in the 1950s, it
became obvious that we needed a
place to put this material away for
ever and ever.
The federal government asked the
National Academy of Sciences (NAS)
to come up with the best strategy
and, in 1957, they reported that
deep geologic disposal (half-a-mile
or so below the Earth’s surface) was
best (National Academy of Sciences,
1957). And they had a particular rock
in mind (NAS, 1957), the massive
Permian salt, which is the best rock for
isolating anything for a very long
time. And a rock that is pretty common
around the world, especially in North
America.
This recommendation led directly
to the only operating deep geologic
repository in the world, the Waste
Isolation Pilot Plant (WIPP) in New
Mexico (Conca, 2017). A splinter
strategy in the 1970s, involving
retrievability of spent nuclear fuel
from the depths, then led to the 1982
Nuclear Waste Policy Act (NWPA,
| Fig. 1.
The Yucca Mountain site was chosen in 1987 to be the first of two high level and spent nuclear fuel
repositories in America, but the basis for the final decision was quite political, causing large uncertainties
in the performance and the cost, as well as political resistance by the State of Nevada. DOE
1982) and its 1987 Amendment. That
legislation chose Yucca Mountain
(Figure 1), mainly through political
means, as the only repository for spent
nuclear fuel (SNF) and high-level
waste (HLW). The State of Nevada has
fought that decision ever since.
We submitted a license application
for Yucca Mt in 2008, but in 2009, the
Obama Administration terminated
the project and formed the Blue
Ribbon Commission (BRC) to recommend
alternative paths.
Meanwhile, WIPP continued on in
the salt (Figure 2), with its license
and permit curtailed to include only
transuranic waste (TRU), the other
| Fig. 2.
The Waste Isolation Pilot Plant (WIPP), the only
operating deep geologic nuclear waste repository,
is located 700 meters (2,130 ft) below
the surface in the massive salt of the Salado
Formation, and has been operating successfully
since 1999. The Salado Salt was chosen
originally based entirely on science. DOE
type of nuclear bomb waste, that is
mainly Pu, U, Am and other actinide
elements along with other nasties
(WIPP Land Withdrawal Act, 1992).
The LWA set aside 16 square miles for
nuclear waste disposal.
Periodically, attempts are made to
resuscitate the Yucca Mt. project, as is
being done once again as of this
writing. At the same time, attempts
are made to return WIPP to its original
NAS mission which was to dispose of
all nuclear waste – SNF, HLW and
TRU – for which it was designed and
built.
Finally, attempts are being made to
site an interim storage facility for SNF
near WIPP, either in New Mexico or
just across the border in West Texas,
under the regulatory umbrella of 10
CFR Part 72. SNF would be stored in
dry casks (Figure 3) after removal
from the spent fuel water pools where
the short-lived radionuclides are allowed
to decay away.
Nuclear Waste
The United States has over 80,000 tons
each of spent nuclear fuel (SNF)
and high-level nuclear waste (HLW)
although the forms of each are
quite different (Figure 4). SNF from
reactors is in a solid form that is easily
handled and easily stored in dry casks
ENVIRONMENT AND SAFETY 325
Environment and Safety
What has Happened to the U.S. Nuclear Waste Disposal Program? ı James Conca

atw Vol. 65 (2020) | Issue 6/7 ı June/July
ENVIRONMENT AND SAFETY 326
| Fig. 3.
When spent fuel is removed from the reactor it requires about five years in water to cool off and allow
the short-lived radionuclides to decay away. It can then transferred to dry cask storage (shown here)
until needed, e.g., burned in Generation IV or V fast reactors in the near-future, or disposed of more
easily in a deep geologic repository as it will be significantly cooler and less radioactive. NRC
once it is removed from the cooling
pools after about five years. HLW is in
different liquid, sludge and solid
forms in various containments such
as the 90 million gallons stored in
large tanks at Hanford, Savannah River
and other DOE facilities. HLW needs
to be soli dified and packaged by
various methods including grouting
( cementing), vitrifying (glassification)
or steam reforming (mineralization).
When dewatered, solidified and
repackaged, this HLW will have somewhat
over 80,000 metric tons of
heavy metals, referred to as MTHM.
In addition to SNF and HLW, a
minor amount of other wastes are
included in the discussion of a
deep geologic repository and include
nuclear navy waste, weapons proliferation-
related international waste,
research materials and greater than
Class C radioactive waste (GTCC).
GTCC includes activated metals from
decommissioned power plants, some
sealed sources from the irradiation,
medical and energy industries,
and non defense-related transuranic
(TRU) waste.
However, spent nuclear fuel may
not actually be waste since it can be
re-used in various forms in present
and future reactors, with or without
additional reprocessing depending
upon the reactor design. Since the
economics of re-use is in question,
SNF should be placed in an interim
storage facility at the surface where it
can be safely stored until needed, a
conclusion agreed upon by the
scientific community and the BRC.
Whatever use is made of SNF, there
will be eventual waste from it, even if
it is disposed of ultimately without
being re-used. So there will be a need
for the final repository regardless of
the future use of SNF. Interim storage
truly is interim, even if it could be a
hundred years. SNF, or its waste after
re-use, will be disposed in a deep geologic
repository at some point in time.
On the other hand, HLW is waste
that should be permanently disposed
as soon as possible since it was generated
primarily from reprocessing
spent fuel from old weapons reactors
and has no future value or use. The
decision to co-mingle SNF and HLW
administratively and physically in the
same repository led to the concept of
retrievability of the SNF, i.e., we might
change our minds about throwing
something so valuable away, so we
should construct the repository so
that we are be able to get only the SNF
back out in 50 years or so.
Unfortunately, retrievability makes
a deep geologic permanent repository
into a deep geologic interim storage
facility that we attempt to engineer or
morph into a deep geologic permanent
repository after we retrieve
the waste or decide to leave it in
place. The engineering and logistics
then becomes extremely difficult and
costly. However, since SNF, or its
waste after re-use, needs a repository
in the long run, co-mingling the
eventual waste may not be a problem
in the manner in which it is presently
considered, as there will not be a
retrievability issue at that time. This is
the problem that interim storage
solves – SNF is not physically comingled
during disposal of HLW in
the same repository but is disposed in
the same repository decades later –
co-mingled in space but not time.
The critical aspect about nuclear
waste, unknown to the general public
and their elected officials, is that there
is not much of it. All the nuclear waste
generated in the United States from its
nuclear power fleet in the last 60 years
can fit in a single soccer field – using
a light-water reactor assembly
dimension of 21.5 cm x 21.5 cm,
approximately 100,000 used assemblies,
and a regulation soccer field of
100 x 60 meters. Including all highlevel
defense waste in the U.S. more
| Fig. 4.
Four categories of nuclear and radioactive waste, their relative amounts and radioactivity, and an
example of SNF, HLW, and TRU, the three categories that require permanent deep geologic disposal by
law. DOE CBFO
Environment and Safety
What has Happened to the U.S. Nuclear Waste Disposal Program? ı James Conca

atw Vol. 65 (2020) | Issue 6/7 ı June/July
than doubles that but it still would
fit in the field with some in the
stadium.
Compared to that, the over
400 million tons of solid waste and
billion tons of CO 2 generated from
coal-fired power plants each year is
staggering. Even worse is the greater
than 500 million tons of solid chemical
and sanitary waste generated each
year, and the 2 quadrillion gallons of
water requiring waste treatment each
year. These are large waste volumes
that require thousands of disposal
sites, if they are regulated at all. On
the other hand, all of the nuclear
waste generated in the United States
in a thousand years could fit into one
repository. Yes, it’s bizarre material,
but easy to handle and relatively easy
to dispose.
Another critical aspect of the
defense HLW is that most of it is no
longer high level, except in name only.
So much time has passed that
significant amounts of radionuclides
have decayed away, and several
campaigns to remove the most radioactive
constituents ( 137 Cs and 90 Sr)
have left most of the HLW tank waste
with such low radioactivity that is
now falls into the category of TRUwaste
or LLW (Conca, 2014). But
bureaucratically and legally, it is
still considered HLW. Embracing the
ramifications of this development will
change our approach to this problem
in ways that would dramatically speed
up disposal and reduce costs over
our present path. DOE has pursued
this reclassification but public and
state opposition has slowed it dramatically.
Most importantly, no one has ever
died in the U.S. from handling, transporting
or disposing of nuclear waste,
and no one has ever died in the U.S. at
an operating nuclear power plant, a
tribute to our technical, industrial and
regulatory system. Because nuclear
waste is sufficiently odd and longlasting,
scientific opinion has long
considered deep geologic burial to be
the optimal method for permanent
disposal (National Academy of
Sciences, 1957; BRC 2011). The earth
is the only system that can operate
as expected for millions of years,
and we understand geologic processes
sufficiently to be able to choose an
optimal place to dispose of these
materials.
Optimal Characteristics Of
A Deep Geologic Repository
Characteristics of a suitable geologic
repository for the disposal of nuclear
waste include the following favorable
characteristics (McEwen 1995, EPRI
2006):
i. a simple hydrogeology,
ii. a simple geologic history,
iii. a tectonically interpretable area,
iv. isolation robustly assured for all
types of wastes (no difficult or
exotic processing needed),
v. minimal reliance on engineered
barriers to avoid extravagant costs
and long time extrapolations of
models for certain types of performance,
vi. performance that is independent
of the canister, i.e., canister and
container requirements are only
for transportation, handling and
emplacement in the repository,
vii. a geographic region that has an
existing and sufficient sociopolitical
and economic infrastructure
that can carry out operations
without proximity to a potentially
rapidly growing metropolis (unlikely
to ever have dense human
habitation near the site).
Two rock types that fit these characteristics
are argillaceous rocks (claystones
and shales) and bedded or
massive salts. Many studies have
focused on argillaceous sites, particularly
in Canada and Europe, with
some strong technical arguments
( Nuclear Energy Agency 2001);
similarly for salt deposits (McEwen
1995, National Academy of Sciences
1970). The primary difference
between salt and argillites is that,
while both have extremely low
permeability (the ability to conduct
water and the contaminants dissolved
in it), argillites have much higher
porosity (the total amount of pore
space, usually filled with water) and
molecular diffusion coefficients (the
ability of molecules and dissolved
contaminants to “randomly walk”
through the material independent of
the flow of water.
Massive salts have extremely low
porosity, molecular diffusion coefficients
and permeability. In fact, in
massive salt, permeability and diffusion
at the depths of a repository are
vanishingly small, so nothing moves
appreciably over millions of years. As
an example, in the massive salt of the
Delaware Basin spanning the borders
of New Mexico and Texas, a half-mile
below the surface it takes water, and
any contaminants in it, a billion years
to move an inch (Beauheim and
Roberts 2002; Conca et al. 1993).
Although salt deposits exist throughout
the world (Zharkov 1984), many
are not sufficiently massive, have too
many clastic interbeds, are tecto nically
affected (faulted and folded), or
are near population centers.
Salt domes and interbedded
salts are less optimal than massive
bedded formations from a hydrologic
standpoint, particularly within the
United States where diapiric movement
can exceed 1 mm/yr (McEwen
1995) and vertical spline fractures can
act as hydraulic conduits. Still, there
are many viable salt deposits in the
U.S. and globally that meet these
criteria (Zharkov 1984, Waughaugh
& Urquhart 1983, Karalby 1983).
The United States does not have
optimal argillites for this purpose.
It should be noted that volcanic
tuffs, like those at Yucca Mountain,
do not generally satisfy these criteria.
The Yucca Mountain tuffs have a
complicated dual-porosity oxidizing
hydrogeology, a complex geologic
and active tectonic history, and a
heavy reliance on engineered barriers
for the performance of a repository.
Finally, it is a mistake for disposal
programs in any country to attempt
to use old abandoned mines, even
in an otherwise good rock. Mines
dug for profit do not necessarily
possess the correct structure or depth
for optimal disposal. Simply taking
what an old mine gives you is not
wise and doesn’t save enough money
to justify the risk of failure. This
was one problem with the German
Asse Salt Repository. Any repository
should be designed and built strictly
with the disposal of nuclear waste
in mind.
Uncertainty
The ultimate basis for any choice of
rock and location is to maximize
properties you think are good, and
minimize properties you think are
bad. The best way to do this is to
use natural systems that have
already minimized these uncertainties
by minimizing or eliminating the
properties themselves. Pick a site
where almost nothing has been
happening for a long time, and where
almost nothing will happen in the
future.
This was WIPP.
Alternatively, you can try to impose
certainty on the system through
engineering, forcing each uncertain
variable to become practically
zero. Unfortunately, the Earth is a
large, active and open system that
resists long-term control by human
engi neering schemes and our understanding
and control of these processes
has always been limited.
ENVIRONMENT AND SAFETY 327
Environment and Safety
What has Happened to the U.S. Nuclear Waste Disposal Program? ı James Conca

atw Vol. 65 (2020) | Issue 6/7 ı June/July
ENVIRONMENT AND SAFETY 328
This was Yucca Mountain.
Yes, the Great Pyramids are fantastic
but that’s about all humans have made
that has lasted anywhere near what
we think of as geologic time, and what
we want for long-term disposal of
nuclear waste.
Beginning in 1982, the Yucca
Mountain Project hoped those
variables that cropped up were
sufficiently minor that they could be
handled by changing the design as we
discovered them.
Instead, what we did was just add
more variables with bigger uncertainties.
We addressed many by testing
and redesigning, or discovering new
information, over the years between
1987 and 2000, but the uncertainties
just grew. And because of that, so did
the projected cost.
There are many factors and properties
of a situation that contribute to
risk. For geologic containment, the
most important properties are the
characteristics of the rock itself,
especially the permeability, chemical
composition, strength, thermal conductivity,
density, porosity, and pore
water chemistry. These characteristics
determine which radionuclides move
in the subsurface, in what chemical
species they exist, and how fast they
migrate.
The host rock for the Yucca
Mountain repository, the Topapah
Spring tuff, is a highly fractured, dual
porosity and variably saturated
volcanic rock with highly oxidizing
pore water, that sits along the edge of
a tectonically-active region called the
Las Vegas Shear Zone in which the
Mojave Block is being rotated between
the San Andreas fault along the south
and the Garlock Fault along the north.
The permeability of water, or
hydraulic conductivity, varies from
10 -10 cm/s to 10 -4 cm/s in the tuff
matrix, from 10 -4 cm/s to 10 -2 cm/s in
small fractures, and greater than
10 -1 cm/s in large fractures and faults.
The ionic diffusion coefficient in
the pore space varies from about
10 -10 cm/s 2 to 10 -6 cm/s 2 depending
upon the volumetric water content
which varies from a few percent to
10 % depending upon the position,
degree of saturation and recent rainfall.
Perched water tables exist. In
some of the proposed engineered barriers
around the waste, the volumetric
water content would exceed 30 %.
The redox potential of the pore
water at Yucca Mountain is oxidizing,
with Eh (oxidation potential) values
greater than +200 mV, and causing
redox-sensitive radionuclides, like Np,
Se and Tc, to be under constant
threat of becoming mobile. This
aspect dominates the performance
assessment of Yucca Mountain.
Yucca Mountain now has several
engineered barriers that are supposed
to reduce the effects of particular
| Fig. 5.
A schematic of the engineered barriers proposed for the waste emplacement drifts at Yucca Mountain. The emplacement of the
barriers correctly is an extremely difficult and expensive logistical activity. The titanium alone for the drip shields will probably exceed
$30 billion. DOE YMP
properties, like the tuff’s relatively
large flux of oxidizing water, in the
hope of reducing their uncertainties.
These include reducing inverts,
shotcrete, robust waste containers
with copper and ceramic coatings,
titanium drip shields, vitrification of
HLW, waste package supports and
reducing gravel backfill (Figure 5).
Unfortunately, these have only
added uncertainty to the repository,
since their correct emplacement, rates
of degradation and time period for
optimal performance are themselves
uncertain.
On the other hand, a better way to
reduce uncertainty is to pick a situation
that has few variables or where
those variables have values that
naturally approach zero, which is
what the NAS did when they chose
Permian Salt in 1957. The Atomic
Energy Commission, and later the
Department of Energy, searched for a
suitable site in Permian Salt and, after
several failed attempts, was invited by
the local community in Carlsbad, New
Mexico to investigate their proposed
site.
Carlsbad was settled in the 1880s
by German miners mining salt above
what is the most optimal rock in the
United States for this purpose – the
massive Permian Salado Salt formation.
The miners and geologists in
Carlsbad understood the engineering
needs of the nuclear repository better
than the DOE, and understood that
the Salado Salt Formation near
Carlsbad would provide all of the
performance required even without
any engineered barriers.
The Waste Isolation Pilot Project
(WIPP) repository was sited in the
Delaware sub-basin of the Permian
salt in southeast New Mexico and
West Texas. It was designed and
built for all nuclear waste of any type.
Later, after the 1982 NWPA, WIPP
was licensed and permitted only for
transuranic nuclear weapons waste,
or TRU waste (Figure 6).
WIPP is just one place that has ideal
massive salt deposits. Although the
following discussion uses data from
the Salado Salt at the WIPP site, we
have well over 100,000 square miles of
appropriate massive salt deposits in
America with similar optimal rock
properties that would suffice for
nuclear waste disposal, 10,000 square
miles of Salado itself (Figure 7).
One particularly important property
of massive salt is something
called creep closure. At depth, under
the pressure of the overlying rocks,
the salt cannot maintain an opening,
Environment and Safety
What has Happened to the U.S. Nuclear Waste Disposal Program? ı James Conca

atw Vol. 65 (2020) | Issue 6/7 ı June/July
| Fig. 6.
Over 10,000 nuclear waste drums and standard waste boxes fills 1 of 56 rooms to be filled at WIPP over
its original 30-year mission, although more rooms are planned. Over 45 rooms have been filled as of this
writing. Note the high-activity remote handled waste (RH) robotically plunged into boreholes in the wall
to the left and plugged, while the contact handled waste (CH) fills the bulk of the room. 15 years of
operation – 100,000 cubic meters of TRU waste dispose. After 20 years of operation, over 120,000 cubic
meters of TRU waste have been disposed, over 600,000 fifty-five gallon drum equivalents, 21 storage
sites have been cleaned of legacy waste, no deaths, with only one minor release to the environment that
resulted in no lasting effects or people contaminated. DOE CBFO
fracture or pore space. It’s why the
salt is essentially impermeable, e.g.,
the trapped water in the salt hasn’t
migrated an inch in 270 million years.
The permeability of water, or the
hydraulic conductivity, is less than
10 -14 cm/s, and the aqueous diffusion
coefficient is less than 10 -15 cm/s 2 ,
amazingly low values that are, for all
practical purposes, zero. Water just
won’t move in this rock and, therefore,
neither will radionuclides.
The redox potential of the pore
water in this salt is exceptionally
reducing, one of the most reducing in
the country, with Eh values less than
-500 mV. This makes redox-sensitive
radionuclides such as Pu, U, Tc, Np,
Se, and I, immobile and unlikely to
migrate out of the repository in the
highly unlikely event that there is a
path out.
Which is unlikely in the extreme.
If a fracture does occur in the salt, or if
we dig out an opening to put waste in,
the salt creeps closed over a relatively
short time, tens of years depending
on the cut of the space The more
asymmetric the cut, like the long
rectangles of the disposal rooms, close
quickly, in years to tens of years. The
more symmetric the cut, like circles or
squares, close in decades. During this
process, the salt naturally recompacts,
or re-anneals, so there is no open
space and the salt becomes essentially
impermeable again.
In fact, any disturbance in rock
properties from a cut only goes out
about 14 feet from the wall of the
repository anyway. Beyond that, the
rock isn’t even disturbed. And these
formations are thousands of feet
thick. At WIPP, the Salado salt formation
is 2000 feet thick over an area of
about 10,000 square miles.
Two to three of those square miles
could hold all of the waste destined for
Yucca Mountain. WIPP already contains
more nuclear waste by volume
than everything that was supposed to
go into Yucca Mountain total, some of
it as radioactive as high-level waste.
In addition, the Salado Salt Formation
in this region has never been
subjected to any adverse geological
processes – no volcanism, no folding
or faulting, not even any tilting after
270 million years, quite unusual.
There was only some regional uplift.
In fact, this area is tectonically the
quietest region in America and will be
for the next 200 million years.
Which is why the National
Academy of Sciences picked this rock
in the first place. All the adverse
properties and possible processes are
either practically zero or non-existent.
Which means the uncertainties are
few and small, and come mainly from
the mining operation or future human
activities that we can never predict
or control. The NAS recommendation
for a salt host rock still stands and, in
fact, has been borne out by 20 years
of successful WIPP operations.
The Nuclear Waste Policy Act and
its Amendments established a 0.1 cent
per kilowatt-hour users fee on nuclear
generated electricity to pay for the
repository and associated costs, called
the Nuclear Waste Fund (NWFund).
The NWFund will have received about
$100 billion by the end of this century
depending on how nuclear energy
evolves in America.
However, recent reports from the
Government Accounting Office (GAO)
have shown how the projected costs
for Yucca Mt, including waste preparation
unique to YMP, have risen from
$80 billion to well over $400 billion.
The NWFund will certainly not cover
| Fig. 7.
Locations of geologic salt deposits in the United States. The Delaware Basin salts (yellow) are the least tectonically deformed,
are the thickest, and are at the most optimal depths for nuclear repository purposes. DOE CBFO
ENVIRONMENT AND SAFETY 329
Environment and Safety
What has Happened to the U.S. Nuclear Waste Disposal Program? ı James Conca

atw Vol. 65 (2020) | Issue 6/7 ı June/July
ENVIRONMENT AND SAFETY 330
these expenses, and taxpayers or
ratepayers would have to step in,
which is highly unlikely.
On the other hand, the cost of a
repository in a host rock like the
Permian salt will only be about
$30 billion, easily handled by the
NWFund. That’s because the cost is
actually a function of the choice of
rock, and the science that should be
determining which site is best.
No one envisioned that a political
choice, like the one that selected
Yucca Mountain, would have such a
profound cost effect because in 1982
no one understood the many aspects
and costs of a deep geologic repository.
But uncertainty increases cost.
The 30-year study of Yucca Mt
by all of us scientists and engineers
led to an amazing understanding of
how water and contaminants move
through the Earth’s subsurface and
how we can affect that to our benefit.
The $12 billion spent on that study
from the NWFund was not wasted at
all, most of it is useful no matter where
we put this waste.
Almost all of that information is
useable elsewhere no matter what
rock we pick. Our understanding
of corrosion, transportation, permeability
and subsurface contaminant
migration, engineered barriers,
shielding, packaging, waste form development
and material science,
among others, have been increased
enormously by studying Yucca Mt.
Using science as the basis of the
decision doesn’t just give you reduced
uncertainty, it also means getting the
lowest cost. That’s how science is
supposed to help society.
References
ı Beauheim, R. L. and R. M. Roberts. 2002. Hydrology and
hydraulic properties of a bedded evaporite formation, Journal of
Hydrology 259:66-88.
ı BRC. 2011. Draft Report to the Secretary of Energy. From the
Blue Ribbon Commission on America’s Nuclear Future (BRC),
Washington, D.C. http://www.brc.gov
ı Conca, J. (2014) “High-Level Nuclear Waste Redefined”, Proc.
2014 Annual Meeting of the American Nuclear Society, ANS,
La Grange, IL, Paper 10041.
ı Conca, J. (2017) “Environmental monitoring programs and public
engagement for siting and operation of geological repository
systems: experience at the Waste Isolation Pilot Plant (WIPP)”,
Chapter 24 in Geological Repositories for Safe Disposal of Spent
Nuclear Fuels and Radioactive Materials, J. Ahn and M. J. Apted
(editors), Woodhead Publishing (Elsevier), Cambridge, UK, 2 nd
Edition ISBN: 978-0-08-100642-9.
ı Conca, J. and J. Wright. 2010. The Cost of a Sustainable Energy
Future, Proceedings of the 2010 Waste Management Symposia,
Phoenix, AZ, paper #10494, p. 1-14. http://www.wmsym.org
ı Conca, J., S. Sage and J. Wright. 2008. Nuclear Energy and
Waste Disposal in the Age of Recycling, New Mexico Journal
of Science, vol. 45, p. 13-21 http://www.nmas.org/NMJoS-
Volume-45.pdf
ı Conca, J. L., M. J. Apted, and R. C. Arthur. 1993. Aqueous
Diffusion in Repository and Backfill Environments. In Scientific
Basis for Nuclear Waste Management XVI, Materials Research
Society Symposium Proceedings. La Grange, IL 294:395-402.
ı DOE. 2008. Disposal of Recycling Facility Waste in a Generic
Salt Repository, GNEP- WAST-MTSD-MI-RT-2008-000245
PREDECISIONAL DRAFT April 2008, U.S. Department of Energy,
Washington, D.C.
ı DOE. 2004. Title 40 CFR Part 191 Subparts B and C Compliance
Recertification Application, DOE/WIPP 2004-3231, March 2004,
U.S. Department of Energy, Washington, D.C.
ı EPRI. 2006. Room at the Mountain: Analysis of the Maximum
Disposal Capacity for Commercial Spent Nuclear Fuel in a Yucca
Mountain Repository, Technical Report 1013523, EPRI Program
on Innovation. Electric Power Research Institute, Palo Alto, CA.
ı Griffith, J. D., S. Willcox, D. W. Powers, R. Nelson and B. K. Baxter.
2008. Discovery of Abundant Cellulose Microfibers Encased in
250 Ma Permian Halite: A Macromolecular Target in the Search
for Life on Other Planets, Astrobiology, Vol. 8, No. 2, p.1-14,
Mary Ann Liebert, Inc. DOI: 10.1089/ast.2007.0196
ı Hamal, C.W., J. M. Carey and C. L. Ring. 2011. Spent Nuclear Fuel
Management: How centralized interim storage can expand
options and reduce costs, a study conducted for the Blue Ribbon
Commission, Washington, D.C. May 16, 2011. http://brc.gov/
sites/default/files/documents/centralized_interim_storage_of_
snf.pdf
ı Karalby, L. S. 1983. High-Integrity Isolation of Industrial Waste
in Salt. In Proceedings of the Sixth International Symposium on
Salt, The Salt Institute 2:211-215.
ı McEwen, T. 1995. Selection of Waste Disposal Sites. Chapter 7
in the Scientific and Regulatory basis for the Geologic Disposal
of Radioactive Waste, pp. 201-238, D. Savage, ed.. John Wiley &
Sons, New York.
ı NWPA. 1982. Pub. L. 97–425 (96 Stat. 2201), Nuclear Waste Policy
Act of 1982 as amended http://epw.senate.gov/nwpa82.pdf
ı National Academy of Sciences. 1957. Disposal of radioactive
waste on land, Report by the Committee on Waste Disposal,
Division of Earth Sciences, The National Academies Press,
Washington, DC. https://www.nap.edu/read/10294/chapter/1
ı National Academy of Sciences. 1970. Disposal of Solid
Radioactive Waste in Bedded Salt Deposits, Board on
Radioactive Waste Management, Wash., D.C.
ı National Academy of Sciences. 2006. Safety and Security of
Commercial Spent Nuclear Fuel Storage: Public Report.
Committee on the Safety and Security of Commercial Spent
Nuclear Fuel Storage, National Research Council. ISBN: 978-0-
309-09645-4. http://www.nap.edu/catalog/11263.html
ı Nuclear Energy Agency. 2001. IGSC Working Group on Measurement
and Physical Understanding of Groundwater Flow through
Argillaceous Media: Self-Healing Topical Session. OECD/NEA
NEA/RWM/ CLAYCLUB(2001)5. Nancy, France.
ı Parkyn, J. 2009. Interim Storage Costs General Accounting Office
(GAO) Study, Wisconsin Public Utility Institute, July 31, 2009.
http://wpui.wisc.edu/files/2009/materials/WPUI_20090731.
Parkyn.John.pdf
ı TSLCC. 2008. The Analysis of the Total System Life Cycle Cost
(TSLCC) of the Civilian Radioactive Waste Management Program,
DOE/RW-0591, U.S. Department of Energy, Washington, D.C.
http://www.rw.doe.gov
ı Waughaugh, D. C. E. and B. R. Urquhart. 1983. The Geology
of Denison-Potacan’s New Brunswick potash Deposit.
In Proceedings of the Sixth International Symposium on Salt,
The Salt Institute 1:85-98.
ı WIPP Land Withdrawal Act. 1992. Pub. L. 102–579, section
2(18) as amended by the 1996 WIPP LWA Amendments,
Pub. L. 104–201.
ı Wright, J. and J. L. Conca. 2007. The GeoPolitics of Energy:
Achieving a Just and Sustainable Energy Distribution by 2040,
BookSurge Publishing (on Amazon.com) North Charleston,
SC. ISBN 1-4196-7588-5. http://www.amazon.com/gp/
product/1419675885/sr=1-10/qid=1195953013/
ı Zharkov, M. A. 1984. Paleozoic Salt Bearing Formations of the
World, Springer-Verlag, Berlin.
Links
Listed in order of appearance in the text and active
as of June 6, 2020
https://www.nap.edu/read/10294/chapter/9
https://www.wipp.energy.gov
https://www.energy.gov/downloads/nuclear-waste-policy-act
https://en.wikipedia.org/wiki/Yucca_Mountain_nuclear_waste_
repository
https://www.nrc.gov/waste/hlw-disposal/yucca-lic-app.html
https://www.energy.gov/sites/prod/files/2013/04/f0/brc_
finalreport_jan2012.pdf
http://nuclearconnect.org/transuranic-waste
https://www.nrc.gov/waste/spent-fuel-storage/cis.html
https://www.neimagazine.com/features/featureclearing-outasse-2
https://www.nap.edu/read/10102/chapter/4#47
https://fas.org/sgp/othergov/doe/lanl/pubs/00818052.pdf
https://en.wikipedia.org/wiki/Reduction_potential
https://www.epa.gov/laws-regulations/summary-nuclear-wastepolicy-act
https://www.elsevier.com/books/geological-repository-systemsfor-safe-disposal-of-spent-nuclear-fuels-and-radioactive-waste/
apted/978-0-08-100642-9
https://en.wikipedia.org/wiki/Hydraulic_conductivity
https://www.comsol.com/multiphysics/diffusion-coefficient
https://www.energy.gov/sites/prod/files/AH-Chap19.pdf
https://www.gao.gov/highrisk/us_government_environmental_
liability/why_did_study
https://xcdsystem.com/wmsym/archives//2012/papers/
12469.pdf
Author
James Conca
UFA Ventures, Inc.
jim@ufaventures.com
2801 Appaloosa Way
Richland, WA 99352, USA
Environment and Safety
What has Happened to the U.S. Nuclear Waste Disposal Program? ı James Conca

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Safely Stored for All Eternity –
How the Bundesgesellschaft für Endlagerung is Conducting
its Search for a Repository for High-level Radioactive Waste
Steffen Kanitz
Germany has spent the last three years with a restart for its attempts at selecting a site for a repository for high-level
radioactive waste. In fall 2020, the Federal Company for Radioactive Waste Disposal (Bundesgesellschaft für
Endlagerung, BGE) is due to present an initial evaluation of data on the deep geological conditions, which is intended
to provide some guidance on which areas are unsuitable and which may be suitable for a geological repository. The
Federal Office for the Safety of Nuclear Waste Management (Bundesamt für die Sicherheit der nuklearen Entsorgung,
BASE) will then take the first step towards formal public participation and issue invitations to the specialist Subareas
Conference. And a new player has come onto the field, the National Monitoring Panel (Nationales Begleitgremium
NBG), which recently completed its line-up. Whereas the 18 th legislative session of the Federal Parliament (2013 to
2017) concentrated on the organizational restructuring of the repository landscape, the focus has now shifted to
realization.
The principle of a blank map of
Germany applies. No site is excluded
right from the outset, no site is
included right from the outset. This
is a political compromise supported
by a broad parliamentary majority
for a task which has spanned several
generations and previously been
marked by major conflicts. This compromise
forms the basis for the work
being undertaken by the BGE to
search for the site in Germany which
offers optimum safety for a million
years.
Looking back
For decades, the disposal of high-level
radioactive waste in Germany seemed
to be a problem that had almost
been solved - at least as far as the two
main political groupings, CDU/CSU
and SPD, were concerned. In 1977,
Ernst Albrecht (CDU), then Minister
President of Lower Saxony, chose
Gorleben as the site of a nuclear
disposal facility and hence the site of a
repository for high-level radioactive
waste, too. This decision was supported
by the Federal Government of
Helmut Schmidt (SPD) in Bonn.
But more than 40 years on from
this, it has to be admitted that the first
attempt at solving the repository issue
in Germany has been a failure. The
selection process, considered by
some sections of the public to be very
intransparent, stirred up resistance
not only in the region around Gorleben
(Wendland) itself, but across the
whole of Germany – and kept it alive
for decades. Wendland, a remote
region in the state of Lower Saxony on
the borders to Saxony- Anhalt and
Brandenburg, has more over even
experienced a large influx of people,
namely those wanting to express their
resistance against nuclear energy and,
at times, against the government,
too. The resistance movement grew
with every Castor transport into the
Gorleben interim storage facility.
Fresh start in the search
for a repository site
The Federal Parliament resolution
of 2011 to phase out nuclear
power after the nuclear disaster in
Fukushima, Japan, which followed in
the wake of a powerful earthquake
and a tsunami, paved the way for a
new attempt at finding a consensus
for a repository. Norbert Röttgen
(CDU), then Secretary of State
for the Environment, and Winfried
Kretschmann (Green Party), Minister
President of Baden- Württemberg,
started a dialog at that time
which produced an initial result two
years later. In 2013, the first
Repository Site Selection Act
(StandAG) was passed, and provided
for a fresh start in the search for a
repository. Between 2014 and
2016, the Repository Commission – a
body of scientists, various social
groups, the Federal Parliament, and
the Federal Council (although the
politicians had no voting rights) –
conducted its deliberations, which
were chaired by Ursula Heinen-Esser
(CDU) and Michael Müller (SPD).
It drew up the scientific criteria
for the procedure, and the principles
for full public participation, too.
These results were taken as the
basis for the amendment to the
StandAG in 2017, and the definition
of the site selection pro cedure:
The search is to be undertaken by
way of a science-based, transparent,
participative, self-scrutinizing and
learning process. This is the foundation
on which the new search for a
repository is to be conducted.
What does the search aim
to do?
Its aim is to find a site in Germany deep
underground where the high- level
radioactive waste can be safely sealed
off from the environment and from us
humans for a million years. The waste
comprises around 10,200 metric tons
of spent fuel elements and approx.
6,000 cubic meters of vitrified waste
from the Sellafield and La Hague
reprocessing plants. At present,
this waste is safely stored in casks
(Castors) in interim facilities. Its
volume is small compared to that of
the low-level and intermediate-level
radioactive waste, but these materials
are the source of more than 99 percent
of the radiation from the radioactive
waste in Germany.
Are there any alternatives
to a repository?
The Repository Commission gave
thorough consideration to alternative
forms of disposal, but ultimately
rejected them for reasons which are
easy to comprehend. In times of global
climate change, a repository in the
pack ice cannot be viewed as a
long-term solution. Although the idea
of disposing of the radioactive waste in
outer space may sound logical, it
would only need a failed rocket
launch to cause a nuclear disaster.
Constructing thick-walled, groundlevel
storage facilities imposes the
responsibility for protecting and maintaining
these facilities in the long term
on future generations, not to mention
the fact that such facilities would be
potential targets for terrorist attacks.
Some people say that so-called
partitioning and transmutation offers
a solution to the nuclear waste
problem. The idea here is that longlived
radionuclides are con verted into
ENVIRONMENT AND SAFETY 331
Environment and Safety
Safely Stored for All Eternity – How the Bundesgesellschaft für Endlagerung is Conducting its Search for a Repository for High-level Radioactive Waste ı Steffen Kanitz

atw Vol. 65 (2020) | Issue 6/7 ı June/July
ENVIRONMENT AND SAFETY 332
shorter-lived radioactive materials
through targeted irradiation in a new
generation of reactors, thus restricting
the length of time they have to
be stored, and that the materials
processed in this way are reused as
nuclear fuel. This tech nology has
remained stuck at the experimental
stage for decades, however. Moreover,
a geological repository would be
needed for the waste from this
process, too. And: Using partitioning
and transmutation to help get rid
of the quantities of radioactive
waste which have already accrued
in Germany would require the technology
to be in operation for at least
150 years – the problem would be
passed on to future generations
instead of relieving them of the
burden of high-level radioactive
waste.
The Repository Commission agrees
with the assessment of international
experts that a geological barrier
deep underground is the only way to
guarantee that the radioactive waste
is permanently and safely sealed in.
How will the site search
be undertaken?
The first phase of the site selection
involves the BGE working with
the data on the deep geological
con ditions which are already available
from the federal government
and federal state authorities. The
exclusion criteria, minimum requirements,
and the geoscientific assessment
criteria defined in the StandAG
will then be applied to the data which
already exist. Thus, the work initially
involves studying the records documenting
our existing knowledge
on the deep geological conditions in
Germany which are available from the
federal state and federal government
authorities. The Geological Surveys
of the federal states, and the Federal
Institute for Geosciences and Natural
Resources (Bundesanstalt für Geologie
und Rohstoffe, BGR), have made
substantial volumes of existing data
available to the BGE.
Exclusion criteria
The BGE now examines whether these
pools of data can already be used to
deduce which areas are not suitable
for a repository (exclusion criteria).
These are areas in which geogenic
uplifts of more than a millimeter
per year are observed, or are to be
expected over the course of a million
years. Regions where mining is still
taking place or used to take place, or
where there are boreholes at depths
of between 300 and 1500 meters,
are also to be excluded, because
the integrity of the rock has been
weakened. Active fault zones, where
the layers of rock are shifting against
each other, are also to be excluded.
Other exclusion criteria are volcanic
activity, seismic zones above zone 1,
and so-called young groundwater.
All the exclusion criteria indicate
rock movements which prevent the
permanent, safe storage of high-level
radioactive waste.
The BGE has developed an exclusion
methodology for each exclusion
criterion. The mining criterion is
subdivided into mines and boreholes,
because the impact of these two
anthropogenic effects is different.
The guideline followed by the BGE
is the maxim of excluding as little as
possible the first time the criteria
are applied, so as not to overlook or
discount an area which may possibly
be suitable. Each exclusion methodology
follows a strict schematic and
is easy to understand. It also covers
how to deal with cases which cannot
be definitively assessed from behind a
desk, such as the question of whether
a fault zone is active or passive, and
what its precise course is, for example.
The BGE therefore proposes that
fault zones reported as active by the
Geological Surveys are taken to be
active in the first step. The BGE
assesses rock movements which are
more recent than 34 million years ago
as a further indication of the activity
of a fault zone. The so-called Rupel
stratum in the geological models
and maps is an indicator of rock
movement which occurred less than
34 million years ago. In addition, the
BGE has specified a further condition
as a result of information from the
online consultation on the methodology:
Fault zones which are located
in tectonically active major systems –
one example would be the Upper
Rhine Graben – are also assessed as
being active. When concrete information
is available on the course of
an active fault zone, a one-kilo meter
protection zone is placed around it,
and the area is then pro jected onto the
surface and “cut out” of the blank
map. If no information is available, it
is initially simply excluded together
with its surrounding protection zone
in the vertical direction.
Minimum requirements
The minimum requirements are then
applied in a second step to establish
which areas in Germany could in
principle host a repository. The BGE is
searching for a stable rock formation
| Exclusion Criteria
Environment and Safety
Safely Stored for All Eternity – How the Bundesgesellschaft für Endlagerung is Conducting its Search for a Repository for High-level Radioactive Waste
ı Steffen Kanitz

atw Vol. 65 (2020) | Issue 6/7 ı June/July
which is as impervious as possible
at a depth of between 300 and
1500 meters. Three rock formations
are suitable to retain radionuclides
over a period of a million years: Rock
salt, clay rock, and crystalline rock.
The BGE will present the definitions
for the host rocks it has used before
the publication of the Subareas
Interim Report. The rock layer in
which a storage location is to be found
must be at least 100 meters thick.
Slightly different requirements apply
to salt in a steeply inclined formation,
i. e., salt domes, and also to crystalline
rock, but these are likewise clearly
defined in the StandAG. Furthermore,
it is important that the rock is as
impervious as possible to water, and
even retains gases, because radionuclides
could migrate with the aid of
water or gas.
To identify areas in which the
minimum requirements are met,
the BGE has utilized a great many
databases and maps, and a wealth of
expertise. When a 3D model of the
deep geological conditions was
available for a federal state or parts of
a federal state, the BGE used it to
determine host rock formations
and their thickness, for example. The
BGE has used paleogeographic and
geological maps, ground profiles
from boreholes, and other suitable
sources of information, to fill the
gaps between the models with
knowledge and justified assumptions.
Geo-scientific assessment
criteria
In the third step, the BGE evaluates
the areas in which all the minimum
requirements are met and no
reason for exclusion exists, in order
to identify subareas that lead one
to expect a favorable geological
situation. To be able to systematically
process the eleven geoscientific
assessment criteria, which are evaluated
with the aid of 40 indicators,
the BGE specialists developed an
Access-based evaluation tool which is
used to individually assess each of
the areas identified. The evaluation
results are documented in a comprehensible
way.
Subareas Interim Report
In fall 2020, the BGE will present a
Subareas Interim Report which will
contain the evaluation of this initial
exploratory phase. The Interim Report
will explain for one how the subareas
identified have been arrived at. The
methodology used to apply the criteria
from the Repository Site Selection
Act will be described, fundamental
stipulations and definitions will be
derived, and an overview of the
database used will be provided. These
steps, as well as the history of how
aspects such as an exclusion methodology
were derived or developed,
will be described in more detail in a
series of supporting documents. Even
before the first step of full public
participation is taken, results of an
online consultation on the methodologies
will be included in the
Interim Report. In addition, the BGE
has organized several specialist
workshops with the Geological
Surveys over the past three years,
and also sought the dialog with
the scientific community. Worthy of
mention is a specialist workshop on
the research needs for the site selection
in January 2019, and in particular
the “Site Selection Conference” in
Braunschweig in December 2019.
The findings from the talks, poster
sessions, and short presentations by
scientists from universities and
institutes are also reflected in the
work, and have sometimes already
been incorporated.
After publication of the Subareas
Interim Report, the BASE will issue
invitations to a specialist Subareas
Conference, at which the BGE will
present the results of the Interim
Report. The BGE will incorporate
the results of the conference into its
subsequent work. At the same time,
the BGE will conduct the first, still
very generalized safety studies in
the subareas which have then been
identified. It will propose survey programs
whereby the conditions underground
can be explored in more
detail. The issue is initially to survey
the areas from the surface, by means
of boreholes, seismic measuring programs,
or other methods.
What happens next?
A BGE proposal for site regions
where a surface survey appears to
be worthwhile marks the end of the
first phase. The BASE will examine
the BGE proposal and either accept it
or make a modified proposal to the
Federal Ministry for the Environment,
which will submit the proposal to the
Federal Parliament in the form of a
draft bill. Parliament will then decide
where surveys are to be undertaken.
The next stage is the surface surveys,
which will then be used to derive a
proposal for the areas where underground
surveys also are to be carried
out. Parliament will again make the
decision on this. Finally (target date
2031), there will be a site proposal on
which the Federal Parliament will
make the decision. The goal is for the
repository to be available in 2050.
After another 50 or so years, the
disposal process will be complete, the
repository will then be sealed. Only
then will the nuclear relics of the
peaceful utilization of nuclear energy
in Germany have been disposed of
safely and permanently.
Author
Steffen Kanitz
Managing Director
Bundesgesellschaft
für Endlagerung mbH (BGE)
dialog@bge.de
Eschenstraße 55
31224 Peine, Germany
ENVIRONMENT AND SAFETY 333
Environment and Safety
Safely Stored for All Eternity – How the Bundesgesellschaft für Endlagerung is Conducting its Search for a Repository for High-level Radioactive Waste ı Steffen Kanitz

atw Vol. 65 (2020) | Issue 6/7 ı June/July
334
RESEARCH AND INNOVATION
Off-site Consequence Analysis During
Severe Accidents in a Nuclear Power Plant
Dahye Kwon and Moosung Jae
To meet the Korean new regulation on Level 3 probabilistic safety assessment, the off-site consequence analysis is
required for the severe accident of nuclear power plant. To consider socio-environmental characteristics and regulation
system for radiation protection in Korea, several important parts in MELCOR Accident Consequence Code System 2
(MACCS2), an off-site consequence analysis code, were modified such as dose conversion factor, shielding factor,
inhalation rate, and food chain model. The modified parts were applied to evaluate the accident consequence of a
Korean reference nuclear power plant, OPR-1000. As a result, the derived health effect consequences were decreased
with reflecting Korean characteristics comparing those with US default values. With the contribution analysis of each
factor comparing non-modified MACCS2, the results were decreased with modified shielding factor and inhalation rate,
but the result was increased with modified food chain model because of Korean diet habits.
1 Introduction
In June 2016, to quantitatively secure
the safety of Nuclear Power Plants
(NPPs), the Nuclear Safety and
Security Commission (NSSC) revised
the notification to apply a safety goal
to Level 3 Probabilistic Safety Assessment
(PSA). In order to carry out the
PSA, the researchers usually use
MELCOR Accident Consequence Code
System 2 (MACCS2) [1], a PSA code
developed in the US [2]. However,
since the default values of input
parameters reflect the representative
environment and radiation protection
systems of US, the calculation result
using default values might not be
guaranteed in reliability of Korean
NPP. Therefore, it is important to
derive the input parameter to reflect
the environmental characteristics to
improve the reliability of results.
In this study, based on expert
elicitation [3], selected four factors as
research objective are Dose Conversion
Factor (DCF), shielding factor,
inhalation rate, and Food Chain Model
(FCM). These Korean specific data
are analyzed and the representative
values are derived. Then, Level 3 PSA
of Korean reference NPP, OPR-1000,
is carried out using derived Korean
specific data, and its results are
compared with those when US specific
data is applied.
2 Methods and materials
2.1 Dose conversion factor
While the US regulation applies the
concept of the ICRP publication 26
to MACCS2, the concept of ICRP
publication 60 is applied to the regulation
in Korea. Hence, we extracted
the DCFs from the FGR-13 DCF
database which follows the ICRP
publication 60.
For the appropriate DCF, the
exposure pathways were selected to
meet the regulatory guideline from
The Korea Institute of Nuclear Safety
(KINS), the Korean regulatory agency.
The organs were adopted 16
critical organs as follows: the 12 major
organs which were suggested by ICRP
publication 60: bone surface, breast,
stomach, bladder, liver, red marrow,
skin, thyroid, esophagus, lung, gonads
and colon; and the others were Lower
Large Intestine (LLI), small intestine,
remainder, and effective dose. The LLI
and small intestine are important
organs in the deterministic health
effect assessment from acute exposure.
The number of radionuclides
was expanded from 60 to 825. It is
important to consider the nuclides as
many as possible, because the source
term of severe accidents can vary
considerably depending on the reactor
type and accident scenarios. As a
result of this study, it is possible to
evaluate various source terms.
2.2 Shielding factor
The shielding factor used in MACCS2
is defined as indoor dose over outdoor
reference dose and it is used to consider
the shielding effect of buildings.
The outdoor reference dose generally
refers to the dose at an elevation of
1 meter above the surface of an infinite
smooth surface source. The shielding
factor can be obtained from Equation
1.
SF = (1 – Indoor) + Indoor × RF
(1)
where Indoor: Indoor residence time
fraction, and
RF: Reduction factor.
The ‘Indoor’ variable was 0.829 for
adults [4]. The adapted ‘RF’s were
0.2 for radioactive plume and 0.01 for
contaminated ground surface [5].
Because the groundshine reduction
factor differs depending on the residential
type, we adopted the value of
apartment in which 60 % of the
Korean live. By substituting the given
values into Equation 1, the shielding
factors for cloudshine and groundshine
were derived as 0.34 and 0.25,
respectively.
2.3 Inhalation rate
The inhalation rate in MACCS2 means
the daily averaged inhalation rate.
It can be obtained as shown in
Equation 2.
(2)
where IR: Average daily inhalation
rate [m 3·d-1 ],
BR i : Short-term inhalation rate
for a specific activity i
[m3·hr -1 ], and
D i : Duration of the activity i
during a day [hr·d -1 ].
In the research on inhalation rate,
researchers usually uses the calcu lated
short-term inhalation rate considering
the respiratory tract model recommended
from ICRP publication 66.
But, it was inappropriate for Korean
because the phantom used in the
calculation is modeled as male and
female Caucasian adults. In this paper,
we used directly measured short-term
inhalation rate for male and female
Korean adults [6], which improves the
reliability. The duration of the activity
for adults was referred from the
national statistics. Thus, the inhalation
rate reflects the physical and
social characteristics of Korean. The
calculated inhalation rate of Korean
adults was 18.51 m 3·d -1 . It is about
20 % lower than the default value.
Research and Innovation
Off-site Consequence Analysis During Severe Accidents in a Nuclear Power Plant ı Dahye Kwon and Moosung Jae

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Food category
2.4 Food chain model
2.4.1 Introduction of COMIDA2
COMIDA2 [7] evaluates the activity
concentration for each radionuclide
and ingestion dose received from the
intake of contaminated food. The
result is compatible with MACCS2
calculation. There are five categories
of vegetable foods and four categories
of animal foods: grains, leafy vegetables,
root vegetables, fruits, and
legumes; beef, milk, poultry and
other.
2.4.2 Improvement of Korean
food chain model
To reflect Korean agricultural and
cultural environment, some input
variables were derived: annual food
consumption and productivity, feedstuff
consumption of each livestock,
transfer coefficient, wet-to-dry weight
ratio, and the processing factors.
Especially, the category ‘other’ means
the ‘pork’ consumption because pork
shows the highest consumption rate
for meat in Korea. These variables
referred from a paper [8] were partially
shown in Table 1.
3 Results
3.1 Accident scenario
3.1.1 Source term
The accident source term was adopted
from the core inventory evaluated
for the Korean reference NPP of
OPR-1000. The initial event was
chosen as Steam Generator Tube
Rupture (SGTR), which was the most
frequent bypass accident among
internal accidents. The core inventory
was evaluated using the FISPACT-2
code, and the result was evaluated
Consumption
(kg·yr -1 )
Productivity
(kg·m -2 )
Vegetable food Grain 25.41 0.06
Leafy vege. 76.48 0.18
Root vege. 46.39 0.11
Legumes 0.57 0.001
Fruits 52.67 0.13
Subtotal 201.53 0.48
Animal food Beef 3.37 0.01
Pork 9.37 0.02
Poultry 13.26 0.03
Milk 18.41 0.04
Subtotal 44.41 0.11
Total 245.94 0.59
| Tab. 1.
Annual food consumption and productivity of Korean
at 10,000 MWd/MTU operation. In
addition, the considered radio nuclides
were 60 species [9].
3.1.2 Meteorological data and
site-specific characteristics
The meteorological data except
atmospheric stability were measured
at the met mast near the reference site
and the atmospheric stability was
measured at the met mast in the site.
The 18 radii were used by the
regulation about radiation environmental
assessment. All of site-specific
characteristic data were based on the
reference NPP site. The population
distribution was the result of the
MSPAR-site code [10], and the others
were the results of the KOSCA-POP
code [11].
3.2 Korean consequence
analysis of severe accident
As a result, we analyzed the comprehensive
consequence and the contribution
of Korean characteristics,
namely shielding factor, inhalation
rate, and FCM. The DCF was fixed to
the modified one, because the DCF
is not a characteristic reflecting Korean
environment, but a regulation requirement.
To focus on the effect of
Korean environmental characteristics,
we excluded the emergency response
scenario.
The analyzed consequences are
three health effects which are early
fatality in early phase, cancer fatality
in early phase, and cancer fatality in
chronic phase. In the figures, the
X axis is the health effect consequence
on the logarithmic scale, and the
Y axis is the relative ratio between the
probabilities exceeding consequence
X before and after the modification on
the linear scale.
3.2.1 Influence of Korean
characteristics and models
In Figure 1, the ratio of the consequence
probability by modification
of shielding factor, inhalation rate,
and FCM is depicted. The result was
reduced in all of health effects. It came
from that the shielding factors were
decreased because most Korean lives
in an apartment, which is a skyscraper
with cement or stone construction,
against wooden building of US citizen.
Moreover, it was because the inhalation
rate was decreased due to the
smaller frame and lung capacity of
Korean than those of Caucasian.
In Figure 2, due to the reduction of
shielding factors, all of results were
highly decreased. In particular, the
decrease of early fatality was very
large. It came from the decreased
cloudshine shielding factor, which
became almost a half of the default
value. Since the cloudshine shielding
factor was applied to the very first
period following the severe accident,
the activity concentration was largely
| Fig. 1.
Relative probability occurring the health effect, (A) early fatality in early
phase, (B) cancer fatality in early phase, and (C) cancer fatality in chronic
phase, considering all of Korean characteristics.
| Fig. 2.
Relative probability occurring the health effect, (A) early fatality in early
phase, (B) cancer fatality in early phase, and (C) cancer fatality in chronic
phase, considering the modified shielding factors.
RESEARCH AND INNOVATION 335
Research and Innovation
Off-site Consequence Analysis During Severe Accidents in a Nuclear Power Plant ı Dahye Kwon and Moosung Jae

atw Vol. 65 (2020) | Issue 6/7 ı June/July
RESEARCH AND INNOVATION 336
| Fig. 3.
Relative probability occurring the health effect, (A) early fatality in early
phase, (B) cancer fatality in early phase, and (C) cancer fatality in chronic
phase, considering the modified inhalation rate.
| Fig. 4.
Relative probability occurring the health effect, (A) early fatality in early
phase, (B) cancer fatality in early phase, and (C) cancer fatality in chronic
phase, considering the modified food chain model.
protected. Furthermore, the probabilities
of cancer fatality in early and
chronic phase were decreased as well.
It is because the decreased groundshine
shielding factor is applied to
overall period.
The results of contribution analysis
are depicted in following three figures.
In Figure 3, because of the reduction of
inhalation rate, the probabilities were
decreased. Especially, the decrease of
cancer fatality in early phase was large.
It was because the thyroid cancer was
included to the cancer fatality in early
phase as an important disease.
In Figure 4, the ingestion is a
valid pathway only in the long-term
period, and affects the cancer fatality
in chronic phase. Distinctively, the
pro bability of the consequence was
increased. The reason of this result
was heavy intake of leafy vegetables,
which means Kimchi, the traditional
food in Korea. The leafy vegetables
are relatively easily contaminated
by radioactive materials because they
cannot be protected by the husk.
On the other hand, the probability
of cancer fatality in chronic phase in
Figure 2 was decreased despite the
increased effect of FCM modification.
It resulted from that the consequence
of decontamination workers was
included to that in chronic phase.
Since they reside longer in the contaminated
area, the exposure of
decontamination workers is much
larger than that of public. In addition,
they are exposed by groundshine and
resuspension inhalation only, which
means no impact by FCM modification.
Therefore, the consequence
of cancer fatality in chronic phase in
Figure1 was mostly affected by the
decontamination workers. In calculating
process, there was no way not to
include the decontamination workers
as the exposed people.
4 Conclusion
In this study, as the Level 3 PSA for
Korean NPP has been required to
meet the quantitative safety goal
through the revision of the notification,
several significant input
parameters of MACCS2 were modified
to reflect the Korean socio-environmental
characteristics and regulation
systems for radiation protection. The
parameters were chosen as the DCF,
the shielding factor, the inhalation
rate, and the FCM.
Each parameter was derived to
meet the regulatory standards and
to reflect the Korean environment.
The DCF DB was established corresponding
to regulatory standards.
The shielding factors were calculated
based on that the public live in apartment,
which is the representative
house of Korea. The inhalation
rate was derived based on directly
measured short-term inhalation rate
and activity duration of Korean adults.
Finally, the FCM reflected the Korean
diet habits and agricultural environment.
With the modified parameters, we
conducted a consequence analysis of
OPR-1000, Korean reference NPP and
the results were analyzed. Due to the
reduction of shielding factors and
inhalation rate, the probabilities of
consequences were decreased. Then,
the contribution analysis of modified
parameters was conducted. The consequence
varied depending on the
critical organ of specific disease and
the time period following the severe
accident. On the other hand, against
the prior results, the probability of
consequence in chronic phase was
increased by reflecting the Korean diet
habits which means much intake of
leafy vegetables. The derived results
of this study can be used to improve
the reliability. These results lead to
secure the sufficient margin of safety
assessment. It also means the hope to
encourage the positive opinion of
public.
If additional accident characteristics
such as heat content of the
release segments, release height and
duration, and emergency response
sce nario are additionally included,
more realistic results can be obtained.
Therefore, the results of this study are
expected to contribute to the improvement
of the reliability of the Level 3
PSA.
Acknowledgement
This work was supported by the
Nuclear Safety Research Program
through the Korea Foundation Of
Nuclear Safety (KOFONS), granted
financial resource from the Multi-Unit
Risk Research Group (MURRG),
Republic of Korea (No. 1705001).
References
[1] Chanin DI, Young ML, Randall J. Code manual for MACCS2:
volume 1, User’s guide. Albuquerque: Sandia National
Laboratories; 1997. (SAND97-0594).
[2] Kang T, Jae M. Consequence analysis for nuclear reactors,
Yongbyon. Journal of Nuclear Science and Technology.
2017;54(2):223-232.
[3] Haskin FE, Goossens LHJ, Harper FT, Grupa J, Kraan BCP,
Cooke RM, Hora SC. Probabilistic accident consequence
uncertainty analysis: Early health uncertainty assessment,
Main Report. Washington DC: US Nuclear Regulatory
Commission; 1997. (NUREG/CR-6545, EUR 15855).
[4] Hanyang University, Analysis of the Korean Socio-Environmental
Factors Based on the New ICRP Recommendations,
Daejeon: Korea Atomic Energy Research Institute; 2017.
(KAERI/CM-2398/2016).
[5] International Atomic Energy Agency. Planning for off-site
response to radiation accidents in nuclear facilities. Vienna:
International Atomic Energy Agency; 1979. (IAEA TECDOC-225).
[6] National Institute of Environmental Research. Korean
Exposure Factors Handbook for children. Incheon: National
Institute of Environmental Research; 2016.
[7] Abbott ML, Rood AS. COMIDA: A Radionuclide Food Chain
Model for Acute Fallout Deposition. Health Physics.
1994:66(1):17-29.
[8] Kwon D, Hwang WT, Jae M. Ingestion Dose Evaluation of
Korean Based on Dynamic Model in a Severe Accident. Journal
of Radiation Protection and Research. 2018;43(2):50-58.
[9] Alpert DJ, Chanin DI, Ritchie LT. Relative Importance of Individual
Elements to LWR Accident Consequence Estimates Using
Equal Release Fractions. Nuclear Safety. 1987:28(1): 77-86.
[10] Ahn B, Seo Y, Park H, Jae M. Development of the MSPAR-SITE
Code for Assessing Multi-Unit Risk. Transactions of the Korean
Nuclear Society Spring Meeting; 2018 May 16-18; Jeju, Korea.
[11] Jang SC, Han SJ, Choi SY, Lee SJ, Kim WS. Establishment of
Infrastructure for Domestic-Specific Level 3 PSA based on
MACCS2. Transactions of the Korean Nuclear Society Spring
Meeting; 2015 May 7-8; Jeju, Korea.
Authors
Dahye Kwon
Moosung Jae
jae@hanyang.ac.kr
Department of Nuclear
Engineering
Hanyang University, 222
Wangsimni-ro, Seongdong-gu,
Seoul, 04763, Korea
Research and Innovation
Off-site Consequence Analysis During Severe Accidents in a Nuclear Power Plant ı Dahye Kwon and Moosung Jae

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Code and Data Enhancements of the
MURE C++ Environment for Monte-Carlo
Simulation and Depletion
Maarten Becker
1 Introduction Irradiation calculations with MCNP [1] are nowadays an accepted industry
standard. Since MCNPX v2.6 followed by MCNP v6 depletion functionality is provided by
in-code-coupling of the CINDER code [2]. Depletion statements are placed in the MCNP input and
some depletion cycle dependent manipulation options can be used such as total thermal power of
the systems, nuclide adaptions to given experimental concentrations etc. However, flexible
functionality such as fuel cycle calculations with reshuffling of fuel elements and more complex
simulations are not implemented in MCNP at the moment.
The wrapper code MURE combines a flexible and
extendable tool set for MCNP input generation with a
dedicated customizable depletion functionality ([3,4]).
MURE is written in C++ and inputs are compiled against
the MURE libraries similar to the way GEANT [5] inputs
are pro vided. The drawback (or feature) of the code is that
no dedicated nuclear data is provided but is in the general
responsibility of the user.
This paper describes and summarizes development
work concerning the provision of processed nuclear data
for neutron depletion calculation, especially the handling
of isomeric branching reactions, the implementation of a
CRAM [6] Bateman solver to accelerate the depletion
calculation over the 4 th order Runge-Kutta method.
The enhanced code setup is then applied to the “Isotope
Correlation Experiment” [7] and validated against the
experimental results and code-to-code comparison.
2 Implementation
The implementation – as outlined here – is carried out with
the most recent version of MURE v2 [4].
2.1 CRAM
The Chebyshev Rational Approximation Method was
proposed by Pusa for the solution of the Bateman equation,
since the method is very effective for matrix systems where
their eigenvalues are distributed near the negative real
axis [6]. Several formulations for CRAM exist. Because of
better numerical stability the method of incomplete partial
fractions (IPF) is adapted with order N=48 and the CRAM
coefficients are taken from the respective publication [8].
The IPF algorithm foresees a series of N/2 matrix
inversions, which are of type sparse matrix because of the
nature of the burn-up (transmutation) matrix. The matrix
inversion can be done by a Gauss-Seidel iteration or – in
this case – by the solver SparseLU of the Eigen library
v3.3.7 [9]. The sparse solver provides the method
compute() to calculate the matrix inversion A -1 as in
A x = b and the method solve() to give the solution of
A -1 b=x in matrix notation.
The algorithm applied is:
Variable
Meaning
Planned entry for
y Nuclide vector initialized by vector N0 at t=0
matrix
dt
alpha[i], theta[i],
alpha0
SpEye
x
Compressed sparse transmutation matrix
Time step of depletion
Coefficients of CRAM solution
Sparse identity matrix
Temporary solution vector
| Tab. 1.
Variables of the CRAM implementation.
The theoretical accuracy of CRAM has been intensively
investigated by Pusa [8,10]. As a consequence CRAM of
order 48 achieves accurate results for number densities
above 1E-20 atoms/cm 3 .
As a more practical approach to validate the implementation,
the results of the Runge-Kutta solver and the new
CRAM solver are compared when exactly the same
transmutation matrix is given as input. The transmutation
matrix corresponds to the KWO ICE fuel of the first
irradiation step that contains 2512 isotopes with half-lives
T ½ >= 1 s. Activation cross sections are condensed from
the TENDL Ace library which is described in section 2.3.
Both solvers act on the full transmutation matrix without
application of isotope saturation as e.g. the ORIGEN family
of codes does.
For all isotopes with a number density N > 1E-20 # /cm 3
the difference cannot be detected between Runge-Kutta
and CRAM for numbers given at 5 digits accuracy. The
speed-up of CRAM for this case is about 16.
However, the depletion algorithm in MURE is still
designed for the Runge-Kutta method, i.e. at least
5 adaptive time sub-steps are applied for each time step to
achieve reliable results. Since the accuracy of CRAM of
order 48 is not as sensitive to the time step [8], less
sub-steps could be used. The matrix inversion then needs
only be done one time for the first sub-step and can be
reused for any further sub-step. This will render the
speed-up even more impressive.
Best Paper
Award
The paper “Code and
data enhancements
of the MURE C++
environment
for Monte-Carlo
simulation and
depletion” by
Dr. Maarten Becker
and “A geopolymer
waste form for
technetium, iodine
and hazardous
metals” by
Werner Lutze,
Weiliang Gong,
Hui Xu and
Ian L. Pegg (will
be featured in a
future atw) have
been awarded
as Best Papers of
KERNTECHNIK 2020,
which unfortunately
had to be cancelled
due to Covid-19.
RESEARCH AND INNOVATION 337
denoting that the Eigen library allows for elegant vector
and matrix notation with variables meaning as in Table 1.
2.2 Implementation of isomeric state branching
The original MURE code has implemented hard coded
isomeric branching in activation reactions only for some
isotopes. To provide a general isomeric branching scheme,
the code is modified to read in a general table of branching
ratios for any isotope and activation reaction type.
Research and Innovation
Code and Data Enhancements of the MURE C++ Environment for Monte-Carlo Simulation and Depletion ı Maarten Becker

atw Vol. 65 (2020) | Issue 6/7 ı June/July
RESEARCH AND INNOVATION 338
Time
step
Duration
(d)
Z t A t I t Z p A p I p MT BR th BR f
95 241 0 95 242 1 102 0.13 0.15
95 242 0 95 242 1 4 0.09 0.06
95 243 0 95 242 1 16 0.26 0.26
95 244 0 95 242 1 17 0.24 0.25
| Tab. 2.
Isomer production of Am-242m.
Uranium fuel
Enrichment (%} 3.1
Temperature (K) 1028
Radius (cm) 0.465
Number densities
U-235 7.12E-04
U-238 2.20E-02
0-16 4.54E-02
Material
Clad
| Tab. 3.
KWO ICE pin cell definition from Broeders [18].
| Tab. 4.
Power and boron history according to Cao [21]; power is relativ to 219.6 W/cm linear power
as defined in [18].
Zirconium
Temperature (K) 605
Radius (cm) 0.535
Zr
Number densities
Moderator
4.33E-02
Temperature (K) 572
Radius (cm) 0.8449
Number densities
H-1 4.81E-02
0-16 2.40E-02
B-10 7.74E-06
B 10
(10 -6+ )
Power
(++)
Time
step
Duration
(d)
B 10
(10 -6+ )
Power
(++)
1 5.8 7.738 1.0 2 1.0 7.738 0.0
3 4.6 7.756 1.0 4 50.0 6.6836 1.0
5 25.0 5.831 1.0 6 2.0 5.831 0.0
7 3.5 5.395 1.0 8 30 5.031 1.0
9 41.5 5.031 0.0 10 6.5 4.495 1.0
11 50 3.726 1.0 12 75 2.027 1.0
13 5.8 2.027 0.0 14 5.9 0.7605 1.0
15 31 0.2558 1.0 16 28 0.2558 0.0
17 6.9 7.494 1.0 18 30 6.663 1.0
19 30 6.663 0.961 20 60 5.447 1.0
21 9.2 5.447 0.0 22 4.7 4.686 1.0
23 40 4.249 1.0 24 40 3.414 1.0
25 3.5 3.414 0.0 26 3.0 2.891 1.0
27 20 2.651 1.0 28 3.0 2.651 0.0
29 4.0 2.338 1.0 30 56 1.711 1.0
31 13.8 1.404 1.0 32 380 1.404 0.0
33 5.3 6.705 1.0 34 65 5.734 1.0
35 60 3.978 1.0 36 3.0 3.978 0.0
37 3.4 3.0 1.0 38 50 2.248 1.0
39 50.0 0.6976 1.0 40 365 0.6976 0.0
The new table contains the data of Table 2 and shows
as an example the production (index p) of AM-242m from
target (index t) isotopes Am-241, Am-242g, Am-243,
Am-244 via the capture (102), inelastic scattering (4),
n2n (16), n3n (17) neutron reactions, respectively.
The branching ratio is given for a thermal LWR neutron
spectrum (index th) and a fast spectrum (index f).
To use the branching data the corresponding reaction
type must be available in the Ace Monte-Carlo library. This
might be not the case for some total reactions like inelastic
scattering which is the sum of all inelastic levels and
continuum stored as MT 51–91.
To generate this table, the nuclear data processing code
PREPRO [11] is used to generate one group activation
cross sections for any isotope of the TENDL [12] nuclear
data evaluation. Branching ratios are then calculated
from the relation
where σ is the one group cross section which leads to
isomeric state i (ground, 1 st , 2 nd , … state) of the target
nucleus.
In PREPRO the weighting spectrum options of the
NJOY code [13] for spectrum type 5 (EPRI CELL) LWR
and type 8 fast neutron spectrum were applied. In total,
30562 reactions of almost 3000 isotopes leading to
isomeric states are generated.
In the course of the depletion calculation MURE
requests reaction rates from the MCNP run and applies the
branching ratios depending on the target and production
isotope to fill the transmutation matrix.
2.3 Nuclear data
MURE allows for prioritizing the nuclear data sources.
If data is requested and not available in the evaluation
of highest priority, the next defined source is researched.
As nuclear data with highest priority in all further
calculations ENDF/B VII.1 [14] is used. The TENDL library
is taken, if isotope reaction data is not available in
ENDF/B VII.1.
Both nuclear data libraries have been processed with
the nuclear data processing code NJOY 2016 [13]. It was
necessary to patch the NJOY code, to get the total inelastic
scattering MT4 cross section under all circumstances
into the Ace library, where otherwise only the level data
MT51–91 is delivered.
The ENDF/B VII.1 cross section data has been processed
in temperature steps of 50 K from 300 K, whereas the
TENDL library only for the temperatures 500 K and 900 K.
With help of MURE utility codes also the fission product
distributions for all fissile isotopes were included into the
MURE data set.
Radioactive decay data was used given by MURE and
is derived from JEFF 3.1.1 data [15]. The collection of
isotopes of the decay data files define the full set of isotopes
of which the evolution will be simulated.
3 Validation
The KWO ICE benchmark defines a simple fuel rod with
fresh Uranium dioxide of 3.1% enrichment. Further details
are given in Table 3 and the irradiation history is shown in
Table 4. The achieved burn-up is about 29 GW d/tHM.
After about one year irradiation at full power, the fuel
was removed from the core and placed back after 380 d.
The experimental results were determined by four
independent laboratories.
Research and Innovation
Code and Data Enhancements of the MURE C++ Environment for Monte-Carlo Simulation and Depletion ı Maarten Becker

atw Vol. 65 (2020) | Issue 6/7 ı June/July
| Fig. 1.
U-235 atoms/initial HM atoms.
| Fig. 2.
U-238 atoms/initial HM atoms.
RESEARCH AND INNOVATION 339
| Fig. 3.
PU-238 atoms/initial HM atoms.
| Fig. 4.
PU-239 atoms/initial HM atoms.
| Fig. 5.
XE-131/XE-134 atoms ratio.
| Fig. 6.
ND-146/ND-145 atoms ratio.
The fuel rod is modeled with the MURE code to generate
automatically a MCNP input, where the material data
is always updated according to the power history. The
evolution scheme is that of constant power irradiation
without predictor-corrector scheme. The reaction rates
are not directly calculated by tallies in MCNP but are
constructed from integrating a very fine flux tally result
(17901 groups) of the cell together with the Ace reaction
cross section. Only for U-238, effective reaction rates are
tallied directly in MCNP. This shortens the simulation time
of MCNP dramatically without losing much of accuracy in
terms of burn-up.
The result of MURE is not only compared to experimental
values but also to the result of the deterministic KAPROS
code ([16,17]). KAPROS has a long history of nuclear data
assessment and simulation of advanced reactor systems.
A standard test case for new developments is the KWO ICE
benchmark. Results of KAPROS have been described by
Broeders [18], Send [19] and Kern [20], independent
analysis were performed by Cao [21], Hesse [22] and
Fischer [23]. The used KAPROS results are recent
evaluations carried out by Broeders [24].
The applied KAPROS data base for burn-up calculation
is the JEFF 3.1.1 nuclear data evaluation [15] together
with the radioactive decay and activation data [25]. The
neutron flux solver is a discrete ordinate 1-D code based on
ANISN [26], the burn-up module stems from the ORIGEN
code [27].
The KWO benchmark provides 31 experimental results
for isotopes and isotope ratios. Estimated errors are given
for some actinides only. The fuel burnup was determined
with a spread of -4.4 % – +3.4 % [23]. Overall, very good
Research and Innovation
Code and Data Enhancements of the MURE C++ Environment for Monte-Carlo Simulation and Depletion ı Maarten Becker

atw Vol. 65 (2020) | Issue 6/7 ı June/July
RESEARCH AND INNOVATION 340
agreement with experiment and code comparison is
observed. In the next figures dedicated examples are shown
for main isotopes U-235 (Figure 1) and U-238 ( Figure 2)
with very good reproduction of experimental values
within the error bars. The prediction of PU-238 (Figure 3)
by MURE is enhanced compared to the KAPROS,
whereas the results for PU-239 (Figure 4) are of almost
equal agreement. As examples for fission product ratios
XE-131/XE-134 (Figure 5) and ND-146/ND-145 ( Figure 6)
show significant enhancement in the prediction of the
experimental values.
4 Summary
The purpose of this work is to provide an accurate and
efficient depletion solver within the MURE code. The
functionality and accuracy was tested against the 4 th order
Runge-Kutta solution of MURE. Considerable speed-up
was achieved without losing accuracy.
Moreover, nuclear data enhancement in handling of
isomeric state branching was implemented which allows
to take into account the full reaction data given by
dedicated activation or the TENDL library.
The complete MURE setup was then tested against the
KWO ICE benchmark. The results are very promising,
although an in-depth analysis of the impact of alternative
cross section data should be performed to learn the root
cause of the observed enhancements in the isotope
prediction.
[22] Hesse, U.: Verification of the OREST (HAMMER-ORIGEN) depletion program system using
post-irradiation analyses of fuel assemblies 168, 170, 171 and 176 from the Obrigheim Reactor
(Nr. ORNL/tr-88/20; GRS-A-962): Gesellschaft fuer Reaktorsicherheit (GRS) mbH, Garching
(Germany), 1984
[23] Fischer, Ulrich; Wiese, Hans-Werner: Verbesserte konsistente Berechnung des nuklearen Inventars
abgebrannter DWR-Brennstoffe auf der Basis von Zell-Abbrand-Verfahren mit KORIGEN,
Kernforschungszentrum Karlsruhe, KfK-3014. Karlsruhe: Kernforschungszentrum Karlsruhe, 1983
[24] Broeders, Cornelis H.M.; Cao, Yan; Gohar, Yousry; Alvarez-Velarde, F.: Reactor Fuel Burn-up
Qualification / Validation of the Isotope Correlation Experiment in NPP Obrigheim
(not published), IAEA CRP ADS Research. Vienna, 2010
[25] Koning, A.; Forrest, R.; Kellett, M.; Mills, R.; Henriksson, H.; Rugama, Y. (Hrsg.): The JEFF-3.1
Nuclear Data Library, JEFF Report 19, NEA No. 3711: OECD, 2005 — ISBN 92-64-01046-7
[26] Becker, Maarten: Determination of kinetic parameters for monitoring source driven subcritical
transmutation devices, Universität Stuttgart (2014)
[27] Bell, M. J.: ORIGEN: The ORNL Isotope Generation and Depletion Code, ORNL-4628:
Oak Ridge National Laboratory, 1973
Author
Dr. Maarten Becker
iUS Institut für Umwelttechnologien und Strahlenschutz GmbH
becker@ius-online.eu
Obernauer Straße 94
63743 Aschaffenburg, Germany
References
[1] Werner, C. J. (Hrsg.): MCNP Users Manual – Code Version 6.2, Los Alamos National Laboratory,
LA-UR-17-29981, 2017
[2] Fensin, Michael L; James, Michael R; Hendricks, John S; Goorley, John T: The New MCNP6 Depletion
Capability. In: International Congress on Advances in Nuclear Power Plants (ICAPP), 2012, S. 10
[3] Méplan, O.; Nuttin, A.; Laulan, O.; David, S.; Michel-Sendis, F.; Wilson, J.: MURE: MCNP Utility for
Reactor Evolution – Description of the methods, first applications and results. In: European
Nuclear Society, 2005
[4] Méplan, O.; Hajnrych, Jan; Bidaud, A.; David, S.; Capellan, N.; Leniau, B.; Nuttin, A.;
Havluj, Frantisek; u. a.: MURE 2: SMURE, Serpent-MCNP Utility for Reactor Evolution User Guide –
Version 1 (report): Laboratoire de Physique Subatomique et de Cosmologie, 2017
[5] Agostinelli, S.; Allison, J.; Amako, K.; Apostolakis, J.; Araujo, H.; Arce, P.; Asai, M.; Axen, D.; u. a.:
Geant4—a simulation toolkit. In: Nuclear Instruments and Methods in Physics Research Section A:
Accelerators, Spectrometers, Detectors and Associated Equipment Bd. 506 (2003), Nr. 3, S. 250–303
[6] Pusa, Maria: Numerical methods for nuclear fuel burnup calculations: Aalto University, 2013 –
ISBN 978-951-38-8000-2
[7] Koch, L.; Schoof, S. (Hrsg.): The isotope correlation experiment, ICE. Final report
(Nr. KfK-3337 EUR 7766 EN ESARDA 2/81): Institut für Radiochemie (IRCH), 1982
[8] Maria, Pusa: Higher-Order Chebyshev Rational Approximation Method and Application to Burnup
Equations. In: Nuclear Science and Engineering Bd. 182 (2016), Nr. 3, S. 297–318
[9] Guennebaud, Gaël; Jacob, Benoît; others: Eigen v3. URL http://eigen.tuxfamily.org
[10] Pusa, Maria; Leppänen, Jaakko: Computing the Matrix Exponential in Burnup Calculations.
In: Nuclear Science and Engineering Bd. 164 (2010), Nr. 2, S. 140–150
[11] Cullen, D. E.: PREPRO 2019: ENDF/B Pre-processing Codes, IAEA-NDS-39, Rev. 19, 2019
[12] Koning, A. J.; Rochman, D.; van der Marck, S. C.; Kopecky, J.; Sublet, J. Ch.; Pomp, S.; Sjostrand, H.;
Forrest, R.; u. a.: TENDL-2014: TALYS-based evaluated nuclear data library.
[13] MacFarlane, R. E.; Kahler, A. C. (Hrsg.): The NJOY Nuclear Data Processing System,
Version 2016, LA-UR-17-20093, 2019
[14] Chadwick, M. B.; Herman, M.; Obložinský, P.; Dunn, M. E.; Danon, Y.; Kahler, A. C.; Smith, D. L.;
Pritychenko, B.; u. a.: ENDF/B-VII.1 Nuclear Data for Science and Technology.
In: Nuclear Data Sheets Bd. 112 (2011), Nr. 12, S. 2887–2996
[15] Santamarina, A.; Bernard, D.; Blaise, P. (Hrsg.): The JEFF-3.1.1 Nuclear Data Library,
JEFF Report 22, NEA No. 6807: OECD, 2009 — ISBN 978-92-64-99074-6
[16] C.H.M. Broeders; R. Dagan; V. Sanchez; A. Travleev: KAPROS-E: Modular Program System
for Nuclear Reactor Analysis, Status and Results of Selected Applications. In: 2004
[17] Becker, M.; Criekingen, S. V.; Broeders, C. H. M.: The Karlsruhe Program System KAPROS and
its successor the Karlsruhe Neutronic Extendable Tool KANEXT, 2013
[18] Broeders, C. H. M.: Entwicklungsarbeiten fuer die neutronenphysikalische Auslegung von
Fortschrittlichen Druckwasserreaktoren (FDWR) mit kompakten Dreiecksgittern in hexagonalen
Brennelementen, University of Karlsruhe, Dissertation, 1992
[19] Send, Ludwig: Investigations for Fuel Recycling in LWRs. Karlsruhe, University of Karlsruhe,
Diplomarbeit, 2005
[20] Kern, Kilian: Advanced Treatment of Fission Yield Effects and Method Development for
Improved Reactor Depletion Calculations, Karlsruher Institute of Technology, Dissertation, 2019
[21] Cao, Yan; Gohar, Yousry; Broeders, Cornelis H.M.: MCNPX Monte Carlo burnup simulations
of the isotope correlation experiments in the NPP Obrigheim. In: Annals of Nuclear Energy
Bd. 37 (2010), Nr. 10, S. 1321–1328
Research and Innovation
Code and Data Enhancements of the MURE C++ Environment for Monte-Carlo Simulation and Depletion ı Maarten Becker

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Modelling Thermal-hydraulic Effects
of Zinc Borate Deposits in the PWR Core
After LOCA – Experimental Strategies
and Test Facilities
Wolfgang Kästner, Sören Alt, André Seeliger, Frank Zacharias, Ulrich Harm, René Illgen, Uwe Hampel
and Holger Kryk
1 Introduction During the sump recirculation phase after loss-of-coolant accidents (LOCA) in
pressurized water reactors (PWR), coolant outpouring of the leak in the primary cooling circuit will
take place (see Figure 1).
| Fig. 1.
Scheme of a PWR LOCA scenario including locations of zinc corrosion and zinc borate deposition effects.
The collected coolant in the reactor
sump will be recirculated to the reactor
core by residual-heat removal pumps
as part of the emergency core cooling
system (ECCS). The long-term contact
of the boric acid containing coolant
with hot-dip galvanized containment
internals (e.g. grating treads, supporting
grids of sump strainers) is assumed
to cause corrosion of the corresponding
materials with the consequence of
rising concentrations of dissolved zinc
(Zn) in coolant. As it was shown in
previous research projects, the subsequently
formed zinc borates (ZnB)
have a retrograde solubility with
increasing temperatures, which could
lead to zinc borate precipitations
(ZBP) in hot spots of the reactor core
in the later stage of the sump recirculation
operation [1-8].
Generic experimental investigations
including the analysis of such Zn
corrosion processes with sub sequent
ZBP in the reactor core have been
started as joint research project of the
Helmholtz-Zentrum Dresden-Rossendorf
(HZDR), TU Dresden (TUD),
and Zittau-Görlitz University of
Applied Sciences (HSZG). The aim is
to provide data sets and correlations
in order to build up a realistic data
based computer simulation tool
( extensions of the ATHLET code) for
this processes by the further project
partner GRS.
Planned entry for
2 Objectives of the project
The German software tool ATHLET
(Analysis of THermohydraulics of
Leaks and Transients) is continuously
being developed for the simulation
of plant behaviour in the event of
transients and accidents. The focus of
the current project named “ATHLET-
Modul Zinkborat” (AZora) is the
development and validation of an
ATHLET module on the basis of the
current state of research on chemical
long-term effects according to PWR
LOCA. The module shall be used for
p resilient deterministic safety
assess ments of PWR plants,
p simulations under consideration of
treatments of the consequences of
an accident (e.g. inclusion of the
coolant purification system for Zn
removal) and
p simulations of scenarios considering
the unavailability of
measures for the treatment of
accident consequences.
The ATHLET module to be developed
consists of different partial models
(PM) for the processes of release and
precipitation/accumulation, while the
transport is realised by increasing
the material flow balances (see
Figure 2):
p PM “release” represents the release
of ionic Zn into the coolant by
corrosion of the galvanized surfaces
(e.g. gratings, platforms,
supporting grids) in the reactor
sump. Model input parameters will
e.g. be the local volume flow,
which represents flow conditions
near the corroding surface, and
the position of the Zn source in
the PWR sump (see [5] for the
corresponding categories).
p PM “precipitation” simulates the
precipitation and deposition of
ZnB as a function of local parameters
such as coolant temperature,
concentration of Zn and
RESEARCH AND INNOVATION 341
Research and Innovation
Modelling Thermal-hydraulic Effects of Zinc Borate Deposits in the PWR Core After LOCA – Experimental Strategies and Test Facilities
ı Wolfgang Kästner, Sören Alt, André Seeliger, Frank Zacharias, Ulrich Harm, René Illgen, Uwe Hampel and Holger Kryk

atw Vol. 65 (2020) | Issue 6/7 ı June/July
RESEARCH AND INNOVATION 342
| Fig. 2.
Input and output variables as well as partial models of the ATHLET module “Zinc borate”.
boron and coolant velocity. The
ATHLET model consideration of
growing ZnB layers and their influence
on the thermal hydraulics
in the PWR core makes it necessary
to measure layer thicknesses and
their thermal impact online.
p For PM “transport”, the existing
multi-component material flow
balances are going to be extended
in such a way that the transport of
two or more substances (Zn and
mobile ZnB) within the coolant
flow can be considered simultaneously
without affecting the
thermal hydraulics [9].
p PM “separation” takes into account
a removal of released Zn ions
from the coolant, which can be
achieved by accident follow-ups
using the coolant cleaning system
(ion exchangers). In ATHLET, a
GCSM-based FILL model will act as
a sink term [9].
3 Concept development
The influence of flow conditions like
air entrance and leak flow rate on Zn
corrosion in the sump was sufficiently
documented. Therefore, parameterization
of PM “release” will be based
on the results of an explorative data
analysis, applied on the experimental
database created during previous
experiments [4,7]. Despite the generic
character of the planned experiments,
transformation of known boundary
conditions from technical to semitechnical
scale is necessary for determining
p course of coolant temperatures at
simulated sump and core,
p course of heating power of fuel rod
simulators,
p corrosion inventory, estimated on
the basis of the corrosive surfaces
inside the containment exposed
to the coolant and the average
thickness of Zn coatings and
p maximum Zn concentration in the
coolant, based on the circulating
coolant volume and the corrosion
inventory available (see [5]).
When assessing the impact of ZnB on
core cooling, its different appearances
must be taken into account: the layerforming
ZnB (see Figure 3) as well as
the mobile ZnB in smaller particle
sizes, which can only be recognised as
turbidity of the coolant.
| Fig. 3.
Microscopic image of a solidified ZnB layer
with ca. 300 µm thickness, taken from
cladding tube.
For the first of the two, preferably
image-based measuring technologies
have to be implemented or enhanced,
which make some requirements
with regard to the observability of
ZBP during experimental operation.
Furthermore, the heat transfer
proper ties of the porous ZnB
layers must be determined experimentally
depending on their thermalhydraulic
and chemical formation
conditions.
Mobile ZnB was often determined
as light adhesions and sediments in
passive downstream components. Its
complete removal from the coolant
and its balancing was not possible
before. For the planned empirical
parameter determination for PM
“transport”, it has to be considered in
the test rig design.
The chemical boundary conditions
of 2000 ppm boron and 0.2 ppm
lithium (Li) correspond to the average
coolant chemistry occurring during
LOCA [8]. For this coolant chemistry,
the dependency between the electrical
conductivity of the coolant and
the additional Zn concentration due
to corrosion must be determined
empirically for online measurements
of Zn concentrations in the experimental
facilities.
Other components of a reactor
pressure vessel than fuel rods and
spacers should not be considered
experimentally.
4 Resulting design of
experimental facilities
4.1 Core simulator design
at semi-technical scale
The high requirements gave rise to the
design and construction of the core
simulator THETIS (Twofold HEaTIng
rod configuration for core Simulation),
which includes a double 3×3
sub-geometry of PWR core, where
both channels are connected with a
transverse flow channel at spacer
height (see 1-3 in Figure 4).
For their technical parameters,
existing ATHLET calculation results
Research and Innovation
Modelling Thermal-hydraulic Effects of Zinc Borate Deposits in the PWR Core After LOCA – Experimental Strategies and Test Facilities ı
Wolfgang Kästner, Sören Alt, André Seeliger, Frank Zacharias, Ulrich Harm, René Illgen, Uwe Hampel and Holger Kryk

atw Vol. 65 (2020) | Issue 6/7 ı June/July
| Fig. 4.
3D scheme of pipe system and housing of the core-representing test rig THETIS, containing heating
sections with 3×3 core subgeometries 12 and a cross flow section 3.
from GRS for 15 LOCA scenarios were
taken up and evaluated in detail [10].
In THETIS, the fuel rods are simulated
by heating rods with heating power
equally distributed over the heating
length. Their cladding tubes consist of
Zircaloy cladding tubes (Ø 10.75 mm).
The lower ends are sealed with final
caps, applied by resistance pressure
welding. The rod configurations are
partially equipped with 3×3 segments
of a 16×16 spacer of type HTP.
Each 3×3 core subgeometry is
separately enclosed as a channel in a
stainless steel housing with several
observation windows. Each channel
has its own inlet and outlet. The
coolant supplied can be heated up by a
preheater component. This subsystem
as shown in Figure 4 becomes part
of a whole coolant circuit, in which
the test rig “Zittau flow tray” acts
as a PWR sump simulator: Here, as a
part of the test setup, up to 16 m 3 of
coolant can be enriched with ionic
zinc, boric acid and LiOH.
The thickness of ZnB layer and the
percentage of the visible area (outer
surface) of the spacer covered with
mobile ZnB can be optically determined
online. With knowledge of
an average layer thickness and the
density of the ZnB, the mass of the
attached ZnB can be approximated.
Furthermore, the ZnB is completely
removed after each experiment and
the total dry mass is determined.
The layer surface condition can be
determined at the end of the test by
measuring layer fragments under a
microscope, e.g. by extreme values of
locally measured layer thicknesses.
In addition to this, the rig is
equipped with a filtering system in
downstream direction, containing
fine filter cartridges with a mesh size
of 1 µm. This redundantly designed
system allows the mass balancing of
the mobile parts of ZnB in the coolant.
Several taps allow sampling and
chemical analysis during an experiment
in progress.
Any specification and thermalhydraulic
influence induced by ZnB
precipitations will be measured technically,
e.g.:
p precipitation rate of layer-forming
ZnB at the cladding tubes: With the
knowledge of the ZnB bulk density,
the total mass of the ZnB layer can
be approximated by the optically
determined layer thickness.
p total dry mass of layer-forming
ZnB: by removal, tempering and
weighing of ZnB at the end of an
experiment
p precipitation rate of mobile ZnB in
the coolant: by periodic removal of
the fine filters, tempering and mass
balance
p formation rate of ZnB in the
simulated core: approximation of
the masses of ZnB produced by the
evaluation of the Zn concentration
in the coolant to be recorded at
core inlet and outlet
p differential pressure: detection
with pressure sensors placed at the
spacer segments
Attributes of the coolant itself will be
measured, e.g.:
p horizontal transversal flows by
ultrasonic flow measurement
p inlet/outlet flows by electromagnetic
flow meter
p pH value by sample taking and
analysis
p online/offline measurement of
electrical conductivity, which is an
indicator of Zn concentration for a
defined coolant chemistry
For the ATHLET module, these
experi mental data supports the
implemen tation of balance equations
for Zn and ZnB. The hydraulic consequences
of ZnB precipitations will be
considered by dynamically adjusted
drag coefficients at the spacers, in
connected objects in axial direction,
and in parallel- connected objects in
radial direction [9]. The data basis,
which should enable the modelling of
the resulting thermal effects, is provided
by tests at laboratory scale of
the project partners HZDR and TUD.
4.2 Coolant loop design
at laboratory scale
The main investigations on ZBP in
boric acid containing PWR coolants
and ZnB deposition at hot surfaces of
PWR fuel rod cladding tubes are
carried out in a modified KorrVA test
facility representing the ECCS during
sump recirculation operation in a very
simplified manner [2]. In particular,
this includes the determination of the
following parameters:
p ZnB deposition rates at hot Zry
surfaces (3 dimensional growth
rate of ZnB layers) depending
on thermal- hydraulic and water
chemical parameters,
p roughness of ZnB layers,
p formation rates of ZnB particles
in the coolant and
p heat transfer properties of the
ZnB layers depending on their
formation conditions.
A simplified scheme of the lab-scale
facility is shown in Figure 5.
Basically, it consists of a zinc
dissolution unit (including flowed
basket with zinc granules) and a bath
section (representing sump / coolant
reservoir). A heat exchanger with
thermostat heats up the coolant to a
defined temperature. The courses of
the fluid temperatures, flow rates and
Zn concentrations are monitored
online during the experiments by
| Fig. 5.
Simplified scheme of modified laboratory corrosion test facility (KorrVA).
RESEARCH AND INNOVATION 343
Research and Innovation
Modelling Thermal-hydraulic Effects of Zinc Borate Deposits in the PWR Core After LOCA – Experimental Strategies and Test Facilities
ı Wolfgang Kästner, Sören Alt, André Seeliger, Frank Zacharias, Ulrich Harm, René Illgen, Uwe Hampel and Holger Kryk

atw Vol. 65 (2020) | Issue 6/7 ı June/July
RESEARCH AND INNOVATION 344
| Fig. 7.
Reaction calorimeter RC1e.
| Fig. 6.
Design of the flow channel measurement
system (FCMS) to investigate the growth of
ZnB layers on hot Zry surfaces and the heat
transfer properties of the ZnB layers.
means of integrated temperature and
flow sensors and by electrical conductivity
sensors, respectively, using
correlations between conductivity
and corresponding zinc concentration
at a distinctive temperature. After the
experiments, the courses of the Zn
concentrations will additionally be
determined by analysis of liquid
samples taken during the experiments,
e.g. by ICPMS (inductively
coupled plasma mass spectrometry).
Details of the experimental facility as
well as of the analytical methods can
be found in [2].
To determine the above-named
parameters by optical and calorimetric
methods, an extension of the
KorrVA test facility using the flow
channel measurement system (FCMS)
is under construction.
In Figure 6, a simplified model of
the FCMS as main part of the KorrVA
coolant loop is shown. The aim is
to execute generic experiments regarding
the dependency of ZnB layer
formation rate on thermal and fluid
dynamic parameters and to evaluate
the dependency of the heat transfer
coefficient (formed ZnB layer) on ZnB
formation parameters.
The thickness, profile and surface
structure of the ZnB layers
formed on the hot surface of the
electrically heated Zircaloy block
(representing the cladding tube wall)
can be measured by means of profile
measure ments using a laser measurement
system (laser triangulation
displacement sensor). Any mobile
ZnB particles forming in the coolant
or spalling from the ZnB layer are
collected in a filter downstream the
FCMS. The contents of Zn and boron
in the precipitated ZnB can be determined
analytically. Deposition rates of
layer-forming ZnB (on the surface)
and mobile ZnB can be estimated as a
function of thermal hydraulic ( fluid
temperature, surface temperature,
Reynold number) and chemical parameters
(Zn concen tration).
The cell also allows in-situ measurements
of the heat transfer through
the ZnB layers in order to derive
statements on the thermal conductivity
of ZnB layers and on the convective
heat transfer between the ZnB layer
and the coolant. The usage of heat
flow calori metry is intended for this
purpose.
4.3 Investigations on zinc
solubility at laboratory
scale
A series of ZBP experiments (formation
of mobile ZnB particles in
coolant) is planned to be conducted in
a reaction calorimeter (see Figure 7)
to determine the following parameters
necessary for the simulation of
ZBP processes:
p solubility of Zn in typical PWR
coolants depending on the coolant
temperature and
p nucleation behavior of ZnB crystals
in PWR coolants.
The calorimeter consists of a stirred
tank reactor having a volume of 1.8 L,
where its temperature is controlled by
a double jacket. During the experiments,
the course of fluid temperature
is controlled by a computer program
and the fluid temperature as well
as the electrical conductivity are
monitored online. Additionally, the
courses of the Zn concentrations
are determined by analysis of liquid
samples using ICPMS.
5 Summary
In coordination with all project
participants, the experimental parameter
intervals, the transfer parameters
relevant for the interface
between experiment and simulation
as well as the representative reference
parameters to be additionally included
in the test matrix were defined. The
dependencies of all parameters to be
simulated were jointly defined in
several discussions and possible
model representation in ATHLET was
evaluated.
In addition, the following provisions
were made with regard to
p material property data and quantities,
e.g. of the zinc inventory in
the PWR affected by corrosion
p maximum time frame after LOCA
to be covered simulatively
p PWR core areas to be considered
p core geometries and reactor
pressure vessel internals to be
simulatively included in the ZnB
problem
Internally, the following definitions
were made:
p Principle test sequences for experiments
at laboratory and semitechnical
scale
p Methodology for filtering and
balancing mobile parts of the ZnB
6 Outlook
In the next step, the model development
for position- and area-related
corrosion rates for Zn inventory under
LOCA conditions (PM “release” acc. to
Figure 2) will be continued, including
experimental data from earlier experiments
of the HSZG aiming at secured
sump suction. Experimental and theoretical
work for model development
for the simulation of ZnB precipitates
and deposits in the PWR core under
LOCA conditions will start after extension
of the core simulator THETIS
and lab-scale facility KorrVA for the
measure ment of thermal-hydraulic
parameters at hot Zry surfaces, cladding
tubes and spacer. Furthermore,
experiments will take place at THETIS
to assign the determined layer thicknesses
to local differential pressures
and flow vectors. Here, the formulation
and parameterization of the
models for the simulation of thermalhydraulic
consequences of ZnB depositions
in the coolant, on cladding tubes
and on spacer stands as final result.
LOCA-related, combined Zn release
and ZnB separation experiments on a
Research and Innovation
Modelling Thermal-hydraulic Effects of Zinc Borate Deposits in the PWR Core After LOCA – Experimental Strategies and Test Facilities ı
Wolfgang Kästner, Sören Alt, André Seeliger, Frank Zacharias, Ulrich Harm, René Illgen, Uwe Hampel and Holger Kryk

atw Vol. 65 (2020) | Issue 6/7 ı June/July
semi-technical scale are planned for
the validation of the ATHLET module.
For this, the existing core simulator
CORVUS [10], which represents a
3×3 PWR core subgeometry in original
length and with a cosine-shaped power
distribution, will be enhanced with
horizontal channels for ZnB-induced
transversal flows.
Both, the direct experimental
results of the project and the models
and simulation resulting therefrom can
be used for the safety assessment and
optimisation of the plants by licensing
authorities and operators. This leads to
an increased range of applications for
the ATHLET simulation tool, including
the non-nuclear engineering sector, in
which crystalline layer growth plays a
significant part.
Acknowledgements
The reported investigations of the project
“Generische thermohydraulische
und physikochemische Analysen zur
Implementierung eines ATHLET-
Moduls für die Simulation thermohydraulischer
Folgen von Zinkborat-
Ablagerungen im DWR-Kern/Kurz titel:
ATHLET-Modul Zinkborat ( AZora)” are
funded by the German Federal Ministry
for Economic Affairs and Energy
( BMWi) under the grant nrs.
1501585A, 1501585B, and RS1571 on
the basis of a decision by the German
Bundestag. The responsibility for the
content of this publication lies with the
authors.
References
[1] Kryk, H.; Hoffmann, W.: Partikelentstehung und -transport
im Kern von Druckwasserreaktoren – Physikochemische
Mechanismen. Final report of BMWi project grant
no. 1501430, 2014
[2] Hampel, U.; Harm, U.; Kryk, H.; Ding, W.; Wiezorek, M.; Unger,
S.: Lokale Effekte im DWR-Kern infolge von Zinkborat-
Ablagerungen nach KMV, Final report to BMWi project grant
no. 1501496, 2019
[3] Kryk, H.; Harm, U.; Hampel, U.: Reducing in-core zinc borate
precipitation after loss-of-coolant accidents in pressurized
water reactors, Proceedings of the Annual Meeting on Nuclear
Technology (AMNT 2016), Hamburg, 2016
[4] Seeliger, A.; Alt, S.; Kästner, W.; Renger, S.; Kryk, H.; Harm, U.:
Zinc corrosion after loss-of-coolant accidents in pressurized
water reactors – thermo- and fluid-dynamic effects. Nuclear
Engineering and Design, 2016, 305, 489-502
[5] Alt, S.; Kästner, W.; Renger, S.: Safety-related analysis of
corrosion processes at zinc-coated installations inside the
PWR sump. Proceedings of the Annual Meeting on Nuclear
Technology (AMNT), Berlin, 2017
[6] Harm, U.; Kryk, H.; Hampel, U.; Generic zinc corrosion studies
at PWR LOCA conditions. Proceedings of the 48 th Annual
Meeting on Nuclear Technology (AMNT 2017), Berlin, 2017
[7] Renger, S.; Alt, S.; Gocht, U.; Kästner, W.; Seeliger, A.; Kryk, H.;
Harm, U.: Multiscaled Experimental Investigations of Corrosion
and Precipitation Processes After Loss-of-Coolant Accidents in
Pressurized Water Reactors. Nuclear Technology, 2018, 205,
248-261
[8] Alt, S.; Kästner, W.; Renger, S.; Seeliger, A.: LOCA Scenariorelated
Zinc Borate Precipitation Studies at Semi-technical
Scale; Proceedings of the Annual Meeting on Nuclear
Technology (AMNT), Berlin, 2019
[9] Kästner, W.; Hampel, U.; Kryk, H.; Harm, U.; Seeliger, A.; Alt, S.;
Renger, S. & Palazzo, S.: Vorstellung des Verbundvorhabens
“ATHLET-Modul zur Simulation thermohydraulischer Folgen
von Zinkborat-Ablagerungen im DWR-Kern”, Proceedings of
Kick-Off Meeting “ATHLET-Modul Zinkborat (AZora)”, 2019
[10] Kästner, W.; Seeliger, A.; Renger, S.; Alt, S.: Lokale Effekte im
DWR-Kern infolge von Zinkborat-Ablagerungen nach KMV,
final report of the BMWi project grant no. 150 1491,
Hochschule Zittau/Görlitz, IPM, 2019
Authors
Prof. Dr.-Ing. Wolfgang Kästner
w.kaestner@hszg.de
Sören Alt
Dr. André Seeliger
Frank Zacharias
Zittau/Goerlitz University
of Applied Sciences
Theodor-Körner-Allee 16
02763 Zittau, Germany
Dr. Ulrich Harm
René Illgen
Prof. Uwe Hampel
Technische Universität Dresden
01062 Dresden, Germany
Dr. Holger Kryk
Helmholtz-Zentrum
Dresden-Rossendorf (HZDR)
Bautzner Landstraße 400
01328 Dresden, Germany
RESEARCH AND INNOVATION 345
Advertisement
www.kerntechnik.com
All-in-one:
Proceedings of AMNT 2019
Get presentations of all sessions on disc.
p Order online: 86.75 € *
www.KernD.de/kernd/shop/
* Price excluding postage. Price including 16 % VAT for Germany and all EU member states without VAT number. For EU
member states with VAT number and all other countries VAT will not be added. For participants of the AMNT 2019 free of charge.
AMNT2019 Proceedings atw2006 179x124v3.indd 1 Research and Innovation
17.06.20 14:20
Modelling Thermal-hydraulic Effects of Zinc Borate Deposits in the PWR Core After LOCA – Experimental Strategies and Test Facilities
ı Wolfgang Kästner, Sören Alt, André Seeliger, Frank Zacharias, Ulrich Harm, René Illgen, Uwe Hampel and Holger Kryk

atw Vol. 65 (2020) | Issue 6/7 ı June/July
RESEARCH AND INNOVATION 346
Investigation on
PWR Neutron Noise Patterns
Marco Viebach, Carsten Lange and Antonio Hurtado
Planned entry for
1 Introduction Investigation of the unexplained changes of neutron flux fluctuation magnitudes
observed in KWU-built PWRs (cf. [1,2]) has drawn attention to long known (cf. [3]) but still
incompletely understood spatial correlation patterns of the neutron flux fluctuations in the
frequency range 0 –2 Hz (cf. [4]). These patterns, namely an out-of-phase behavior of signals from
oppositely located core quadrants and an in-phase behavior of signals from axially aligned locations,
are the dominant fluctuation phenomena because the range 0 –2 Hz carries more than 95 % of the power of the signal
fluctuations and the coherence functions of respective signal pairings have values between 0.5 and 1.0 in this frequency
range (cf. [4]). Therefore, finding the mechanism effecting the measured fluctuation patterns is believed to be key to
explain the changes of the fluctuation amplitudes.
140 144
Recent attempts try to understand the
patterns as being triggered from a
long-range perturbation. Synchronized
lateral fuel-assembly vibrations
are suggested to provide such kind of
perturbation (cf. [4]). A synchronous
vibration of the entire core (as also
proposed in Ref. [3]), leading to a
perturbation possibly called “reflector
effect”, results in signal correlations
similar to those of the measurements.
But the corresponding magnitudes
are found roughly one order of magnitude
lower than observed in the
measurements (cf. [5]).
As a new attempt, synchronized
lateral vibrations that do not involve
the entire reactor core are suggested
as an approach to overcome the shortcoming
of a low fluctuation magnitude
in the model (cf. [5]). Such vibration
mode corresponds to a perturbation
that is located in regions more central
225
· ·· 254 255 256 257
than for the “reflector effect”. Simulations
of corresponding scenarios
give magnitudes of the neutron flux
fluctuations that are within the range
of the measured values (i. e. percents)
and correlation patterns that qualitatively
agree with the measured ones
(cf. [6,7]).
The work at hand investigates a special
case of synchronous lateral fuelassembly
vibration that involves all
fuel-assembly rows, though with
unequal amplitudes. It is assumed that
large-scale coolant flow fluctuations
drive the fuel-assembly vibration such
that the central fuel assembly has the
largest amplitude, both in x- and
y-direc tion. The vibration amplitudes
of the surrounding fuel assemblies are
lower with the lowest amplitude for
the outermost ones. As an extreme
case, this assumption is represented
by a synchronous fluctuation of all
34
.
16
z
f(z)
fuel-assembly gaps. This scenario is
simulated for a KWU Vor-Konvoi PWR
by the neutron-noise tool CORE SIM
[8] in the frequency domain. The model
is based on a corresponding input
(cf. [9]) of the reactor dynamics code
DYN3D [10]. A simulation of similar
type for the above-mentioned “reflector
effect” is presented in Ref. [11].
The simulation shown here aims at
studying the neutron flux fluctuation
patterns that are introduced by the
described scenario. Furthermore, it
investigates whether this scenario
may adequately approximate the
actual picture in KWU-built PWRs.
Therefore, the work at hand tries to
broaden the set of potentially relevant
perturbation sources that can lead to
the observed phenomena. Note that it
is not primarily intended to provide
quantitative results.
The article is structured as follows.
After the introduction, the model is
described in detail before outlining
the concept of CORE SIM and the
preparation of its input. Then, the
simulation results are shown by means
of spatial distributions of absolute
values (amplitudes) and phases of the
neutron flux fluctuations. After a
discussion, the article is closed by
drawing conclusions.
y
x
10 11 12 13 · ·· 33
1 2 3 4 · ··
(a)
Spatial setup. Radial-azimuthal.
z
y
x
| Fig. 1.
Spatial (nodal) setup (a, b) used for the simulation and illustration of the fuel-assembly bow (c). The reflector regions are filled gray
( side with dark and corner with light shading). Channels with detector signals referenced in the results section are shaded red. Numbers
n Ch = 1, 2, . . . , 257 denote channel indices (a) and n z = 1, 2, . . . , 34 axial levels (b), resp. The leading, central fuel assembly is
represented hatched. Considering a given instant, expansion arrows label fuel-assembly gaps that are expanded, and contraction
arrows label those that are contracted (a). The bow shape is illustrated by a dashed line against the straight, nominal shape (c).
.
2
1
(b)
Spatial setup. Axial.
(c)
Bow shape.
2 Simulation of neutron
flux fluctuations
2.1 Models and methods
2.1.1 Modelling of coherent
fuel-assembly gap
fluctuation
The simulation considers a 4-loop
KWU Vor-Konvoi reactor at nominal
power at end of cycle. Figures 1a and
1b illustrate the spatial (nodal) setup
for the neutron-kinetics part and
the thermal-hydraulics part of the
simulation. The steady-state system is
Research and Innovation
Investigation on PWR Neutron Noise Patterns ı Marco Viebach, Carsten Lange and Antonio Hurtado

atw Vol. 65 (2020) | Issue 6/7 ı June/July
p n (z,t) := 1 4 ·
perturbed by the vibration of the fuel
assemblies. The perturbation enters
the calculation via time-dependent
variations of the group constants of
the neutron-kinetics part (cf. Sec.
2.1.3). For simplicity, the vibration
w n (z, t) of the n th fuel assembly (n = 1,
2, . . . , 193) is considered only in
x-direction. It is approximated by
a sinusoidal axial shape function
f(z) = sin(πz/L) (cf. 1c), with L
representing the axial fuel-assembly
length, and a time-dependent elongation
A n (t), i. e. w n (z, t) = f(z)A n (t).
The scenario studied here assumes
that all fuel assemblies vibrate
synchronously, but their elongations
A n (t) have unequal magnitude with
the central one having the largest. The
scenario is motivated by the idea
that in the central core region, the
fluctuations of the coolant flows of
each of the four loops act on the
fuel assemblies there, leading to
correspon dingly large vibration magnitudes.
The outer fuel assemblies are
less affected, responding with smaller
magnitudes. For simplicity, it is
assumed that the magnitude linearly
decreases with increasing distance
from the core center. At the outer fuelassembly
row, the magnitude is zero.
This assumption leads to uniform
fluctuations of all fuel-assembly gaps.
The situation is illustrated in Figure
1c. For each fuel assembly n, the variations
of the center-to-center distances
d njn to its four adjacent fuel assemblies
j n ∈ {north, south, east, west}
are averaged forming an effective fuel-assembly
pitch variation
p n (z,t) := 1 4 ·
∑
∑
j n∈{north,south,east,west}
j n∈{north,south,east,west}
d njn (z,t) .
(1)
2.1.2 Calculation of neutron
flux fluctuations
with CORE SIM
The code CORE SIM solves the neutron
transport equation using diffusion
theory, two energy groups, and one
group of delayed neutrons [8],
input
δΣ (r,ω)
CORE SIM
(Σ 0 (r),φ 1,0 (r), φ 2,0 (r))
| Fig. 2.
Illustration of CORE SIM, calculating the neutron flux fluctuations triggered
by perturbations of the macroscopic cross-sections.
with all symbols carrying their usual
meaning, in the frequency domain.
For this purpose, all variables are
expanded about their steady-state
values, X (r, t) = X 0 (r) + δ X (r, t).
Products of the (time- dependent)
deviations δ X (r, t) are neglected in
order to linearize the equations.
Fourier transformation of the deviations
δ X (r, t) → δ X (r, ω) finally leads
to the frequency-domain equations.
The employed numerical techniques
to solve them are given in Ref. [8].
Practically, CORE SIM calculates
the variations δφ 1 (r, ω) and δφ 2 (r, ω)
of the neutron flux (output) based
on a given distribution δΣ (r, ω) of perturbations
(input) of the macroscopic
cross-sections. The procedure is illustrated
in Figure 2. The calculation is
based on externally provided distributions
of the cross-sections Σ 0 (r) and
on the steady-state distribution (φ 1,0
(r), φ 2,0 (r)) of the neutron flux. The
latter is calculated by CORE SIM in a
steady-state calculation prior to the
calculation of the fluctuations δφ 1
(r, ω) and δφ 2 (r, ω). CORE SIM sets
criticality by renormalizing the fission
cross-sections with the multiplication
factor k eff .
d
2.1.3
njn (z,t)
Preparation
.
of the
CORE SIM simulation
Both the cross-sections Σ 0 (r) and
their perturbations δ Σ (r, ω) are
provided via a DYN3D calculation that
precedes the CORE SIM run. Using this
strategy, the complex configuration of
the reactor’s material data is covered
in the simulation. Furthermore, the
specific impact of the fuel- assembly
gap variations on the cross- sections,
which depends on various parameters
(T fue , T mod , ρ mod , c bor , burnup, fuelassembly
type), gets incorporated.
1 ∂
φ v ∂t 1 (r,t) = ∇(D 1,0 (r) ∇φ 1 (r,t))
1 ∂
φ v 1 ∂t 1 (r,t) = + ∇(D ((1 − 1,0 β)νΣ (r) ∇φ f,1 (r,t) 1 (r,t)) − Σ a,1 (r,t) − Σ r (r,t)) φ 1 (r,t)
1 ∂
φ + +(1 ((1 −β)νΣ f,2 f,1 (r,t) (r,t) φ 2
−(r,t)+λC Σ a,1 −(r,t)+S Σ r (r,t)) 1 (r,t) φ 1 (r,t),
v 1 ∂t 1 (r,t) = ∇(D 1,0 (r) ∇φ 1 (r,t))
∂
φ v 2 ∂t 2 (r,t) = +(1 ∇(D − 2,0
β)νΣ (r) f,2 ∇φ (r,t) 2 (r,t)) φ 2 (r,t)+λC (r,t)+S 1 (r,t) ,
+ ((1 − β)νΣ f,1 (r,t) − Σ a,1 − Σ r (r,t)) φ 1 (r,t)
1 ∂
φ (2)
v 2 ∂t 2 (r,t) = +Σ +(1 ∇(D r −(r,t) 2,0 β)νΣ (r) φ 1 (r,t) ∇φ 2 (r,t))
f,2 (r,t) −φΣ 2 a,2 (r,t)+λC φ 2 (r,t)+S 21 (r,t),
,
1 ∂
C +Σ βν r (r,t) (Σ ∂t φ f,1 (r,t) φ 1 (r,t) φ 1 (r,t)+Σ − Σ a,2 (r,t) f,2 (r,t) φ 2 (r,t)+S φ 2 (r,t)) 2 (r,t) − λC , (r,t)
v 2 ∂t
∂
C 2 (r,t) = ∇(D 2,0 (r) ∇φ 2 (r,t))
(r,t) = βν (Σ ∂t +Σ f,1 (r,t) φ 1 (r,t)+Σ f,2 (r,t) φ 2 (r,t)) − λC (r,t) (3)
r (r,t) φ 1 (r,t) − Σ a,2 (r,t) φ 2 (r,t)+S 2 (r,t) ,
∂
C (r,t) = βν (Σ ∂t f,1 (r,t) φ 1 (r,t)+Σ f,2 (r,t) φ 2 (r,t)) − λC (r,t)
(4)
output
δφ 1 (r,ω), δφ 2 (r,ω)
Figure 3 illustrates the procedure.
Based on a model of a PWR (with
straight fuel assemblies), DYN3D
performs a steady-state calculation,
yielding the steady-state distribution
of the cross-sections Σ 0,DYN3D (r)
with also the thermal-hydraulics
variables converged. The distribution
of effective fuel-assembly pitches
{p n (z m ), n = 1, . . . , 193, m = 2, . . . , 35}
(cf. Eq. (1)), representing the homogeneous
fuel-assembly gap elongation
and the sinusoidal axial shape, is
denoted as Π. A modified version of
DYN3D with a cross-section library
covering variations of the effective
fuel-assembly pitch p n (z m ) (cf. [7])
interpolates the set of cross-sections
Σ Π,DYN3D (r) that corresponds to
the distribution of fuel-assembly
pitches Π on the one hand and to the
complex distribution of the parameters
listed above on the other
hand. The actual perturbation δ Σ
of the cross-sections Σ is their deviation
against the steady-state Σ 0 . 1
The cross-section perturbations that
can be applied in CORE SIM are
calculated as follows:
δ Σ a,1 (n Ch , n z , ω) =
(Σ a,1,Π,DYN3D (n Ch , n z ) −
Σ a,1,0,DYN3D (n Ch , n z )) · δ (ω − ω 0),
(5a)
δ Σ a,2 (n Ch , n z , ω) =
(Σ a,2,Π,DYN3D (n Ch , n z ) −
Σ a,2,0,DYN3D (n Ch , n z )) · δ (ω − ω 0),
δ Σ r (n Ch , n z , ω) =
(Σ r,Π,DYN3D (n Ch , n z ) −
Σ r,DYN3D (n Ch , n z )) · δ (ω − ω 0),
δ Σ f,1 (n Ch , n z , ω) =
(Σ f,1,Π,DYN3D (n Ch , n z ) −
Σ f,1,0,DYN3D (n Ch , n z )) · δ (ω − ω 0),
(5b)
(5c)
(5d)
δ Σ f,2 (n Ch , n z , ω) =
(Σ f,2,Π,DYN3D (n Ch , n z ) −
Σ f,2,0,DYN3D (n Ch , n z )) · δ (ω − ω 0),
(5e)
with the discrete spatial setup (n Ch , n z )
according to Figures 1a and 1b.
Note that the DYN3D levels m = 4 and
m = 5 are homogenized, making
CORE SIM level n Ch = 4. Similarly,
m = 31 and m = 32 make n Ch = 30.
1) Note that coefficients
translating
the elongations
to cross-section
deviations, as used
in the simulations
shown in Ref. [11],
are obsolete for the
current approach.
RESEARCH AND INNOVATION 347
Research and Innovation
Investigation on PWR Neutron Noise Patterns ı Marco Viebach, Carsten Lange and Antonio Hurtado

atw Vol. 65 (2020) | Issue 6/7 ı June/July
RESEARCH AND INNOVATION 348
DYN3D
PWR input
DYN3D
steady-state calculation
steady state
φ 0,DYN3D (r) , Σ 0,DYN3D (r)
fuel pitch distribution
input (Π)
DYN3D
cross-section interpolation
perturbed cross-sections
Σ Π,DYN3D (r)
CORE SIM
steady-state calculation
steady state
φ 0,CORESIM (r) , Σ 0,DYN3D (r)
The symbol δ (ω − ω 0) indicates that
the cross-section perturbation acts at
the frequency ω = ω 0 . Finally, with
the steady- state distribution Σ 0 , DYN3D
and the perturbations δ Σ at hand,
CORE SIM calculates the neutron
flux fluctuations as described in
Sec. 2.1.2.
(cf. Eq. (5))
perturbation calculation
CORE SIM fluctuations
input (δΣ(r,ω 0 ))
CORE SIM
fluctuations calculation
fluctuations
δφ (r,ω 0 )
| Fig. 3.
Procedure of coupled calculations performed in order to simulate the homogeneous fuel-assembly gap variation with CORE SIM.
one CORE SIM run
2.2 Results
For the simulation, the chosen
gap-fluctuation amplitude is 1.6 mm,
which is the nominal gap width [1].
The chosen oscillation frequency is
ω 0 = 2π ∙ 1.0 Hz. Figure 4 presents the
simulated neutron flux fluctuations
for the thermal group. The maximum
magnitude is approx. 4.5 %. It is
located in the outer regions in
x-direction (Figure 4a) at mid axial
level (Figure 4b). The lowest magnitude
is found in the central region in
x-direction and at the bottom and the
top in axial direction. The axial shape
of the magnitudes is C-like. Figure 4c
shows that the fluctuations are out- ofphase
for comparing the left and
the right core half. Along the axial
direction (Figure 4d), the fluctuations
are in-phase. Comparing different
channels with one another, either
in-phase or out-of-phase behavior
is found. The behavior corresponds
to the phase relations seen in the
horizontal view (Figure 4a).
2.3 Measured values
For convenience, Figure 5 briefly
presents measured data of a 4-loop
Vor-Konvoi reactor at nominal power
at end of cycle (details about the
data can be found in Ref. [4]). The
standard deviation takes values in the
range of percents. Along the central
lines (G, J), the magnitude is lower
than in the outer lines (≥N, ≤C). In
the axial view, the magnitude has a
bulgy shape. The phase (determined
by the cross-spectral densities of the
con sidered signals with the signal of
node number in y-direction
17
16
15
14
13
12
11
10
9
8
7
6
5
4
3
2
1
1 2 3 4 5 6 7 8 9 1011121314151617
node number in x-direction
(a) Amplitudes radial-azimuthally at n z = 17 (mid).
4.0
3.5
3.0
2.5
2.0
1.5
1.0
0.5
|δφ2/φ0,2| in %
node number in z-direction
34
30
20
10
5
Ch33
Ch140
Ch144
Ch225
1
0 1 2 3 4 5
|δφ 2 /φ 0,2 | in %
(b) Amplitudes axially.
arg(δφ2) in rad
3π
2
π
π
2
0
− π 2
node no. y-dir.=9
1 2 3 4 5 6 7 8 9 1011121314151617
node number in x-direction
(c) Phase radial-azimuthally at n z = 17 (mid).
node no. in z-dir.
34
15
1
− π 2 0
π
2 π
arg(δφ 2 ) in rad
(c) Phase axially.
| Fig. 4.
Spatial distribution of the induced neutron flux fluctuations δφ 2 for the thermal energy group calculated with CORE SIM for a homogeneous fluctuation of all
fuel- assembly gaps in x-direction at ω 0 = 2π ∙ 1.0 Hz with a sinusoidal axial shape. The upper panel shows the relative amplitudes |δφ 2/φ 0,2| of the fluctuations
and the lower panel shows the phase arg (δφ 2) of the fluctuations. The phase has the input perturbation, for which arg(d S) = 0, as its reference. (In Figure 4d,
the curve of Ch144 overlaps with those of Ch140 and Ch225.)
3π
2
Research and Innovation
Investigation on PWR Neutron Noise Patterns ı Marco Viebach, Carsten Lange and Antonio Hurtado

atw Vol. 65 (2020) | Issue 6/7 ı June/July
P
O
N
M
L
K
J
H
G
F
E
D
C
B
A
1 2 3 4 5 6 7 8 9 101112131415
location B11-6 as the reference)
demonstrates the axial in-phase and
radial out-of-phase behavior known
from this type of reactor (cf. [4]).
2.4 Discussion
The presented simulation overcomes
the defect of the small fluctuation
magnitudes that resulted for the
simulation of the “reflector effect” (cf.
[5,11]) while preserving the characteristic
phase relations of the fluctuations
(see Figures 4c, 4d, and 5b).
The distributions of magnitudes in the
axial and in the radial direction are
similar to the measured ones (see
Figures 4a, 4b, 5). 2 The axial shape
corresponds to the assumed axial
bow shape 3 of the fluctuation magnitudes.
It has to be emphasized that the
scenario considered in this article
is marked by vast simplifications.
Nevertheless, it reproduces relevant
main features of the measured
neutron flux fluctuations. Therefore,
the assumed homogeneous gap
fluctuation is among those scenarios
potentially taking place in the actual
reactor. On the other hand, a proper
mechanism that drives such behavior
has not been found, yet.
Research of the near future
needs to focus on finding plausible
mechanisms that are responsible for
the fuel assembly vibration as a
consequence of coolant. Furthermore,
the trend of the magnitudes in the
horizontal view should be further
investigated. As seen in Figure 4a,
the trend seems to be only little
dependent on the kind of fuel
assemblies; the trend seems to
be a geometrical effect. With regard
to the simplicity of the simulation
shown, the use of the effective fuelassembly
pitch variation has not been
validated, yet. This fact may be tackled
in near future as well.
3 Conclusion
Neutron flux fluctuations of KWU
PWRs show dominant patterns. Based
on the as- sumption that the gaps of
all fuel assemblies fluctuate in a synchronous
manner, the corresponding
neutron flux fluctuations are simulated
with CORE SIM in the frequency
domain. The obtained fluctuation
patterns are similar to the measured
patterns and the obtained fluctuation
magnitudes are in the range of
percents as in the measurements.
Therefore, the assumed scenario is a
potential candidate for being the main
perturbation source triggering the
observed neutron flux fluctuation
patterns. Future research needs to
address the lack of a mechanism
that explains the excitation of fuelassembly
vibrations by coolant-flow
fluctuations.
Acknowledgement
This work was supported by the
German Federal Ministry for Economic
Affairs and Energy (project
NEUS, grant number 1501587). The
responsibility for the content of this
publication lies with the authors.
The authors thank Marcus Seidl for
discussion.
References
3.2
3.0
2.8
2.6
2.4
2.2
2.0
1.8
std(δU/U) in%
axial detector level
1
2
[1] (German) Reactor Safety Commission (RSK), “PWR neutron
flux oscillations,” RSK Statement (457th meeting on
11.04.2013), 2013. http://www.rskonline.de/en/meeting457
3
4
5
6
B11
J02
J06
O05
0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 −π − π 2 0
std(δU/U) in%
[2] M. Seidl, K. Kosowski, U. Schüler, and L. Belblidia, “Review of
the historic neutron noise behavior in german KWU built
PWRs,” Progress in Nuclear Energy, vol. 85, pp. 668 – 675,
2015. http://www. sciencedirect.com/science/article/pii/
S0149197015300652
[3] J. Runkel, “Rauschanalyse in Druckwasserreaktoren,” Ph.D.
dissertation, Universität Hannover, 1987.
[4] M. Viebach, N. Bernt, C. Lange, D. Hennig, and A. Hurtado,
“On the influence of dynamical fuel assembly deflections on
the neutron noise level,” Progress in Nuclear Energy, vol. 104,
pp. 32 – 46, 2018. http://www.sciencedirect.com/science/
article/pii/S0149197017302147
[5] M. Viebach, C. Lange, N. Bernt, M. Seidl, D. Hennig, and
A. Hurtado, “Simulation of low-frequency pwr neutron flux
fluctuations,” Progress in Nuclear Energy, vol. 117, p. 103039,
2019. http://www. sciencedirect.com/science/article/pii/
S0149197019301349
[6] L. Torres, D. Chionis, C. Montalvo, A. Dokhane, and
A. García-Berrocal, “Neutron noise analysis of simulated
mechanical and thermal-hydraulic perturbations in a pwr
core,” Annals of Nuclear Energy, vol. 126, pp. 242 – 252,
2019. http://www.sciencedirect.com/science/article/pii/
S0306454918306303
[7] M. Viebach, C. Lange, M. Seidl, Y. Bilodid, and A. Hurtado,
“Neutron noise patterns from coupled fuel-assembly
vibrations,” in PHYSOR 2020: Transition to a Scalable Nuclear
Future, Cambridge, United Kingdom, March 29 -April 2, 2020,
2020.
[8] C. Demazière, “CORE SIM: A multi-purpose neutronic tool for
research and education,” Annals of Nuclear Energy, vol. 38,
no. 12, pp. 2698 – 2718, 2011. http://www.sciencedirect.
com/science/article/ pii/S0306454911002210
[9] U. Rohde, M. Seidl, S. Kliem, and Y. Bilodid, “Neutron noise
observations in German KWU built PWRs and analyses with
the reactor dynamics code DYN3D,” Annals of Nuclear Energy,
vol. 112, pp. 715 – 734, 2018. http://www.sciencedirect.
com/science/article/pii/S0306454917303687
[10] U. Rohde, S. Kliem, U. Grundmann, S. Baier, Y. Bilodid,
S. Duerigen, E. Fridman, A. Gommlich, A. Grahn, L. Holt,
Y. Kozmenkov, and S. Mittag, “The reactor dynamics code
DYN3D – models, validation and applications,” Progress in
Nuclear Energy, vol. 89, pp. 170 – 190, 2016. http://www.
sciencedirect.com/science/article/pii/S014919701630035X
[11] M. Viebach, N. Bernt, C. Lange, D. Hennig, and A. Hurtado,
“Frequency-Domain Investigation on the Neutron Noise in
KWU PWRs,” in 49th Annual Meeting on Nuclear Technology,
Berlin, Germany, May 14-65, 2018, 2018.
Authors
Marco Viebach
marco.viebach@tu-dresden.de
Dr.-Ing. Carsten Lange
Prof. Dr.-Ing. Antonio Hurtado
Chair of Hydrogen and
Nuclear Energy
Technische Universität Dresden
George-Bähr-Str. 3b,
01069 Dresden, Germany
π
2
phase(δU) in rad
(a) Radial view.
(b) Axial view (phase ref.: B11-6).
| Fig. 5.
Measured data of neutron flux fluctuations. Standard deviation std() of detector signals normalized w. r. t. their mean values and
phase of the fluctuations w. r. t. those at detector B11-6.
2) The radial-azimuthal
pictures are rotated
by 90°, which would
not be the case for
considering the fuelassembly
bow
exclusively in
y- rather than in
x-direction. Note
that the x- and
y-direction are
equivalent in the
underlying model.
In the real reactor,
exclusive consideration
of only one
direction is impossible.
Therefore,
the lack of the
90°-rotational
symmetry in the
measured data indicates
an inherent
asymmetry.
3) See Ref. [11] for a
comparison of the
results for various
bow shapes.
RESEARCH AND INNOVATION 349
Research and Innovation
Investigation on PWR Neutron Noise Patterns ı Marco Viebach, Carsten Lange and Antonio Hurtado

atw Vol. 65 (2020) | Issue 6/7 ı June/July
350
OPERATION AND NEW BUILD
Planned entry for
| Fig. 1.
Control principle of a modern PWR.
Reactor Core Control
Based on Artificial Intelligence
Victor Morokhovskyi
Governing of nuclear reactors worldwide
is currently based on classical
control technology. However, control
technology applied for this task
reaches its applicability limits.
This article proposes a new
approach for governing of Pressurized
Water Nuclear Reactor (PWR) based
on Artificial Narrow Intelligence
(ANI).
2 Current state of the art
Figure 1 shows the control principle
of the modern Nuclear Power Plant
(NPP) with Pressurised Water Reactor
(PWR) and also holds for DWR
( German, DruckWasserReactor) and
VVER (Soviet, Water-Water Energetic
Reactor). In such NPPs the electrical
output is determined by the position
of the turbine valves and the governance
of the electrical power is the
task of turbine controllers. Turbine
controllers possess a set-point for
the electrical power; this set-point can
be adjusted from the control desk.
Adapting of electrical output to grid
demand is usually performed using a
phone connection between the plant
operator and the grid dispatcher.
1 Introduction A nuclear reactor is a complex system, comprehensive control of it is not trivial.
Reactor controllers belong to the most complicated devices created by humans. Besides well-known
control of thermal power and coolant temperature, reactor controllers take care of plenty of other
aspects such as operational safety permitting operation only within given limits, uniforming of
burnup, burnup compensation, compensation of the poisoning, uniforming of the power density
distribution, support of flexible electricity production, operation economy, etc.
Such a governance concept obviously
implies the necessity to keep the
power balance between the nuclear
island and the turbine island, which is,
in this case, the task of the nuclear
island control system. There exist two
process variables, which can be used as
an indicator for the power balance:
Average Reactor Coolant Temperature
(ACT) and the Live Steam Pressure
(LSP). If these process variables remain
constant, power balance is ensured.
PWRs including German design
DWRs achieve the power balance by
keeping the ACT constant, while
VVERs do it by keeping the LSP
constant. In both cases the reactor
power will be influenced by the control
rods of the P-bank (Power bank).
Movements of the P-bank affect,
however, the axial power distribution
in the core and the necessity of
additional measures for keeping of the
Axial Offset (AO) in the appropriate
range occurs. The AO is usually
measured by the core internal neutron
flux measuring system ( incore). Usage
of the external system (excore) for
this purpose is also possible. Depending
on the type of PWR, AO control is
performed using either boration/
dilution system (Bo/Di) or the second
movable control rod bank, called the
H-bank (Heavy bank). Depending on
the plant type, the AO is adjusted
either manually or automatically,
using the so-called AO-controller.
Figure 1 shows also the neutron
flux controller (Φ-controller), needed
for the start-up of the reactor as
well as the bypass controller needed
for the start-up of the secondary
circuit and for the overriding of large
disturbances in energy production.
3 Recent challenge
The biggest modern challenge for
NPP control is flexible operation. This
operation mode has recently gained
importance throughout the world [1].
The reasons for flexibility of old and
new NPPs are fluctuating power of
renewables, the need to follow the
daily and weekly load profiles of
consumer as well as the considerable
share of nuclear power in the energy
mix of some grid segments.
In the case of flexible operation
the power output should continuously
be changed according to the grid
demand. In Germany, for example, it
means, that in addition to primary
frequency control, every 15 minutes a
NPP receives a new set-point for
the electrical output from the grid
dispatcher. This continuous set-point
adjustment causes new additional
effort for the plant operator. Furthermore,
the continuous change of the
turbogenerator power affects the power
balance between the nuclear island
and the turbine island and the control
system of the nuclear island will be
demanded. In this way, the reactor
control system gets a new sophisticated
challenge compared to the former
times, when the power was almost
constant throughout the operation
cycle. Flexible operation yields several
new difficulties. First of all, the Xenon
poisoning will no longer remain
Operation and New Build
Reactor Core Control Based on Artificial Intelligence ı Victor Morokhovskyi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
constant. Xenon poisoning is a very
strong effect having a complex longtime
dynamics well known as Xenon
transients. Spatial distribution of the
Xenon in the reactor core is the next
challenge. Thus, both, ACT (or LSP)
control and AO control are highly
demanded in this case. Even if these
control systems can cope with the
continuous power changes, several
questions remain unanswered. Continuously
changing process variables
may touch the operation limits secured
by the limitation system causing painful
effects, such as the impossibility to
increase the reactor power according
to the current grid demand leading
sometimes to penalties. Continuous
and intensive control actions result in
the higher expenditure of control resources
like boric acid, demineralized
water and movement steps of the control
rods. Increased consumption of
the boric acid and demineralized
water increases the load on the coolant
reprocessing system and can bring
it to its limits. Increased consumption
of the control resources raises the
question of operation economy.
Further effects include the non- uniform
burnup of the nuclear fuel and
unwanted burnup of the control rods.
All of this shows that, with flexible
operation, reactor control receives a
set of new goals. The simple control
of two process variables no longer
suffices. Moreover, due to the complex
long- lasting dynamics of Xenon
poisoning released by each and every
power change and affecting the
reactor core around 20-30 hours after
this change, momentary consideration,
which is common for control
technology, is no longer appropriate.
Prediction of process variables for at
least 24 hours is needed. The next
new challenge for the reactor control
system is keeping the reactor core
and the coolant reprocessing system
apparat from all their limits under
new intricate dynamic conditions.
The reactor control system should
now monitor and/or guarantee the
ability to quickly ramp up to 100 %
power from every operation state to
avoid possible penalties on the side of
the grid operator. Due to the spatial
distribution of Xenon poisoning along
with its complex dynamics, instant
control of AO no longer suffices. The
control system should guarantee
reasonable values of AO in the future,
for example within the prediction
horizon of 24 hours. The control
system should now take care of
operation economy, minimizing rod
movements as well as boron and
demineralized water consumption. At
the same time, the system should
remain simple and its parametrization
trivial. The control system should
allow easy fitting to all possible
changes, e.g. different loadings.
Control technology applied for this
task obviously reaches its applicability
limits. The main difficulties are:
p trying to solve an ambitious inverse
problem for a very complex system
p the number of associated goals
which is significantly larger than
the number of actuators available
for core control like control rod
banks and Bo/Di valves.
Governance of a complex system
like a nuclear reactor with a series of
goals and a number of constraints,
especially in the case of flexible power
operation, is not an issue for classical
control technology, but is a classical
task for Artificial Narrow Intelligence
(ANI).
4 Application of artificial
intelligence for the
governance of a nuclear
reactor
Artificial Narrow Intelligence (ANI) is
a kind of AI designed to perform a
single specific task. This kind of AI
already exists today. The most known
examples of ANI are chess computer
and street navigator, the most recent
example is AlphaGo.
Contrary to control technology,
Artificial Narrow Intelligence allows
consideration of an arbitrary large
number of goals even if quite different
in nature. In the case of reactor
governing, safety, ergonomics, operation
economy and grid services can be
processed simultaneously in accord
with each other. The scope of goals
can be easily extended every time.
Unlike reactor controllers based on
classical control technology, Core
Control Based on Artificial Intelligence
(COCOAI) can generate not
only control commands in real time, it
can also compile comprehensive plans
for control actions for the next
24 hours, continuously update these
plans and permanently display them
to the operator along with predicted
trajectories for all important process
variables for this time horizon.
5 Integration of new
devices into existing
NPPs and application
for new build plants
Generally there are two ways to upgrade
current PWR control systems.
The first is by modernizing of
existing I&C [2], the second by
introducing new supplemental I&C
devices. In case of computerized
control systems, the first way is
relatively inexpensive, since only the
software of the corresponding devices
will be updated. This approach has
been used in several German NPPs and
in the Swiss NPP Gösgen in the last
decade. Within the scope of ALFC (Advanced
Load Follow Control) projects,
the turbine and reactor controller
software of all these NPPs was updated
to make their control systems
more suited to flexible operation.
The modernization approach has
its natural limitations. Modifying of
the existing approved I&C system
implies risks for safety and reliability.
Since only application software will
be updated, the hardware and the
operational system of the plant control
remains as is, limiting the applic ability
and the performance of new possible
algorithms.
Another option to upgrade the
control system means introducing the
new supplemental I&C devices without
intervening in the existing I&C
(Figure 2).
5.1 Load Governor
The first supplemental device is called
Load Governor (see Figure 2; German:
Einsatzrechner). For the first
time Framatome installed the Load
Governor in a NPP in Germany in
2002. The device enables a fully automatic
management of the electrical
output based on a Load Schedule. A
Load Schedule contains a load program
for the current and following
day having typically a time step of
15 minutes and is provided by the grid
dispatcher.
Today’s grid operators have such
Load Schedules for the whole grid as
well as for each single plant and can
share these Load Schedules with plant
operators. The Load Schedule for the
whole grid results from the predicted
grid consumption profile based on
consumption plans and experience as
well as from the predicted renewable
power profile based on the weather
forecast.
On receiving of the Load Schedule
the plant operator observes it on the
screen of the Load Governor, edits it if
necessary, endorses it and saves it. For
redispatch purposes, each time the
Load Schedule can be edited or overwritten
with a new one.
The Load Governor will now
execute all changes of electrical output
according to the Load Schedule, automatically
changing the plant output
each 15 minutes according to the
OPERATION AND NEW BUILD 351
Operation and New Build
Reactor Core Control Based on Artificial Intelligence ı Victor Morokhovskyi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
OPERATION AND NEW BUILD 352
| Fig. 2.
COCOAI governing architecture.
saved power set-points. The interface
of the Load Governor to the existing
turbine controller is kept very simple;
the Load Governor uses relay contacts
in parallel to control desk pushbutton
contacts. This kind of interface is
compatible with all possible turbine
controllers. Effort required from the
turbine operator reduces significantly.
Additionally the Load Governor
can comprise the functions of primary
and secondary frequency control if
the existing turbine controller does
not include them, or if these functions
within the existing turbine controller
do not satisfy recent requirements.
5.2 Core Governor
The second new supplemental device
is called Core Governor (Figure 2).
The Core Governor comprises all the
functions needed for power operation
of the reactor core except the ACT (or
LSP) control.
Such a governing architecture
allows the control system to be subdivided
into two levels (Figure 2):
Controller Level securing the operation
and Governor Level optimizing
the operation. All safety and reliability
related functions remain at Controller
Level. The Governor Level includes
devices which can be assigned to Operator
Assistant Systems (OAS) having
lower safety requirements than
devises at Controller Level.
Moreover, the described governing
architecture makes it possible to use
Artificial Narrow Intelligence technology
for the Core Governor function.
Usage of this new technology enables
to take into consideration all existing
and imaginable goals of core control
simultaneously and not only for the
current moment but also for a significant
time span in the future. COCOAI
can govern the core in the way,
avoi ding possible problems like
touching of process limits not only for
the current moment, but using
pre diction also for the next 24 hours.
The main inputs for this prediction are
the Load Schedule coming from
the Load Governor (see Figure 2)
and poisoning vector calculated by
the Core Governor on the basis of
the power history. Simultaneously,
COCOAI ensures the most economical
operation within given limits in all
phases of the burnup cycle. The
actuators of the Core Governor are the
Bo/Di system and H-bank drives (see
Figure 2). There are three implementation
possibilities for the actuator
interface: manual, semi- automatic
and automatic.
In the first case the Core Governor
serves as an OAS showing the predicted
values of all important process
variables for the next 24 hours,
together with the automatically compiled
plan for control actions for this
time span and proposing these control
actions in real time. If the operator
executes the proposed control action,
process variables remain in the
current plan; if the operator ignores
the proposition or makes some unproposed
actions, COCOAI will, like a
street navigator, quickly compile a
new plan and display it.
In semi-automatic mode the operator
pushes and holds the ‘enable’ button
and the Core Governor automatically
performs the control actions while the
’enable’ button is pushed.
In fully automatic mode, pushing
of the ‘enable’ button is not more
necessary. The operator observes the
process and its prediction on a screen
having the possibility each time to
deselect the fully automatic mode and
to perform further control of Bo/Di
and H-bank manually. The ACT (or
LSP) control, actuating the P-bank,
remains fully automatic, since it is
provided by a simple control device at
Controller Level (see Figure 2).
6 Summary
The new governing concept presented
in this paper can be applied to all
existing PWRs including all variations
such as DWR or VVER as well as for
new build NPPs if they use PWR
principle. It enables highly econo mical
and flexible operation of the NPP
within given operation limits. Additionally
the NPP operator receives a
prediction for all important process
variables for the next 24 hours.
References
[1] Victor Morokhovskyi, Jürgen Rudolph, Tatiana Salnikova,
Transfer of Concepts for Flexible Operation to Different Plant
Types, Printed in AMNT, Berlin, Mai 16-17, 2017.
[2] Klaus-Peter Hornung, Andreas Kuhn, Victor Morokhovskyi,
Advanced Load Following Control with Predictive Reactivity
Management (ALFC-PREDICTOR), London, 26-th ICONE –
International Conference on Nuclear Engineering, July 22-26,
2018.
Author
Dr.rer.nat. Victor Morokhovskyi
Senior Expert at Framatome GmbH
in Erlangen, Germany
Lecturer at Hannover University
of Applied Sciences and Arts
victor.morokhovskyi@
framatome.com
Operation and New Build
Reactor Core Control Based on Artificial Intelligence ı Victor Morokhovskyi

atw Vol. 65 (2020) | Issue 6/7 ı June/July
On the Potential to Increase the Accuracy
of Source Term Calculations for Spent
Nuclear Fuel from an Industry Perspective
Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman
1 Introduction One of the many success factors of nuclear projects and in particular of interim Planned entry for
storage and final repository projects are: the economic viability of the endeavor and the reliability
of the engineering predictions. The better the accuracy of simulation tools and codes is, the smaller
are the required error margins of parameters relevant for nuclear and non-nuclear safety
assessments and the smaller are the required resources to build the above-mentioned facilities. For
example, the decay heat emitted from storage casks at the time of entry is one factor that determines the minimum
spacing between casks in a deep underground repository. The decay heat also determines the minimum required
shutdown cooling time before fuel assemblies can be transported to an interim storage facility and final repositories.
The gamma and neutron source terms determine the shielding requirements for transport casks and packaging facilities.
The planned deep underground repository in Forsmark, Sweden, for example, is designed to have a capacity of
6,000 canisters and requires an excavation mass of about 1.6M tonnes of rock [1]. If the required volume can be reduced
by 10 %, due to more accurate predictions of the minimum canister distance, important costs savings for the ~500M€
[2] worth of tunnel investments would follow. Another important cost driver is the waiting period until all spent fuel
can be removed from a shutdown nuclear power station. Operation of required safety systems for criticality safety and
heat removal cost several 10k€ per day. Therefore, reducing the wait time by several months can make a substantial
contribution to the financial performance of a plant decommissioning project.
Besides project costs an equally important
success factor is the reliability
of engineering predictions regarding
the safety parameters of the spent
nuclear fuel. A high precision estimate
of a safety parameter based on today's
knowledge can turn out to be biased
and predicted with too optimistic
error margins if new research leads
to a revision of taken-for-granted
methods and data. The consideration
of this possibility is especially relevant
for the above-mentioned projects,
with planning phases that can take
many years and execution phases
often spanning many decades. The
need for cost-optimization on the one
hand and the potential of incomplete
knowledge on the other hand, can
result in an overoptimization of a
facility’s engineering design which is
not sufficiently robust to absorb future
revisions of established methods.
This article is structured as follows:
firstly, a short review of the state-ofthe-art
of source term determination
which encompasses nuclide vector
determination of spent fuel, gammaand
neutron source terms and decay
heat is given. Secondly, identification
of potential knowledge gaps and
options to improve the accuracy of
current methods and tools follows.
The role of the EURAD task 8, subtask
2 [3] to contribute to this objective is
explained. Thirdly, given the current
set of data to validate simulation tools
and codes the case for using either
thin-tailed or thick-tailed statistics to
generate robust engineering pre dictions
is discussed.
2 Prediction of source
terms for spent nuclear
fuel
A determination of source terms for
spent nuclear fuel can be divided into
four knowledge domains. First: initial
material composition and geometry.
Second: parameter change during
irradiation. Third: nuclear data
including neutron interaction cross
sections, fission product yields,
neutron and gamma-ray emission
data and radioactive decay data.
Forth: nuclide vector generation
| Fig. 1.
Knowledge domains for making source term predictions.
during irradiation and decay chains
simulation. The domains are shown
in Figure 1.
From a life cycle point of view
reactor operation comes first and
criticality safety considerations and
the determination of the effective
multiplication factor k eff were traditionally
of higher priority compared to
parameters important for backend
activities. Therefore, reactor physics
tools which determine the neutron
field during reactor operation are
mostly validated with high quality
data often obtained from single effects
tests. What constitutes a single effects
test depends on circumstances.
353
DECOMMISSIONING AND WASTE MANAGEMENT
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
DECOMMISSIONING AND WASTE MANAGEMENT 354
Compared to conditions in a commercial
reactor here single effects tests are
meant to have material compositions,
geometries and boundary conditions
which are much better defined and
are relatively simple configurations
compared to the order of 50k of fuel
rods in a commercial reactor. There is
very little uncertainty regarding irradiation
conditions and main emphasis
is on validating microscopic data.
In later stages of the reactor life
cycle nuclides relevant for burnup
credit receive more attention. First,
they are important to predict the
reactivity and other safety parameters
of a reactor during cycle burnup and
core reload. For example: critical
boron concentration as a function of
full power days, power density
peaking and homogenization during
irradiation. Second, these nuclides
are inputs for safety analyses in which
the radioactive inventory is a major
parameter (e.g. decay heat during
regular shutdowns or dose rate
| Fig. 2.
Factors influencing accuracy of source term validation for relatively simple (single effects) tests (top) and integral tests like samples
from commercial nuclear fuel (bottom).
| Tab. 1.
Nuclides of interest identified in [49,50] relevant for criticality, burnup credit and dose rate.
calculations during accidents). Moreover,
they are used as input for safety
analyses regarding transport and
storage of spent fuel. Finally, as
the life cycle ends and interim and
final repository activities increase, the
priorities among nuclides and radioactive
decay modes again changes due
to the much larger time scales for
these projects.
For example, the SCALE code
system, which covers many of the
reactor physics and backend analysis
fields [4], has been exten sively
validated with experiments collected
in the International Handbook of
Evaluated Criticality Safety Benchmark
Experiments (ICSBEP Handbook
[5]). In these experiments the
system configurations are kept as
simple as possible: uranium or
plutonium systems with a wide range,
but accurately defined isotopic vector
variations. Other, simple materials
include light water as primary
moderator, and reflectors consisting
of light water as well as graphite,
beryllium, molybdenum. The geometrical
configurations are often
much simpler than in a commercial
reactor, they are static and typically
no nuclides relevant for burnup credit
are included.
For the purpose of criticality safety
for transport, storage and treatment of
spent fuel the feasibility and reliability
of burnup credit has also seen considerable
effort [6, 28]. While code- tocode
benchmarks are straight forward
[7] a comparison with measured
nuclide vectors requires much more
effort and resources [8, 9, 10]. Firstly,
in many cases irradiated fuel comes
from commercial reactors and
boundary conditions during irradiation
are less well known compared to
single effects tests for criticality benchmarks,
for example. Secondly, a post-irradiation
determination of the nuclide
com po sition is resource intensive and
usually only done for pellet-sized
samples of a fuel assembly. While the
average energy generation of a fuel
assembly is known with relatively high
accuracy, factors such as local parameter
variation due to rod or fuel bowing,
moderator conditions, neutron
field suppression by spacer grids, neutron
spectrum shifts induced by neighboring
fuel assemblies or shielding by
moving control rods increase the uncertainty
of the nuclide vector prediction
at the pellet- scale and therefore
limit validation efforts. Thirdly, nuclide
vector determination at a fixed burnup
point yields only a single snapshot of
the behavior of a non-linear system and
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
therefore limits the ability to extrapolate
the validation to different
burnup conditions. Figure 2 summarizes
relevant factors influencing the
evaluation of samples from commercial
reactors. Under ideal validation
conditions irradiation would be done
with well-known circumstances in a
research reactor and nuclide vectors
would be determined for a series of
burnup steps to eliminate most of the
above-mentioned limitations.
| Fig. 3.
Concentration of U235, Cm244, Nd148, Pm147 for a reference PWR UO2 assembly at 50MWd/kgU;
while the EOL burnup remained fixed; the power history and the cycle durations were randomly
changed for the assembly’s 4-cycle lifetime.
Table 1 marks the most pro mi nent
nuclides for criticality, for burnupcredit
and for radiation dose of spent
fuel. Which nuclides are more relevant
than others depends on time
scales and safety parameters. Nuclides
contributing to neutron emission are
different from nuclides contributing
to decay heat. Nuclides contributing
to decay heat at reactor shutdown are
different from nuclides contributing
to decay heat in a final repository.
Also, final repositories often have
limits on the concentration of particular
nuclides mentioned in other
environmental regulations which fall
outside of the attention of classical
source term determination.
Figure 3 shows the relative concentration
of some actinides and
fission products for a typical 4 wt%
U-235 PWR fuel assembly (determined
with the SCALE code system).
The irradiation history (power and
duration) was randomly changed but
EOL burnup was kept constant and all
values are normalized to the results of
the reference irradiation. For some
nuclides such as Cm-244 or Pm-147
history effects matter because of the
DECOMMISSIONING AND WASTE MANAGEMENT 355
| Fig. 4.
Nuclide vector spread for a representative PWR UO2 fuel assembly at 50MWd/kgU; nuclide concentrations are normalized to burnup of each node
(i.e. if the nuclide concentration would scale linearly with burnup all values would be at 1.0).
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
DECOMMISSIONING AND WASTE MANAGEMENT 356
non-linear character of the nuclide
generation and destruction chains.
Another example is shown in Figure
4. Results for the 32 axial nodes of
the same fuel assembly as above were
analyzed. In the figure the nuclide
concentrations were first normalized
with the mean value and then
scaled with burnup. As expected,
the Nd-148 monitor values are concentrated
at 1.0. But for many other
nuclides the scatter is visibly larger.
This underlines again the difficulty
to get high quality test data from
commercial irradiation.
3 Potential for
improvement of source
term predictions
The validation of source terms has
two legs: first, the simulation tools
and codes which determine them
use as input evaluated nuclear
data such as ENDF/B [11] or JEFF
| Fig. 5.
Using the principles of particle transport and decay to transform an initial nuclide vector with evaluated, measured microscopic data into
a nuclide vector at a future state.
LIB Cumulative yield (%)
Sr-90
Cs-137
JEF-2.2 5.847 6.244
JEFF-3.1.1 5.729 6.221
JEFF-3.3 5.676 6.090
JENDL-4.0 5.772 6.175
ENDF/B-V 5.913 6.220
ENDF/B-VII.1 5.782 6.188
1-sigma 1.20 % 0.40 %
LIB
Sr-90 + Y-90
+ / keV
Cs-137 + Ba-137m
decay data 1129 813
JEFF-3.1.1 1107 812
JENDL/FPD-2011 1130 811
ENDF/B-VII.1 1129 806
1-sigma 1.00 % 0.30 %
LIB
Integral, average cross section
Sr-90 (b)
Cs-137 (mb)
TENDL-2017 3.936 1.071
JENDL-4.0u 4.018 0.926
JEFF-3.3 3.937 1.040
ENDF/B-VIII.0 3.987 1.573
1-sigma 1.00 % 25 %
| Tab. 2.
Simple estimate of uncertainty regarding yield, neutron capture of Cs-137 and Sr-90 and decay energy
from data of different microscopic data libraries.
[51]: microscopic cross sections,
fission product yields and radioactive
decay data. The majority of these
data are provided with covariance
information [12]. By propagating this
input through reactor irradiation
simulations and through decay
periods the source terms and their
uncertainty can be determined [13,
14, 15]. From this perspective the
“theoretical” calculation of source
terms is a transformation of an initial
nuclide vector to a new nuclide
vector by means of the laws of particle
transport and radioactive decay
using evaluated nuclear data, see
Figure 5.
Second, the codes for source term
determination can be validated with
measured nuclide concentrations
such as given in the SFCOMPO database
[16], with integral measurements
of neutron and gamma source
strengths of spent fuel [17, 18, 19] and
decay heat [20, 21] from irradiated
fuel samples and fuel assemblies.
If this information would be the
only source of validation, a code
could be entirely based on empirical
para metrizations and could be sufficiently
accurate if its application
stays within the established parameter
range. For example, the classical
formulas for decay heat in [22] or
[23] are of this kind.
Some of papers published in the
literature suggest that SCALE and
other sophisticated codes used to
predict SNF source terms appear to
perform better in terms of accuracy
than can be justified by the uncer tainty
of the fundamental, microscopic input
data (see following example of decay
heat predictions). In other published
results the measurement- theory comparisons
show much higher deviations
than would be expected from the
uncertainty of the microscopic data
(see following example on nuclide
vector prediction).
In [24] decay heat measurements
on spent nuclear fuel were performed.
50 BWR and 34 PWR assemblies
were selected for measurement
from the Clab inven tory. Shutdown
cooling period was 11 to 27 years in
these cases. The measurement- theory
agree ment in this non-blinded study
was reported excellent and not
larger than the decay heat measurement
uncertainty of 2 %. In a followup
study [25] the overall decay heat
uncertainty from both modeling and
nuclear data was estimated at 1.3 %.
Research in [26] also concluded that
measurement- theory comparisons for
decay heat were mainly limited by the
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
accuracy of the calorimeters used in
these experiments. For the assemblies
considered in this exercise Cs-137 and
Sr-90 are among the main decay heat
contributors from the entire nuclide
inventory. A simple estimate (by comparing
values in different evaluated
data libraries) of the uncertainty of
their number densities due to fission
yield and absorption cross section
uncertainty combined with the
uncertainty of the specific heat makes
the above 1.3 % estimate appear very
optimistic (see Table 2). Furthermore,
research in [29] with coupled
Monte Carlo and burnup calculations
and comparisons with data from post
irradiation examinations concluded
that the inventory of plutonium isotopes
can be predicted within 2-4 % of
measured values. Given the very good
agreement of decay heat measurements
with predictions in the above
example there is the possibility that a
procedure can be formulated about
how the irradiation history simulation
with its many degrees of freedom
must be done to minimize bias. If
codes are validated and are used in a
parameter range defined by available
experiments this can be an acceptable
approach from a safety point of
view. More attention is necessary if
calculations are made for long range
forecasts, which cannot be verified
before a project receives licensing
approval.
Also, decay heat codes have been
validated at short cooling times
against pulse fission experiments (for
example [30, 31]) with estimated
uncertainties for UOX and MOX fuels
of about 7.5 %. The WPEC Subgroup
25 was formed in 2005 to assess and
recommend improvements to the
fission product decay data for decay
heat calculations [32]. It already
considered the question if a reduction
in the uncertainty in decay heat
calculations to about 5 % or better is
achievable. One conclusion was that
more accurate measurements were
required to determine the decay
constants of key radionuclides. However,
in the recommended list for
obtaining better data on 37 nuclides
the emphasis was mostly put on
nuclides with short decay times.
Already in 1976, the impact of the
uncertainties in fission-product yields,
half-lives and decay energies on decay
heat was studied in [33, 34]. This
assessment indicated that decay heat
can be calculated to an accuracy of 7 %
or better for cooling times > 10 sec.
The expected accuracy fell to 3 % for
cooling times larger than 10 3 sec.
| Fig. 6.
C/E values for Cm-244 and Cs-137 from [35].
In [35] predictions by the SCALE
code system for PWR spent fuel
nuclide inventory were compared
with results from measurements.
In this research a total of 118 fuel
samples were analyzed and predictions
for 61 nuclides were included. In
Figure 6 the C/E ratios (experiment
measured over calcu lated) are shown
for Cm-244 and Cs-137 as a function
of sample burn up. The C/E values
follow no clear trend with burnup.
This is the case for most other
nuclides. Variations between samples
of similar burnup can be as large as
variations between samples of large
and small burnup and magnitudes can
be as large as 10 % and higher.
Even if observables like neutron
emission or decay heat can be predicted
well through fortunate circumstances
of error elimination in some
parameter domain, three challenges
remain: first, the error cancellation
might not occur for those states and
time scales which cannot be experimentally
verified. Second, for some
projects the nuclide number densities
themselves are important and the
reasons for the observed, relatively
large C/E variations must be understood.
Third, in order to formulate an
improvement strategy of existing
codes samples whose irradiation
conditions are known with higher
accuracy are necessary.
Concerning the second point, the
C/E variations in Figure 6 appear
rather random without a trend or bias
with burnup. For most nuclides and
experiments the stated nuclide measurement
uncertainties are very small
compared to the observed range of
C/E variations. Moreover, nuclear
input data such as fission product
yield and microscopic cross section
uncertainties do not fully explain the
observed variations. For example,
results in Figure 7 show the impact
of these uncertainties for the fuel
DECOMMISSIONING AND WASTE MANAGEMENT 357
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
DECOMMISSIONING AND WASTE MANAGEMENT 358
| Fig. 7.
Estimating Cm-244 and Cs-137 concentration uncertainties (relative units) due to cross section uncertainties, fission yields and decay
parameters for a representative UO2 PWR fuel assembly at 50 MWd/kgU.
example from section 2. Calculations
were done with the SAMPLER module
from SCALE which uses the therein
provided covariance information
[36]. Also, this source of uncertainty
should manifest itself as a slowly
varying bias as a function of burnup,
not randomly changing between
samples with similar burnup.
Research in [38] made detailed
calculations on how the uncertainty of
the boron concentration, of the fuel
and moderator temperature, of the
final burnup, of the initial U-235
enrichment, of the fuel assembly pitch
and of the type of fuel assembly
neighbors affect C/E results. Assuming
expert guesses for plausible input
parameter ranges, the results show
that expected uncertainties for C/E
due to these factors for most of the
relevant nuclides are smaller than
5 % (Table 3) and are unlikely to explain
C/E variations in the order
of 10 % or more.
As already mentioned, one possible
explanation is that irradiation conditions
on the scale of pellet-sized
samples have much higher uncertainties
than typically assumed. But
they should also average out over the
irradiation lifetime. Another explanation
is that the experimental uncertainties
of the radiochemical nuclide
inventory data may be biased due to
systematic effects depending on the
laboratory or method that is used.
A third explanation is that burnup
monitors like Nd-148 are not sufficiently
reliable to establish similarity
between samples and that more
variables are necessary to create
meaningful classes of samples.
Finally, unrecognized sources of
uncertainty [37] have been introduced
among researchers responsible for
providing evaluated nuclear data to
address the issue that uncertainties
based on existing covariance information
sometimes appear to be
inconsistent and underestimated
with observed scatter of predicted
mean values for cross sections or
benchmarks. In the context at hand
irradiation conditions at pellet-scale
or lack of an adequate set of irradiation
history variables could be
examples thereof.
4 Options for
improvement of source
term predictions
One of the simplest methods used in
industry practice to reliably predict
source terms (i.e. conservatively
overpredict concentration or source
strength) uses the minimum from a
set of C/E results and applies this
value as penalty factor in future calculations.
For example, the C/E values in
Figure 6 suggest that the calculated
Cm-244 concentration is under estimated
at most by a factor of 0.6. All
future calculation results would be
multiplied with a penalty factor of 1.7.
The disadvantage of this approach
is that it depends only on a single
minimum value which could also be
an outlier. Another downside is that
in this approach no information is
generated for situations which are not
covered by the existing validation
database. Also, any burnup dependence
of the penalty factor is ignored.
Moreover, the information of all the
other samples’ C/E result is discarded.
An appropriate statistical analysis
framework is necessary to account
for all the information which is available
in the data. The main condition
to decide is whether the observed,
seemingly random variations of
the C/E results are thin- (optimistic
approach: statistical independent
sample irradiation and evaluation
conditions, averaging over C/E results
converges to true bias) or thick-tailed
(conservative approach: sample irradiation
and evaluation conditions
are not independent, outliers are
important pieces of information).
If the C/E variations are relatively
small or within plausible uncertainty
margins one can assume that the
randomness comes from a Gaussian
distribution with unknown mean and
variance. There are various statistical
tests available to check if this assumption
should be rejected. Table 4
shows, for example, that C/E results
for Cm-244 are more likely to be
Gaussian distributed than results for
Cs-137. If there is sufficient confidence
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Fuel pitch
(δ=0.005cm)
Surrounding
(depleted vs
reference)
Enrichment
(δ=0.05wt%)
| Tab. 3.
Relative uncertainties (%) due to irradiation boundary condition changes estimated in [38].
Fuel-T
(δ=50K)
Moderator-T
(δ=2K)
Burnup
(δ=2%)
Cm-244 1.2 3.0 2.4 0.1 0.7 9.0
Cm-243 1.2 2.0 1.1 0.4 0.7 5.3
Cm-242 1.0 1.0 0.6 0.4 0.4 3.3
Am-243 0.5 2.2 1.7 0.3 0.4 6.1
Am-241 1.2 2.0 0.5 0.9 0.6 0.2
Pu-242 0.1 1.0 1.3 0.2 0.0 4.4
Pu-241 1.2 1.0 0.0 0.7 0.6 1.2
Pu-240 0.4 0.0 0.3 0.2 0.3 1.6
Pu-239 1.4 0.0 0.5 0.7 0.6 0.1
Pu-238 1.1 1.0 0.2 0.2 0.6 4.3
Np-237 0.7 1.0 0.4 0.3 0.3 2.2
U-236 0.0 0.0 1.0 0.1 0.0 0.7
U-235 1.0 1.0 3.1 0.6 0.5 4.0
U-234 0.1 1.0 0.5 0.2 0.1 1.5
Eu-155 1.3 1.0 0.2 0.2 0.5 3.1
Eu-154 0.6 1.0 0.4 0.0 0.3 3.7
Eu-153 0.1 1.0 0.2 0.0 0.1 2.5
Sm-152 0.3 0.0 0.0 0.0 0.2 1.5
Sm-151 1.5 1.0 0.7 0.5 0.9 0.5
Sm-150 0.1 1.0 0.0 0.0 0.1 2.3
Sm-149 1.1 1.0 1.1 0.7 0.9 0.3
Sm-147 0.2 0.0 0.5 0.1 0.2 0.1
Pm-147 0.2 0.0 0.5 0.2 0.1 0.5
Gd-155 1.4 1.0 0.1 0.2 0.5 3.0
Cs-137 0.0 1.0 0.1 0.0 0.0 2.0
Cs-134 0.3 1.0 0.4 0.1 0.2 4.0
Cs-133 0.1 1.0 0.2 0.0 0.0 1.6
Ag-109 0.2 1.0 0.7 0.3 0.1 2.8
Rh-103 0.1 0.0 0.1 0.2 0.0 1.3
Ru-101 0.0 1.0 0.0 0.0 0.0 2.0
Tc-99 0.1 0.0 0.1 0.0 0.0 1.7
Mo-95 0.1 0.0 0.2 0.0 0.1 1.7
DECOMMISSIONING AND WASTE MANAGEMENT 359
Cm-244
Cs-137
Statistic P-Value Statistic P-Value
Anderson-Darling 0.208448 0.87023 Anderson-Darling 0.734219 0.0541662
Baringhaus-Henze 0.341385 0.790368 Baringhaus-Henze 0.706856 0.0646831
Cramér-von Mises 0.0295401 0.858128 Cramér-von Mises 0.121485 0.0568223
Jarque-Bera ALM 0.147449 0.928628 Jarque-Bera ALM 7.19179 0.0466512
Kolmogorov-Smirnov 0.00851119 0.952135 Mardia Combined 7.19179 0.0466512
Kuiper 0.0155151 0.93949 Mardia Kurtosis 1.58167 0.113726
Mardia Combined 0.147449 0.928628 Mardia Skewness 3.27423 0.0703758
Mardia Kurtosis -0 .239727 0.810542 Pearson x2 14.9452 0.0924521
Mardia Skewness 0.1006 0.751111 Shapiro-Wilk 0.968223 0.0621728
Pearson x2 51.712 0.170192
Shapiro-Wilk 0.999592 0.90984
Watson U2 0.0284101 0.836886
| Tab. 4.
Statistical tests to check if distributions of C/E for Cm-244 and Cs-137 in Figure 6 are consistent with a Gaussian distribution.
in the existence of a Gaussian process
governing the tests, the unknown
mean and variance can be estimated
with the usual maximum likelihood
method and the confidence interval
by using a Student-t distribution [39].
For the shown example of Cm-244 the
95 % confidence intervals are:
μ ∈ [0.96, 1.03], σ ∈ [0.12, 0.17],
Cs-137: μ ∈ [0.99, 1.02], σ ∈ [0.04,
0.06]. The results for μ can be interpreted
as systematic bias and can be
used to improve codes with empirical
factors or to confirm that updated,
microscopic cross sections result in
improved C/E values. For example,
the path to Cm-244 is through neutron
capture of Pu-242. In the thermal
energy range most evaluations refer to
cross sections from 1971 [40] and
1966 [41] and in ENDF/B-VII.1 and
JEFF-3.2, for example, evaluations
differ up to 20 %. Hence this cross
section would be a suitable candidate
for further improvement.
In previous research using Bayesian
updating [42,43] it has been demonstrated
that a combination of
infor mation from measurements of
microscopic data and from integral
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
DECOMMISSIONING AND WASTE MANAGEMENT 360
[46,47,48] to cover the large variations
of output parameters. In short,
the C/E data can be used for the
preparation of a set of upper-order
statistics and from it the characteristic
of threshold exceedances can be
deduced. The main distributional
model for exceedances over thresholds
is the generalized Pareto distribution
G ξ,β (x). For a given level u, a number
of N u datapoints will exceed the
threshold and the excesses are used to
fit the parameters of G by maximum
likelihood. The threshold is typically
determined from a mean excess
plot, see Figure 8 top (u≈0.2 for
Cm-244 and u≈0.1 for Cs-137 in this
example). The bottom of Figure 8
shows the Q-Q plots for both nuclides
together with the reference line from
fitted G ξ,β . The advantage of this
approach is that all the information of
the existing datapoints is used and
that very conservative, quantitative
estimates can be given how likely
unseen outliers or extreme values are.
The disadvantage is that there is no
explanation why the outliers exist.
Extreme value theory assumes that
more often than not unknowns in
irradiation conditions, code theory
and nuclear data and radiochemical
sample analysis do not fortuitously
cancel each other out.
| Fig. 8.
Distribution of excesses for Cm-244 and Cs-137 (top) and Q-Q plot using a Generalized Pareto
distribution as reference.
tests like the above can lead to an
improvement of microscopic data. One
of the objectives of the EURAD work
package 8 subtask 2 is to provide highly
accurate integral test results and
provide recommendations for nuclear
data that need to be improved.
The other alternative to interpret a
relatively thin database is to embed it
into a thick-tailed model (i.e. a model
which allows higher probabilities
for events outside of conventional
domain). This can be reasonable for
three purposes: first, observed outliers
cannot be discarded and are a hint for
unidentified sources of uncertainty.
Second, in some applications simulation
tools must make predictions
in parameter ranges which are not
accessible by current experiments and
prudence and conser vatism is important.
Third, the system belongs to the
complex class of systems in which
often small changes of boundary
parameters can have over proportionally
large effects on results
[44,45]. In these cases, the methods
of extreme value theory can be applied
5 Conclusion
A large database of single effects tests
and integral tests has been built for
source term validation since the start
of the civil nuclear programs. Efforts
were mainly focused on criticality
safety, burnup credit and decay heat.
There is little coherence between
these efforts and requirements
concerning long-term storage only
recently received higher priority.
Increasing the accuracy of existing
source term predictions faces several
hurdles:
p Different source terms and different
time scales require setting
different priorities on nuclides.
Resource constraints exist to complement
existing data.
p High quality tests for measurement
of source terms are scarce and
significantly improving knowledge
about irradiation boundary conditions
for most samples of commercial
fuel appears unrealistic at
the moment.
p Many integral tests show relatively
large differences between measurements
and theory which cannot
easily be explained by known
uncertainties of microscopic data
and irradiation conditions.
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Among others, research in the EURAD
WP8 subtask 2 addresses these issues
by:
p Reevaluating data from samples
from commercial fuel for which
irradiation boundary conditions
are known with relatively high
accuracy.
p Detailed sensitivity analysis to
define reliable uncertainty margins
for nuclide inventory and corresponding
source terms predictions
and identify nuclear data requirements
to improve the predictive
power of codes.
p Identifying a potential for improvement
of the robustness of industrystandard
code predictions. Both by
embedding existing C/E results
into a suitable statistical framework
and by comparison with
latest, sophisticated codes.
Acknowledgement
Co-funding from European Commission
under Grant Agreement number
847593 is highly acknowledged.
References
1. SKB Technical Report TR-11-01, Long-term safety for the final
repository for spent nuclear fuel at Forsmark, Main report of
the SR-Site project, Stockholm, March 2011.
2. SKB Technical Report TR-17-02, Basis for fees and guarantees
for the period 2018–2020, Stockholm, April 2017.
3. EURAD Deployment Plan 2019-2024,
https://www.ejp-eurad.eu/sites/default/files/2020-01/4.%20
EURAD%20Deployment%20Plan.pdf
4. B. T. Rearden and M. A. Jessee, Editors, SCALE Code System,
ORNL/TM-2005/39, Version 6.2, Oak Ridge National
Laboratory, Oak Ridge, TN, 2017.
5. International Handbook of Evaluated Criticality Safety
Benchmark Experiments, NEA/NSC/DOC(95)03, Nuclear
Energy Agency/OECD, Paris, France, 2016.
6. M.C. Brady et al, International Studies on Burn-up Credit
Criticality Safety by an OECD/NEA Working Group, Intl.
Conference on the Physics of Nuclear Science & Technology,
Long Island, NY, 5-8 Oct. 1998.
7. Burnup credit criticality safety benchmark – Phase VII, UO2
Fuel: Study of spent fuel compositions for long-term disposal,
NEA No. 6998, Nuclear Energy Agency/OECD, Paris, France,
2012.
8. G. Ilas, I. C. Gauld, and B. D. Murphy, Analysis of Experimental
Data for High Burnup PWR Spent Fuel Isotopic Validation –
ARIANE and REBUS Programs UO2 Fuel, NUREG/CR-6969,
ORNL, Tenn., February 2010.
9. G. Ilas, I. C. Gauld, F. C. Difilippo, and M. B. Emmett, Analysis
of Experimental Data for High Burnup PWR Spent Fuel Isotopic
Validation - Calvert Cliffs, Takahama, and Three Mile Island
Reactors, NUREG/CR-6968, Oak Ridge National Laboratory,
Oak Ridge, TN, February 2010.
10. I.C. Gauld, U. Mertyurek, Validation of BWR spent nuclear fuel
isotopic predictions with applications to burnup credit, Nuc.
Eng. Des., Vol. 345, pp.110-124, 2019.
11. M. B. Chadwick et al., ENDF/B-VII.1 Nuclear Data for Science
and Technology: Cross Sections, Covariances, Fission Product
Yields and Decay Data, Nuclear Data Sheets, Vol. 112,
pp. 2887–2996, 2011.
12. M. Salvatores et al., Needs and Issues of Covariance Data
Application, Nuclear Data Sheets, Vol. 109, pp. 2725–2732,
2008.
13. D. Rochman et al., Nuclear Data Uncertainty Propagation:
Total Monte Carlo vs. Covariances, J. Korean. Phys. Soc.,
Vol. 59, pp. 1236-1241, 2011.
14. D. Rochman, C. Sciolla, Nuclear data uncertainty propagation
for a typical PWR fuel assembly with burnup, Nuc. Eng. Tech.,
Vol. 46, pp. 353-362, 2014.
15. D. Rochman et al., Nuclear Data Uncertainties for Typical LWR
Fuel Assemblies and a Simple Reactor Core, Nuclear Data
Sheets, Vol. 139, pp. 1-76, 2017.
16. F. Michel-Sendis et al., SFCOMPO-2.0: An OECD/NEA
database of spent nuclear fuel isotopic assays, reactor design
specifications, and operating data, Ann. Nucl. Energy, Vol. 110,
pp. 779–788. 2017.
17. B.B. Bevard at al., Review of information for spent nuclear fuel
burnup confirmation, NUREG/CR-6998, Oak Ridge National
Laboratory, Oak Ridge, TN, 2009.
18. A. Tanskanen, Assessment of the neutron and gamma sources
of the spent BWR fuel, STUK report STUK-YTO-TR 170, Helsinki,
2000.
19. A. Rimpler, Bonner sphere neutron spectrometry at spent fuel
casks, Nucl. Instruments Methods Phys. Res. Sect. A Accel.
Spectrometers, Detect. Assoc. Equip., Vol. 476, pp. 468–473,
2002.
20. G. Ilas, I.C. Gauld, SCALE analysis of CLAB decay heat measurements
for LWR spent fuel assemblies, Ann. Nucl. Energy,
Vol. 35, pp. 37–48. 2008.
21. T. Yamamoto, D. Iwahashi, Validation of decay heat calculation
results of ORIGEN2.2 and CASMO5 for light water reactor fuel,
J. Nucl. Sci. Technol., Vol. 53, pp. 2108-2118, 2016.
22. ANSI/ANS-5.1-2005, Decay Heat Power In Light Water
Reactors, ANSI, Washington, DC, 2005.
23. DIN-25463-1:1990-05, Decay heat power in nuclear fuels
of light water reactors; non-recycled nuclear fuels, DIN e.V.,
Berlin, 1990.
24. SKB report R-05-62, Measurements of decay heat in spent
nuclear fuel at the Swedish interim storage facility, Clab,
Stockholm, 2006.
25. G. Ilas, H. Liljenfeldt, Decay heat uncertainty for BWR used
fuel due to modeling and nuclear data uncertainties, Nucl.
Eng. Des., Vol. 319, pp. 176–184, 2017.
26. I.C. Gauld et al., Validation of SCALE 5 decay heat predictions
for LWR spent nuclear fuel, NUREG/CR-6972, Oak Ridge
National Laboratory, Oak Ridge, TN, 2010.
27. I.C. Gauld et al., Uncertainties in predicted isotopic
compositions for high burnup PWR spent nuclear fuel,
NUREG/CR-7012, Oak Ridge National Laboratory, Oak Ridge,
TN, 2011.
28. IAEA-TECDOC-1241, Implementation of burnup credit in
spent fuel management systems, International Atomic Energy
Agency, Vienna, 2001.
29. H.R. Trellue, M.L. Fensin, J.D. Galloway, Production and
depletion calculations using MCNP, LA-UR-12-25804,
Los Alamos, NM, 2012.
30. M. Akiyama, S. An, Measurements of Fission-product Decay
Heat for Fast Reactors, Proc. Int. Conf. on Nuclear Data for
Science and Technology, pp. 237-244, Antwerp, Belgium,
1982.
31. J.K. Dickens, T.A. Love, J.W. McConnell, Fission-product Energy
Release for Times Following Thermal-neutron Fission of 239Pu
and 241Pu Between 2 and 14 000 Seconds, Nucl. Sci. Eng.,
Vol. 78, pp. 126-146, 1981.
32. NEA/WPEC-25, International evaluation cooperation volume
25, Assessment of fission product decay data for decay heat
calculations, Report NEA no. 6284, Nuclear Energy Agency/
OECD, Paris, 2007.
33. F. Schmittroth, Uncertainty analysis of fission-product
decay- heat summation methods, Nucl. Sci. Eng., Vol, 59,
pp. 117-139, 1976.
34. F. Schmittroth, R.E. Schenter, Uncertainties in fission product
decay-heat calculations, Nucl. Sci. Eng., Vol. 63, pp. 276-291,
1977.
35. G. Radulescu, I.C. Gauld, G. Ilas, SCALE 5.1 Predictions of PWR
Spent Nuclear Fuel Isotopic Compositions, ORNL/TM-2010/44,
Oak Ridge National Laboratory, Oak Ridge, TN, 2010.
36. W.J. Marshall et al., Development and Testing of Neutron
Cross Section Covariance Data for SCALE 6.2, Proceedings of
International Conference on Nuclear Criticality Safety,
Charlotte, NC, 2015.
37. R. Capote et al., Unrecognized sources of uncertainties (USU)
in experimental nuclear data, Nuclear Data Sheets, Vol. 163,
pp. 191-227, 2020.
38. NEA/NSC/WPNCS/DOC(2011)5, Spent nuclear fuel assay data
for isotopic validation, State-of-the-art report, Nuclear Energy
Agency/OECD, Paris, 2011.
39. Student, The probable error of a mean, Biometrika, Vol. 6(1),
pp. 1-25, 1908.
40. T.E. Young et al., The low-energy total neutron cross section
of plutonium-242, Nucl. Sci. Eng., 43, pp. 341-342, 1971.
41. G.F. Auchampaugh et al., Neutron total cross section of Pu242,
Phys. Rev., Vol 146, p.840, 1966.
42. A.J. Koning, Bayesian Monte Carlo method for nuclear data
evaluation, Eur. Phys. J. A, Vol. 51, 184, 2015.
43. E. Alhassan, Bayesian updating for data adjustments and
multi-level uncertainty propagation within Total Monte Carlo,
Ann. Nucl. Energy, Vol. 139, 107329, 2020.
44. A. Majdandzic et al., Spontaneous recovery in dynamical
networks, Nature Physics, Vol. 10, pp. 34-38, 2013.
45. S.V. Buldyrev et al., Catastrophic cascade of failures in
interdependent networks, Nature, Vol. 464, pp. 1025-1028,
2010.
46. P. Embrechts, C. Klüppelberg, T. Mikosch, Modelling extremal
events for insurance and finance, Springer-Verlag, Berlin, 1997.
47. J. Pickands, Statistical inference using extreme order statistics,
Ann, Statist., Vol. 3, pp. 119-131, 1975.
48. E.J. Gumbel, Statistics of Extremes, Columbia University Press,
New York, 1958.
49. I.C. Gauld, G. Ilas, G. Radulescu, Uncertainties in predicted
isotopic compositions for high burnup PWR spent nuclear fuel,
ORNL/TM-2010/41, Oak Ridge National Laboratory, Oak
Ridge, TN, 2010.
50. G. Zerovnik et. al., Observables of interest for the
characterization of spent nuclear fuel, JRC Technical Report JRC
112361, Luxembourg, 2018.
51. R. Jacqmin, M. Kellett, A. Nouri, The JEFF-3.0 Nuclear Data
Library, PHYSOR 2020, Seoul, Korea, 2002.
Authors
Marcus Seidl
PreussenElektra GmbH
marcus.seidl@preussenelektra.de
Tresckowstraße 5
30457 Hannover, Germany
Peter Schillebeeckx
EC Joint Research Center
Retieseweg 111
2440 Geel, Belgium
Dimitri Rochman
Reactor Physics and Thermal
Hydraulic Laboratory
Paul Scherrer Institut
Villingen, Switzerland
DECOMMISSIONING AND WASTE MANAGEMENT 361
Decommissioning and Waste Management
On the Potential to Increase the Accuracy of Source Term Calculations for Spent Nuclear Fuel from an Industry Perspective ı Marcus Seidl, Peter Schillebeeckx and Dimitri Rochman

atw Vol. 65 (2020) | Issue 6/7 ı June/July
DECOMMISSIONING AND WASTE MANAGEMENT 362
Experimental Investigations into Flow
Conditions of Konrad Exhaust Air Channel
Steffen Wildgrube, Anton Anthofer, Michael Haas, Alexander Kratzsch and Clemens Schneider
The minesite Konrad is going to be converted into a final storage facility for solid or consolidated radioactive waste with
negligible heat generation. To investigate the flow in the exhaust channel “chimney” a test facility with 1:5 scaled mockup
was built. 2D-PIV measurement technology was used to analyze the flow at the envisaged sample taking point. The
main purpose of the tests was to forecast if the different criteria for homogenous flow defined by DIN ISO 2889 could be
met. Two test parameters have been examined: (total) air volume flow and particle size. Only one of three investigated
criteria was passed for all particle sizes and volume flows. Further investigations into adaptions of the exhaust channel
“chimney” are necessary to fulfill all requirements for homogenous flow at the sample taking spot for all particle sizes
and all (normal) operation status.
| Fig. 1.
Schematic layout of the exhaust channel “chimney“.
Introduction
The minesite Konrad is going to be
converted into a final storage facility
for solid or consolidated radioactive
waste with negligible heat generation.
To prohibit unacceptable emission of
activity by exhaust air from underground
facilities (exhaust channel
“diffuser”) or aboveground facilities
(exhaust channel “chimney”), the exhausted
air is monitored continuously
by taking samples. For a representative
sample taking it is necessary
to have (among other boundary conditions)
a homogenous flow where
the sample is taken according to DIN
ISO 2889.
The current design of the exhaust
channel “chimney” is not optimal
for taking samples. Therefore, the
operator BGE (Bundesgesellschaft für
Endlagerung) mbH want to ensure
that the sample taking in the exhaust
channel “chimney” will be conforming
to the standards after it is built.
For this reason, BGE assigned the
VPC Nukleare Dienstleistungen GmbH
(VND) to investigate the flow conditions
in the exhaust channel
“ chimney” by using a mock-up in a
reduced scale of 1 to 5. Multiple experiments
were performed including
all relevant operating status and
particle sizes. To minimize influences
by the measurement setup 2D-particle
imaging velocimetry (2D-PIV) was
used as major measurement device.
Layout of the test facility
The exhaust channel “chimney” has a
cross sectional area of 4.40 m x 2.20 m
at the envisaged sampling point.
Before and after the sampling point
the exhaust channel is split into
two equal subchannels with a cross
section of 2.20 m x 2.20 m each. The
undisturbed entry length before the
sampling point is circa 16 m long. This
equals a 5.8 fold hydraulic diameter of
the channel.
To investigate the flow conditions in
the exhaust channel “chimney” a test
facility was erected at the Zittau/
Goerlitz University of Applied Sciences
(HSZG). The test facility contains a
mock-up of the exhaust channel “chimney”
in a scale of 1:5. The mock-up
includes the relevant area before the
sampling point till the ventilators and
after the sampling point till the beginning
of the chimneys. To achieve fluidic
similarity all important installations of
the channel like fire flaps were considered
in the mock-up (see Figure 1).
Three relevant operational status
were defined for the exhaust channel
“chimney”. These are “Reposition
min.”, “Reposition max.” and “Reposition
max. + frost”. The operational
status differs in the corresponding air
volume flow. Regarding to these
volume flows average flow velocities
can be derived. The correlation for the
different operational status is given in
Table 1.
To achieve similar particle behavior
in the exhaust channel and in the
mock-up the Stokes similarity number
should be in the same range. The
Stokes number expresses the behavior
of particles suspended in a fluid flow.
Operational
status
V [m 3 /s]
| Tab. 1.
Volume flow and flow velocities for relevant
operational status.
In other words, it describes how fast a
particle can adopt to changes in the
surrounding fluid flow expressed by
the ratio of the characteristic time of a
particle to the characteristic time of
the flow. A Stokes number much
smaller than one indicates that the
particles follow the surrounding fluid
flow (streamlines) closely. Adopted to
the given scenario here the Stokes
number can be expressed as follows:
| Form. 1.
Stokes number (= 1.82*10-5 kg/m/s).
v − [m/s]
Reposition min. 32 3.3
Reposition max. 40 4.1
Reposition max.
+ frost
46 4.8
Hereby ρ P means the particle density,
d ae the aerodynamic particle diameter
(AED), v¯ the average fluid flow
Decommissioning and Waste Management
Experimental Investigations into Flow Conditions of Konrad Exhaust Air Channel ı Steffen Wildgrube, Anton Anthofer, Michael Haas, Alexander Kratzsch and Clemens Schneider

atw Vol. 65 (2020) | Issue 6/7 ı June/July
AED of
original particles
| Tab. 2.
Chosen particles for test facility.
Material of
scaled particles
velocity, η L the dynamic viscosity of
air and L U the length of directional
change of the flow in the channel.
In the exhaust channel “chimney”
particles with an aerodynamic
diameter of up to 60 µm are possible.
Nevertheless, particles with an AED
of 1, 5 and 20 µm are most important
for the operator of the facility and
have been chosen for closer investigation.
Therefore, they have been scaled
to the mock-up by using the Stokes
number. By applying a value for L U of
0,1 m the Stokes numbers in a range
of 1*104 to 5*101. As a result, the
following particle sizes have been
chosen to be used in the test facility
(see Table 2).
Setup of the test facility
The facility was built with an open
cycle layout. The particles are
generated by using a dust disperser.
The disperser is located at the open
front end of the facility. Therefore, the
particles are injected directly in the
circular suction channel made of
metal sheet with a diameter of circa
500 mm. The suction channel is split
afterwards in two mirror inverted
channel. They have also a diameter of
circa 500 mm and are made of metal
sheet. The suction channels are
connected to two equal ventilators
that are operated in parallel to
generate a total volume flow in a
range of 32 to 46 m³/s. During the
experiments that have been run so far,
the ventilators have been operated
always with the same load (normal
operation status of the future exhaust
channel “chimney”). For future
Density of
scaled particles
| Fig. 2.
Test facility with a mock-up of the exhaust channel “chimney“ in scale 1:5.
AED of
scaled particles
1 µm Aluminium oxide 4 g/cm 3 0.4 µm
5 µm Silicium dioxide 2 g/cm 3 2 µm
20 µm Borosilicate glass 1 g/cm 3 10 µm
experiments it is also possible to
run the ventilators with different
loads to simulate abnormal operational
status (malfunction of ventilators,
maintenance etc.). The exits
of the ventilators are connected to
440 mm x 440 mm rectangular
channels made of metal sheet. From
this point on the dimensions of the
construction elements are scaled 1:5
to the original exhaust channel
“ chimney” planed for the final disposal
facility Konrad.
By using a Y-section the two channels
are united to a 880 mm x 440 mm
rectangular channel. This channel
section is the most important part of
the facility for the experiments and
will be named as main channel in the
following text. It is made of multiple
channel sections with different length.
Two of these sections have a top and a
side wall made of acrylic glass to
ensure optical transparency for the
2D-PIV measurement. Depending on
the arrangement of the different
| Fig. 3.
Coordinate system of the test facility (not to scale).
sections the 2D-PIV measurement can
be performed at different locations of
the channel. The results presented in
this article are based on measurement
data taken at the actual planned
sample taking point, which is at
z=3200 mm in the test facility.
At the end of the main channel a
second Y-section separated the main
channel again in two subchannels of
440 mm x 440 mm. After 90° elbows
the subchannels are directed to a
rectangular header. At this header
4 circular pipes with a diameter of circa
400 mm are connected, that simulate
the (shortened) chimneys. The pipes
end in a two staged HEPA filter to
ensure the retention of the dispersed
particles in the facility. Beyond the
filter the cleaned air flow is emitted to
the laboratory. The whole test facility
can be seen in Figure 2.
According to the planned layout of
the exhaust channel “chimney” in
Konrad multiple installations like fire
flaps and flow grids are installed in the
test facility at their assigned place as
dummy installation in scale 1:5.
The test facility was built in the
laboratories of the Institute of Process
Technology, Process Automation and
Measuring Technology (IPM) of the
Zittau/Goerlitz University of Applied
Sciences (HSZG).
The coordinate system of the facility
refers to the main channel and was
defined as follows (see Figure 3):
p x-axis: The x-axis defines the
width of the main channel
beginning from the left
channel wall (in flow
direction).
p y-axis: The y-axis defines the
height of the main channel
beginning from the bottom
wall of the channel.
p z-axis: The z-axis defines the
length of the main channel
in flow direction. It begins
(z=0 mm) after the first
Y-section.
DECOMMISSIONING AND WASTE MANAGEMENT 363
Decommissioning and Waste Management
Experimental Investigations into Flow Conditions of Konrad Exhaust Air Channel ı Steffen Wildgrube, Anton Anthofer, Michael Haas, Alexander Kratzsch and Clemens Schneider

atw Vol. 65 (2020) | Issue 6/7 ı June/July
DECOMMISSIONING AND WASTE MANAGEMENT 364
Measurement equipment
The local flow velocity of the
air was measured with a spherical
anemometer. The uncertainty of the
anemometer is below ±0.28 m/s at
flow velocities smaller than 5 m/s.
| Fig. 4.
2D-PIV measurement setup in the test facility (example).
The characteristics of the particle
flow have been analyzed by using a
2D-PIV measurement system. For
using a 2D-PIV system a light section
is created with a laser at the optical
transparent section of the channel.
The particles dispersed in the flow
reflect the light of the laser in any
direction. Due to this the particles
can be seen on pictures taken with a
camera orientated 90° to the laser. By
using a high-speed camera and taking
at least two pictures in a very short
time the movement of the particles
can be derived by comparing the
pictures. A special PIV software is
taken for this kind of image processing
and to calculate particle speed, orientation
and distribution (within the
light section). The used 2D-PIV system
consists of one laser and one camera.
Therefore, within one measurement
setup one certain x-z or y-z plane can
be investigated depending on the
position of the laser (and the camera).
To achieve information for the x-y
cross-sectional area at z=3,200 mm
(envisaged sample taking point)
3 different horizontal and 5 different
vertical measurement setups (laser
positions) have been chosen (see
Figure 4).
Results
The homogeneity of the flow was evaluated
according to the criteria of DIN
ISO 2889. The following properties
were examined:
p Velocity profile of the air flow
p Vorticity of the air flow
p Velocity profile of the particle flow
p Concentration profile of aerosol
particles
| Fig. 5.
Measuring points of the anemometer measurement.
| Fig. 6.
Horizontal velocity profile of air) and particles (2D-PIV measurement) with the respective standard deviation.
Experiments without particles
At first, the air flow in the channel was
analysed without particle transport.
For this purpose, the flow velocity
of the air was recorded by means
of a spherical anemometer at
15 measuring points evenly distributed
over the cross section of the
channel (see Figure 5).
In all three set operating con ditions
a reduction of the flow velocity in the
middle of the horizontal flow profile
could be detected. This results in the
formation of a “double hump profile”
(see Figure 6). This is probably
attributed to a too short inlet length.
This leads to the fact that the flow profile
cannot evolve completely after the
two partial channels have been
merged. Furthermore, it was determined
that the flow velocity is
highest in the lower channel area
and decreases towards the top. This
is also attributed to the short inlet
length as well as the 90° elbow before
the Y-section, which can lead to
negative effects at the planned
sampling location. Further investigations
using CFD analysis are ongoing
Decommissioning and Waste Management
Experimental Investigations into Flow Conditions of Konrad Exhaust Air Channel ı Steffen Wildgrube, Anton Anthofer, Michael Haas, Alexander Kratzsch and Clemens Schneider

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Particles Operating status Criteria:
non­ turbulent flow
to get deeper knowledge to verify
these assumptions.
In the experiments without particles,
the coefficient of variation
(COV) of the flow velocity was in the
range of 11 % to 13 %. Thus, the
homogeneity criterion of less than
20 % is maintained under all simulated
operating conditions. However,
the currently planned position of the
particle sampler is in the middle of the
channel. Therefore, it would be at a
position with reduced flow velocity.
A sampling at this position would
not be conservative.
Experiments with particles
Test series were carried out with
particle sizes of 0.4 µm, 2 µm
and 10 µm at different stationary
operating conditions. The particle
velocity and distribution in the
channel was measured by 2D-PIV
in several planes, both horizontally
( x-z planes) and vertically (y-z
planes). By overlaying the results
from the horizontal and vertical
planes, the particle velocity and distribution
in the channel cross section
(x-y-plane) at the planned sampling
point (z=3,200 mm) could be determined.
The flow velocity profile of the
particles (PIV measurement) and
the air (anemometer measurement) is
qualitatively identical (see Figure 6),
i.e. the considered particles follow the
air flow and a corresponding double
hump profile of the particle flow
velocities is formed.
Quantitatively, the variation of the
flow velocity of the particles over the
cross section is partially, i.e. especially
for the “larger” 10 µm particles, higher
comparing to air. As a result, the
coefficient of variation of the particle
velocity for the 10 µm particles is over
20 % and thus the homogeneity
criterion based on DIN ISO 2889 is
not fulfilled.
In addition to the uniform velocity
distribution over the channel cross
section, the even distribution of
the particles in general is a crucial
criterion to verify a homogeneous
particle flow. For this purpose, the
variation coefficient of the particle
concentration over the channel
cross-section must be below 20 %.
During the experiments, this criterion
was only met for the smallest particles
(aerodynamic diameter of 0.4 µm)
under all operating conditions.
While a homogeneous particle distribution
can still be determined for
the particles with an aerodynamic
diameter of 2 µm depending on the
operating status, the coefficient of
variation of the particle concentration
for the large particles (aerodynamic
diameter of 10 µm) is above 20 % in
all operating status.
The criterion for a vorticity free
flow is met, when the angular
deviation of the flow velocity to
the z-axis is below 20°. This criterion
was achieved for all particle sizes in
all operating conditions.
Summary and outlook
The homogeneity of the expected
particle flow in the envisaged exhaust
channel “chimney” of the final disposal
facility Konrad was investigated.
Therefore, a test facility was erected
at the HSZG including a mock-up of
the channel in scale of 1:5. The air
flow and the particle flow in the test
facility were measured by using a
spherical anemometer and a 2D-PIV
measurement system. Experiments
were conducted representing all
relevant operating status of the final
disposal facility and the relevant
particle sizes. Concerning the criteria
given in DIN ISO 2889 for a homogeneous
flow the following results
could be reached (see Table 3):
Based on these findings the
following options arise:
A) Constructive changes in design of
the channel
B) Adopt the sample taking concept
e.g. design change of sampler
After consultations with the operator
option A was chosen for further
investigations as a first step. To avoid
changes of the overall channel design,
the influence of installations into the
channel to homogenize the flow will
be subject for future work. To identify
the impact of different options they
are analyzed by using CFD methods.
The most promising option will be
installed in scale 1:5 in the test facility.
Authors
Criteria: homogenous
velocity distribution
None (Pre-experiment) Reposition min. n/a Passed n/a
None (Pre-experiment) Reposition max. n/a Passed n/a
None (Pre-experiment) Reposition max. + frost n/a Passed n/a
Aluminium oxide (AED 400 nm) Reposition min. Passed Passed Passed
Aluminium oxide (AED 400 nm) Reposition max. Passed Passed Passed
Aluminium oxide (AED 400 nm) Reposition max. + frost Passed Passed Passed
Silicium dioxide (AED 2 µm) Reposition min. Passed Passed Passed
Silicium dioxide (AED 2 µm) Reposition max. Passed Passed Passed
Silicium dioxide (AED 2 µm) Reposition max. + frost Passed Passed Not passed
Borosilicat glass (AED 10 µm) Reposition min. Passed Passed Not passed
Borosilicat glass (AED 10 µm) Reposition max. Passed Not passed Not passed
Borosilicat glass (AED 10 µm) Reposition max. + frost Passed Not passed Not passed
| Tab. 3.
Overview of current results.
Dr. Steffen Wildgrube
steffen.wildgrube@vpc-group.biz
Dr. Anton Anthofer
VPC Nukleare Dienstleistungen
GmbH (VND)
Fritz-Reuter-Straße 32 c
01097 Dresden, Germany
Michael Haas
Bundesgesellschaft
für Endlagerung mbH (BGE)
Eschenstraße 55
31224 Peine, Germany
Prof. Dr. Alexander Kratzsch
Dr. Clemens Schneider
Hochschule Zittau/Görlitz
Theodor-Körner-Allee 16
02763 Zittau, Germany
Criteria: homogenous
particle distribution
DECOMMISSIONING AND WASTE MANAGEMENT 365
Decommissioning and Waste Management
Experimental Investigations into Flow Conditions of Konrad Exhaust Air Channel ı Steffen Wildgrube, Anton Anthofer, Michael Haas, Alexander Kratzsch and Clemens Schneider

atw Vol. 65 (2020) | Issue 6/7 ı June/July
366
KTG INSIDE
Inside
KTG Junge Generation –
Kamingespräch bei Framatome
| Carsten Haferkamp (2. v. l.), Managing Director bei Framatome GmbH,
mit Teilnehmenden beim Kamingespräch in Erlangen
Die aktuelle Corona-Krise beeinflusst alle Bereiche
unseres Lebens, auch die Zeiträume für Berichte und
Nachrichten. Aus diesem Grund erscheinen die Berichte
der Jungen Generation derzeit mit Verzögerung. Bereits
am 5. März dieses Jahres fand erneut unser Kamingespräch
statt. Diesmal hatten Studenten und Young Professionals
die Möglichkeit, auf Einladung von Carsten Haferkamp,
Geschäftsführer der Framatome GmbH, nach Erlangen zu
kommen.
Auch diese Veranstaltung wurde bereits durch die
Vorzeichen der Krise beeinflusst und wurde vielleicht
deshalb zu etwas Besonderem. Die Vorbereitung nahm
einige Zeit ein, da die Terminkalender der Führungskräfte
meist gut gefüllt sind und diese gerade in einer solchen
Krise besonders gefordert sind. Doch trotz all der Vorzeichen
nahm sich Carsten Haferkamp die Zeit, um sich
mit den Teilnehmern in der Gaststätte „Alter Simpl“ in
Erlangen zu treffen.
Hier erfuhren die Teilnehmer viel Interessantes über die
aktuellen und geplanten Aktivitäten von Framatome und
über Carsten Haferkamp selbst. Es wurde auch über die
Beschäftigungschancen in der Branche gesprochen und
über die Zukunft der Kerntechnik in Deutschland. Auch
wenn die Vorzeichen in der Tagespresse und öffent lichen
Meinung negativ erscheinen, gibt es doch span nende und
zukunftsträchtige Perspektiven in Deutschland.
Die Rückmeldung der Teilnehmer nach der Veranstaltung
war sehr positiv. Gerade die Möglichkeit, alle
Fragen stellen zu können und auch offene Antworten zu
erhalten wurde sehr positiv angenommen. Dass eine Krise
auch Spontanität macht, zeigte sich auch am Nachmittag
vor dem Kamingespräch. Thomas Hahn, Vice President
Customer Relationship, organisierte kurzfristig noch eine
Führung auf dem Gelände von Framatome, bei dem die
Mehrzahl der Teilnehmer Testanlagen besuchen konnten.
Der Vorstand der Jungen Generation der KTG möchte
sich daher in besonderem Maße bei Carsten Haferkamp
und Thomas Hahn bedanken, trotz der Vorzeichen das
Kamingespräch ermöglicht zu haben.
Thomas Romming
Stellvertretender Sprecher der Jungen Generation
Nachruf
Johann Waldmann
Wach im Geist und streitbar, trotz zunehmender
körperlicher Beeinträchtigung, so hat man
Johann Waldmann bei seinen Vorträgen und
Diskussionen in den letzten Jahren erlebt.
Nach schwerer Krankheit ist unser langjähriger
Mitstreiter und Gründungsmitglied von uns gegangen.
Diskussionen ohne ihn waren eine Seltenheit, aber er
hatte wirklich immer etwas zum Thema zu sagen.
Er konnte zornig sein bei Unexaktheit und pauschalem
Geschwafel, jedoch milde bei Unwissenheit.
Wenn einer etwas von ihm wissen wollte, schaute
er nicht auf die Uhr.
Er wird uns auf unseren Tagungen fehlen, aber wir
werden weiterhin auch in seinem Sinne für eine
realistische und sichere Energieversorgung kämpfen.
Kerntechnische Gesellschaft e. V. (KTG)
Fachgruppe Nutzen der Kerntechnik und Energiesysteme
2. August 2019 ı
Dipl.-Ing. Johann Pisecker
Tulln
23. Februar 2020 ı
Dipl.-Ing. Hubert Andrae
Rösrath
30. Mai 2020 ı
Dr. Hans Schuster
Aachen
Die KTG verliert in ihnen langjährige
aktive Mitglieder, denen sie ein
ehrendes Andenken bewahren wird.
Ihrer Familie gilt unsere Anteilnahme.
KTG Inside

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Herzlichen Glückwunsch!
Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag
und wünscht ihnen weiterhin alles Gute!
Juli 2020
70 Jahre | 1950
01. Prof. Dr. Helmut Keutner,
Oberkrämer OT Schwante
04. Dr. Gerhard Eiselt, Groß-Umstadt
19. Dipl.-Ing. Gerhard Hanetzog, Peine
75 Jahre | 1945
13. Prof. Dr. Eckhard Rückl,
Bodenwerder
15. Walter Burchhardt, Karlsruhe
76 Jahre | 1944
17. Dipl.-Ing. Jürgen Krellmann,
Le Puy Ste. Réparade/FR
20. Günter Langer, Rosbach
77 Jahre | 1943
10. Dipl.-Ing. Dieter Eder, Alzenau
80 Jahre | 1940
31. Dr. Peter Schneider-Kühnle, Worms
81 Jahre | 1939
23. Heinz Stahlschmidt, Erlangen
26. Dipl.-Ing. Ewald Passig, Bochum
82 Jahre | 1938
30. Dr. Philipp Dünner, Odenthal
83 Jahre | 1937
06. Dipl.-Ing. Paul Börner, Steinau-Uerzell
29. Dr. Herbert Reutler, Köln
86 Jahre | 1934
14. Prof. Dr. Walter-H. Köhler, Wien/AT
88 Jahre | 1932
24. Dipl.-Ing. Joachim May, Burgwedel
27. Dr. Rainer Schwarzwälder, Glattbach
31. Dr. Theodor Dippel,
Eggenstein-Leopoldsh.
August 2020
40 Jahre | 1980
07. Dipl.-Ing. Ingmar Koischwitz, Reken
60 Jahre | 1960
23. Dipl.-Ing. Hans-Werner Fedler,
Leverkusen
76 Jahre | 1944
24. Dr. Gerd Uhlmann, Dresden
29. Dipl.-Phys. Harald Scharf,
AX Goes/NL
78 Jahre | 1942
28. Dipl.-Ing. Hans-J. Fröhlich, Berzhahn
79 Jahre 1941
17. Dipl.-Ing. Jörg-Hermann Gutena,
Emmerthal
21. Dipl.-Phys. Peter Kahlstatt, Hameln
81 Jahre | 1939
01. Dipl.-Ing. Gerhard Becker,
Neunkirchen-Seelscheid
29. Dr.-Ing. E. h. Adolf Hüttl,
Monte Estoril (Parque Palmela)/PT
82 Jahre | 1938
06. Prof. Dr. Rudolf Avenhaus, Baldham
21. Dr. Gerhard Schücktanz, Altdorf
84 Jahre | 1936
31. Dr. Hartwig Poser, Radeberg-Rossendorf
85 Jahre | 1935
29. Dr. Hans-Jürgen Engelmann, Peine
86 Jahre | 1934
15. Dipl.-Phys. Heinrich Glantz,
Eggenstein-Leopoldsh.
91 Jahre | 1929
02. Dipl.-Phys. Wolfgang Schwarzer,
Weilerswist
96 Jahre | 1924
01. Prof. Dr. Wolfgang Stoll, Hanau
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.
Lektorat:
Natalija Cobanov,
Kerntechnische
Gesellschaft e. V.
(KTG)
Robert-Koch-Platz 4
10115 Berlin
T: +49 30 498555-50
F: +49 30 498555-51
E-Mail:
natalija.cobanov@
ktg.org
www.ktg.org
367
KTG INSIDE
Nachruf
Willi Marth
Dr. Willy Marth ist im April 2020
in Karlsruhe verstorben.
Willy Marth, geboren 1933 im Fichtel gebirge,
promovierte in Physik an der Technischen
Hochschule in München und erhielt
anschließend ein Diplom in Betriebswirtschaft
der Universität München. Ein Post-
Doc- Aufenthalt in den USA vervollständigte seine Ausbildung. Am „Atomei“
FRM in Garching war er für den Aufbau der Bestrahlungseinrichtungen
verantwortlich, am FR 2 in Karlsruhe für die Durchführung der Reaktorexperimente.
Als Projektleiter wirkte er bei den beiden natrium gekühlten
Kernkraftwerken KNK I und II, sowie bei der Entwicklung des Schnellen
Brüter SNR 300 in Kalkar. Ab Oktober 1978 war er Leiter des Projektes
Schneller Brüter. Beim europäischen Brüter EFR war er ab November 1989
als Executive Director zuständig für die gesamte Forschung an 12 Forschungszentren
in Deutschland, Frankreich und Großbritannien. Im Jahr 1994 wurde
Dr. Marth Leiter der Stabsabteilung Finanzen und Controlling des Geschäftsbereichs
„Stilllegung nuklearer Anlagen“ am Forschungszentrum Karlsruhe
GmbH. Diese Aufgabe umfasste vier Reaktoren und Kernkraftwerke sowie
um die Wieder aufarbeitungsanlage Karlsruhe, wo er für ein Jahresbudget
von 300 Millionen Euro verantwortlich war.
Dr. Milli Marth war von der friedlichen Nutzung der Kernenergie und ihrem
Beitrag überzeugt. Dies drückt sich nicht nur in seinem beruflichen
Werdegang und seinem stetigen Engagement für dieses Thema aus,
sondern auch darüber hinaus. Im Jahr 2007 entdeckte er das World-Wide-
Web für sich und begleitete bis Ende 2019 unter dem Titel „Rentner blog“ *
mit 456 Posts Themen aus Energietechnik, Energie politik, Technik und
weiteren des öffentlichen Lebens mit spitzer Feder.
Die Entwicklungen, an denen Willy Marth mitgearbeitet und die er
vorangetrieben hat, sind heute international anerkannt und werden fortentwickelt,
auch wenn die Innovationen im eigenen Land teils nicht zum
Tragen kommen. Mit ihm ist ein engagierter Energie- und Kerntechniker
gegangen, seine Ideen und sein Wirken werden Bestand haben.
atw, Redaktion
*Der „Rentnerblog“ ist aktuell noch erreichbar: www.rentnerblog.com
KTG Inside

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Operating Results February 2020
368
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
NEWS
OL1 Olkiluoto BWR FI 910 880 696 644 053 1 334 106 270 799 576 100.00 100.00 99.77 99.89 100.58 100.70
OL2 Olkiluoto BWR FI 910 880 696 645 120 1 334 236 260 698 321 100.00 100.00 100.00 99.96 100.75 100.71
KCB Borssele PWR NL 512 484 696 348 194 728 625 168 710 059 97.45 98.55 97.41 98.53 97.92 99.08
KKB 1 Beznau 7) PWR CH 380 365 696 268 218 555 147 130 863 967 100.00 100.00 100.00 100.00 101.48 101.52
KKB 2 Beznau 7) PWR CH 380 365 696 266 501 551 612 137 848 395 100.00 100.00 100.00 100.00 100.84 100.89
KKG Gösgen 7) PWR CH 1060 1010 696 741 717 1 534 451 323 650 686 100.00 100.00 99.99 99.89 100.54 100.53
CNT-I Trillo PWR ES 1066 1003 696 726 362 1 512 612 257 260 638 100.00 100.00 100.00 100.00 97.42 98.08
Dukovany B1 PWR CZ 500 473 696 346 951 721 350 116 605 533 100.00 100.00 99.58 99.80 99.70 100.19
Dukovany B2 PWR CZ 500 473 696 345 219 716 356 111 759 674 100.00 100.00 99.70 99.75 99.20 99.49
Dukovany B3 2) PWR CZ 500 473 0 0 284 866 110 536 602 0 40.28 0.14 39.64 0 39.56
Dukovany B4 2) PWR CZ 500 473 436 219 348 219 348 110 926 305 62.64 30.28 62.60 30.25 63.03 30.46
Temelin B1 PWR CZ 1080 1030 664 712 309 1 424 618 123 339 431 88.51 88.89 87.58 87.99 94.59 91.43
Temelin B2 PWR CZ 1080 1030 696 815 233 1 630 466 119 113 084 100.00 100.00 100.00 100.00 108.25 104.65
Doel 1 2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0
Doel 2 2) PWR BE 454 433 0 0 0 136 335 470 0 0 0 0 0 0
Doel 3 PWR BE 1056 1006 696 753 758 1 559 721 264 671 371 100.00 100.00 100.00 100.00 102.16 102.08
Doel 4 PWR BE 1084 1033 696 765 543 1 583 350 271 221 625 100.00 100.00 99.98 99.99 99.79 99.86
Tihange 1 2) PWR BE 1009 962 0 0 0 307 547 424 0 0 0 0 0.01 0
Tihange 2 PWR BE 1055 1008 696 726 088 1 507 394 259 561 912 100.00 100.00 99.48 99.74 99.84 100.18
Tihange 3 PWR BE 1089 1038 696 754 690 1 561 434 282 124 011 100.00 100.00 99.99 99.99 100.20 100.19
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 696 838 310 1 785 757 362 506 780 100.00 100.00 94.30 94.22 80.81 83.31
KKE Emsland DWR 1406 1335 696 912 520 1 959 034 359 559 235 100.00 100.00 100.00 100.00 93.05 96.71
KWG Grohnde DWR 1430 1360 696 899 593 1 894 442 390 169 287 100.00 100.00 100.00 99.98 89.72 91.36
KRB C Gundremmingen SWR 1344 1288 696 921 286 1 925 418 343 248 970 100.00 100.00 98.36 99.21 97.67 98.84
KKI-2 Isar DWR 1485 1410 696 983 460 2 085 538 367 848 007 100.00 100.00 100.00 99.99 94.72 97.20
GKN-II Neckarwestheim DWR 1400 1310 696 912 300 1 954 300 342 192 544 100.00 100.00 100.00 100.00 93.60 97.05
Top
IEA recovery plan says
investing in nuclear
will generate jobs and
help secure a sustainable
clean energy future
World Nuclear Association
response to the IEA World
Energy Outlook Special Report
on Sustainable Recovery
(wna) The International Energy
Agency (IEA) has released an energyfocussed
COVID-19 recovery plan
identifying actions that will “move the
world towards a cleaner and more
resilient future.” Investment in existing
nuclear plants, new nuclear build
and supporting innovation in small
modular reactors are among measures
proposed to support a broad range of
clean energy technologies.
Responding to the launch of the
report Agneta Rising, Director General
of World Nuclear Association, said:
“This IEA report confirms that extending
the operations of existing nuclear
plants will support thousands of
jobs and avoid more emissions per
GW than other low-carbon options.
Govern ments stimulus packages
should also accelerate the deployment
of new nuclear build, to bring immediate
employment and economics
benefits through policies aimed at
delivering a clean energy future.”
The report says that extending the
lifetimes of nuclear power plants
would improve electricity security by
lowering the risk of outages, boosting
flexibility, reducing losses and helping
integrate larger shares of variable
renewables such as wind and solar PV.
Additionally, extending the operation
of existing nuclear plants would
reduce fossil fuel imports, improve
electricity security by adding to power
system flexibility, and improve the
affordability of electricity to consumers
The IEA also conclude that modernising
and upgrading existing nuclear
facilities would avoid a steep decline
in low-carbon electricity generation;
new construction would further boost
low-carbon generation.
The report identifies small modular
reactors (SMRs) as offering the
pos sibility of providing low-carbon
nuclear power with lower initial capital
investment and better scal ability with
the potential to provide a large number
News

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Operating Results March 2020
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
369
OL1 Olkiluoto BWR FI 910 880 722 637 308 1 971 414 271 436 884 97.17 99.04 92.55 97.39 93.23 98.16
OL2 Olkiluoto BWR FI 910 880 743 687 549 2 021 785 261 385 870 100.00 100.00 99.90 99.94 100.58 100.67
KCB Borssele PWR NL 512 484 743 379 306 1 107 931 169 089 365 99.45 98.86 99.45 98.84 100.03 99.40
KKB 1 Beznau 7) PWR CH 380 365 743 286 312 841 459 131 150 279 100.00 100.00 100.00 100.00 101.48 101.51
KKB 2 Beznau 7) PWR CH 380 365 743 282 715 834 327 138 131 110 100.00 100.00 99.34 99.78 100.19 100.65
KKG Gösgen 7) PWR CH 1060 1010 743 791 841 2 326 292 324 442 527 100.00 100.00 99.98 99.92 100.54 100.53
CNT-I Trillo PWR ES 1066 1003 743 771 681 2 284 293 258 032 319 100.00 100.00 100.00 100.00 96.93 97.69
Dukovany B1 PWR CZ 500 473 743 372 060 1 093 410 116 977 594 100.00 100.00 100.00 99.87 100.15 100.17
Dukovany B2 PWR CZ 500 473 743 369 383 1 085 739 112 129 057 100.00 100.00 100.00 99.84 99.43 99.47
Dukovany B3 2) PWR CZ 500 473 0 0 284 866 110 536 602 0 26.57 0 26.15 0 26.10
Dukovany B4 PWR CZ 500 473 743 375 002 594 350 111 301 307 100.00 54.01 99.99 53.99 100.94 54.45
Temelin B1 PWR CZ 1080 1030 309 332 124 1 756 742 123 671 555 41.59 72.79 41.29 72.10 41.31 74.38
Temelin B2 PWR CZ 1080 1030 743 810 637 2 441 103 119 923 721 100.00 100.00 100.00 100.00 100.83 103.35
Doel 1 2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0
Doel 2 2) PWR BE 454 433 0 0 0 136 335 470 0 0 0.01 0 0 0
Doel 3 PWR BE 1056 1006 743 803 500 2 363 221 265 474 871 100.00 100.00 100.00 100.00 102.03 102.06
Doel 4 PWR BE 1084 1033 743 815 094 2 398 443 272 036 718 100.00 100.00 99.55 99.84 99.53 99.75
Tihange 1 2) PWR BE 1009 962 0 0 0 307 547 424 0 0 0 0 0 0
Tihange 2 PWR BE 1055 1008 743 780 842 2 288 235 260 342 754 100.00 100.00 99.99 99.83 100.60 100.32
Tihange 3 PWR BE 1089 1038 743 803 085 2 364 519 282 927 095 100.00 100.00 99.89 99.96 99.83 100.07
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 743 971 299 2 757 056 363 478 079 100.00 100.00 94.28 94.24 88.01 84.91
KKE Emsland DWR 1406 1335 743 1 032 994 2 992 028 360 592 229 100.00 100.00 100.00 100.00 98.93 97.47
KWG Grohnde DWR 1430 1360 743 1 011 150 2 905 592 391 180 437 100.00 100.00 100.00 99.99 94.69 92.50
KRB C Gundremmingen 3) SWR 1344 1288 469 623 989 2 549 407 343 872 960 63.12 87.45 62.51 86.72 61.89 86.27
KKI-2 Isar DWR 1485 1410 743 1 057 082 3 142 620 368 905 089 100.00 100.00 100.00 99.99 95.40 96.58
GKN-II Neckarwestheim DWR 1400 1310 743 1 013 700 2 968 000 343 206 244 100.00 100.00 99.88 99.96 97.57 97.23
of jobs in design, manufacturing,
supply and construction activities. The
report recommends that governments
provide investment support, foster
cost-sharing agreements and supporting
regulatory authorities in the
validation of innovative safety features
and factory assembly.
The IEA’s recovery plan also includes
investment in new nuclear
build. However, the report underestimates
the number of new nuclear
power projects ready to start construction,
as well as the thousands of
supply chain jobs that would be
created years before construction
would begin on later reactor projects.
Agneta Rising commented: “For a
sustained transition to a clean energy
future, new nuclear plants must play a
substantial role. With more than
100 new reactors already planned to
be in operation in the 2020s, strong
governmental policy support could
stimulate hundreds of billions of
dollars of investment and tens of
thousands of jobs in the supply chain
long before construction begins.“
In addition to construction, the
operation phase of nuclear power
plants, lasting 60 years or more,
would create a large number of longterm
high-skilled jobs that would
particularly benefit local communities.“This
acce leration of nuclear new
build would support sustainable
economic growth, and would make
a major contribution to the global
nuclear industry’s Harmony goal,
which targets 1000 GWe of new
nuclear capacity by 2050.”
www.iaea.org (201121401)
World
New IAEA reports
on response to the COVID–19
Pandemic
(iaea) As the world grapples with
COVID‐19, the IAEA has adjusted
ways of working to ensure its
operations continue with minimal
disruptions under the extraordinary
circumstances. At the meeting of the
Board of Governors, which is taking
place virtually this week, IAEA
Director General Rafael Mariano
Grossi presented three reports on the
Agency’s COVID‐19 related work. The
reports on support to Member States
in the fight against the pandemic,
support to nuclear and radiation
facility operators and safeguards
implementation during the crisis,
News

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Operating Results April 2020
370
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
NEWS
OL1 Olkiluoto BWR FI 910 880 699 590 640 2 562 054 272 027 524 97.04 98.54 89.17 95.35 89.17 95.93
OL2 Olkiluoto 4) BWR FI 910 880 720 652 413 2 674 197 262 038 283 100.00 100.00 99.23 99.76 98.49 100.13
KCB Borssele PWR NL 512 484 720 366 399 1 474 330 169 455 764 99.47 99.01 99.05 98.89 99.58 99.45
KKB 1 Beznau 1,2,6,7) PWR CH 380 365 395 151 006 992 465 131 301 285 54.86 88.80 54.63 88.75 54.86 89.94
KKB 2 Beznau 6,7) PWR CH 380 365 720 273 585 1 107 912 138 404 695 100.00 100.00 100.00 99.83 100.05 100.50
KKG Gösgen 7) PWR CH 1060 1010 720 761 249 3 087 541 325 203 776 100.00 100.00 99.99 99.94 99.74 100.34
CNT-I Trillo PWR ES 1066 1003 720 661 602 2 945 895 258 693 921 100.00 100.00 100.00 100.00 85.00 94.54
Dukovany B1 PWR CZ 500 473 720 358 018 1 451 428 117 335 611 100.00 100.00 100.00 99.90 99.45 100.00
Dukovany B2 PWR CZ 500 473 720 355 806 1 441 545 112 484 863 100.00 100.00 100.00 99.88 98.84 99.31
Dukovany B3 PWR CZ 500 473 1 29 284 895 110 536 631 0.14 20.01 0.01 19.67 0.01 19.63
Dukovany B4 PWR CZ 500 473 720 360 754 955 104 111 662 061 100.00 65.42 99.65 65.31 100.21 65.80
Temelin B1 PWR CZ 1080 1030 0 0 1 756 742 123 671 555 0 54.74 0 54.22 0 55.93
Temelin B2 PWR CZ 1080 1030 720 782 217 3 223 320 120 705 938 100.00 100.00 100.00 100.00 100.41 102.62
Doel 1 1,2) PWR BE 454 433 0 0 0 137 736 060 0 0 0 0 0 0
Doel 2 1,2) PWR BE 454 433 0 0 0 136 335 470 0 0 0 0 0 0
Doel 3 PWR BE 1056 1006 720 775 809 3 139 030 266 250 680 100.00 100.00 100.00 100.00 101.65 101.96
Doel 4 PWR BE 1084 1033 720 788 060 3 186 503 272 824 778 100.00 100.00 99.26 99.70 99.26 99.63
Tihange 1 2) PWR BE 1009 962 0 0 0 307 547 424 0 0 0 0 0 0
Tihange 2 PWR BE 1055 1008 720 754 589 3 042 824 261 097 343 100.00 100.00 99.98 99.87 100.34 100.33
Tihange 3 PWR BE 1089 1038 720 778 353 3 142 871 283 705 448 100.00 100.00 100.00 99.97 99.91 100.03
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 720 951 530 3 708 586 364 429 609 100.00 100.00 93.80 94.13 89.00 85.93
KKE Emsland 4) DWR 1406 1335 720 1 003 407 3 995 435 361 595 636 100.00 100.00 100.00 100.00 99.17 97.89
KWG Grohnde 2) DWR 1430 1360 266 359 894 3 265 485 391 540 331 36.96 84.36 37.84 84.58 34.73 78.17
KRB C Gundremmingen 3) SWR 1344 1288 590 746 240 3 295 647 344 619 199 81.99 86.09 77.15 84.34 76.49 83.84
KKI-2 Isar DWR 1485 1410 720 972 470 4 115 090 369 877 559 100.00 100.00 99.98 99.99 90.37 95.04
GKN-II Neckarwestheim DWR 1400 1310 720 962 560 3 930 560 344 168 804 100.00 100.00 100.00 99.97 95.54 96.81
*)
Net-based values
(Czech and Swiss
nuclear power
plants gross-based)
1)
Refueling
2)
Inspection
3)
Repair
4)
Stretch-outoperation
5)
Stretch-inoperation
6)
Hereof traction supply
7)
Incl. steam supply
BWR: Boiling
Water Reactor
PWR: Pressurised
Water Reactor
Source: VGB
have also been made available to the
public.
“I said when the crisis began that
there were two areas of the Agency’s
work which would not be halted, no
matter what happened,” said the
Director General in his introductory
statement to the Board of Governors.
“We would continue to implement
safeguards to prevent any misuse of
nuclear material and activities for
non-peaceful purposes. And we would
do everything we possibly could to
assist Member States in confronting
the coronavirus.”
The Report on IAEA Support to
Member State Efforts in Addressing
the COVID-19 Pandemic, describes
the IAEA’s delivery of support to
120 countries and territories that
have requested Agency support to
use the nuclear-related RT-PCR technology
for the detection of COVID-19
infections. The shipments have included
detection equipment, that is,
real time RT‐PCR and kits, together
with reagents and laboratory consumables,
as well as biosafety supplies
such as personal protection equipment
for the safe analysis of samples.
| www.iaea.org (201711216)
Research
Europe: Joint letter urges
Commission to support
hydrogen production from
nuclear
(nucnet) Energy companies, research
institutes and associations have signed
a joint letter to the European Commission
highlighting the possibility of
low-carbon energy sources such as nuclear
to produce clean hydrogen.
One of the signatories, Romania’s
state-controlled nuclear energy producer
Nuclearelectrica, said the
use of low-carbon sources to produce
hydrogen can help achieve
European decarbonisation targets
set for 2050.
A report last year by the International
Energy Agency said “now is the
time” to scale up technologies and
bring down costs to allow hydrogen to
become widely used.
Hydrogen is created using steam
methane reforming, which basically
uses high temperatures to convert
steam and methane into hydrogen gas
and carbon dioxide.
News

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Supplying hydrogen to industrial
users is a major business. But its
production is responsible for CO 2
emissions of around 830 million
tonnes of carbon dioxide per year,
equivalent to the CO 2 emissions of the
UK and Indonesia combined.
A wide variety of fuels are able to
produce hydrogen, including nuclear,
renewables, natural gas, coal and oil.
It can be transported as a gas by
pipelines or in liquid form by ships,
much like liquefied natural gas (LNG).
It can be transformed into electricity
and methane to power homes and
feed industry, and into fuels for cars,
trucks, ships and planes.
The IEA said nuclear power plants
are an option for the provision of heat
for hydrogen production. Depending
on local conditions, using steam from
nuclear power could be cheaper
than using steam from natural gas, as
well as reducing the carbon intensity
of the hydrogen produced. It could
also provide a useful additional
revenue stream for nuclear power
plants.
The joint letter, addressed to senior
European Commission officials,
recom mends that proposals for clean
hydrogen generation be considered in
the strategy documents which due to
be published by the Commission as
part of the European Green Deal.
The proposals consist of encouraging
the production of, and the
demand for, hydrogen from clean
sources, which will allow the replacement
of hydrogen supplied from
sources with significant CO 2 emissions.
“Hydrogen production through the
use of nuclear energy, greenhouse
gas-free energy, is considered by the
major producing and consuming
companies in Europe as a potential
way to recover from the economic
crisis generated by the Covid-19
pandemic,” the letter says.
It calls for the Commission to begin
a debate on the essential role of
nuclear energy in decarbonising the
energy and industrial sector. This
would include producing hydrogen
through existing nuclear capacity in
Europe and through new capacity that
will be built in Romania and Europe.
In the US, the Department of
Energy is looking at ways to develop
new technologies to efficiently scale
up the production of hydrogen using
all of the nation’s energy sources,
including nuclear. The DOE said a
single 1,000-MW nuclear reactor
could produce more than 200,000
tonnes of hydrogen each year.
IAEA launches initiative
to help prevent future
pandemics
(iaea) The Director General of the
International Atomic Energy Agency
(IAEA), Rafael Mariano Grossi,
launched an initiative today to
strengthen global preparedness for
future pandemics like COVID-19.
The project, called ZODIAC, builds
on the IAEA’s experience in assisting
countries in the use of nuclear and
nuclear-derived techniques for the
rapid detection of pathogens that
cause transboundary animal diseases,
including ones that spread to humans.
These zoonotic diseases kill around
2.7 million people every year.
The IAEA Zoonotic Disease Integrated
Action (ZODIAC) project will
establish a global network to help
national laboratories in monitoring,
surveillance, early detection and
control of animal and zoonotic
diseases such as COVID-19, Ebola,
avian influenza and Zika. ZODIAC is
based on the technical, scientific and
laboratory capacity of the IAEA
and its partners and the Agency’s
mechanisms to quickly deliver equipment
and know-how to countries.
The aim is to make the world
better prepared for future outbreaks.
“ Member States will have access to
equipment, technology packages,
expertise, guidance and training.
Decision-makers will receive up-todate,
user-friendly information that
will enable them to act quickly,” Mr
Grossi told a meeting of the IAEA
Board of Governors.
Mr Grossi said COVID-19 had
exposed problems related to virus
detection capabilities in many
countries, as well as a need for
better communication between health
institutions around the world. While
the IAEA has been doing important
work to help countries in these areas,
such as through the provision of
COVID-19 tests, he said it was “essential
to pull these diverse strands
together into a coherent and comprehensive
framework of assistance”.
Nuclear-derived techniques, such
as tests using real time reverse
transcription-polymerase chain reaction
(RT-PCR), are important tools in
the detection and characterization of
viruses. The IAEA is providing emergency
assistance to some 120 countries
in the use of such tests to rapidly
detect COVID-19.
Zoonotic diseases are caused by
bacteria, parasites, fungi or viruses
that originate in animals and can be
transmitted to humans. Many of these
| Based on the IAEA's technical, scientific and laboratory capacity, ZODIAC
will establish a global network to help national laboratories in monitoring,
surveillance, early detection and control of animal and zoonotic diseases
such as COVID-19 and Ebola. (Photo: D. Calma/IAEA)
diseases are treatable if medication is
available, such as E. coli- and brucella
bacterial infections. But others
have the potential to severely affect
humans, such as Ebola, SARS and
COVID-19.
ZODIAC builds on the experience
of VETLAB, a network of veterinary
laboratories in Africa and Asia that
was originally set up by the Food and
Agriculture Organization of the
United Nations (FAO) and the IAEA to
combat the cattle disease rinderpest.
VETLAB now supports countries in
the early detection of several zoonotic
and animal diseases, such as African
swine fever and pest des petit
ruminants (PPR).
“About 70 per cent of all diseases in
humans come from animals,” said
Gerrit Viljoen, Head of the Animal
Production and Health Section of the
Joint FAO/IAEA Programme for
Nuclear Techniques in Food and
Agriculture.
ZODIAC aims to help veterinary
and public health officials identify
these diseases before they spread.
“We have seen an increase in the
number of zoonotic epidemics in the
last decades: first Ebola, then Zika,
and now COVID-19. It’s important to
monitor what is in the animal kingdom
– both wildlife and livestock – and to
act quickly on those findings before
the pathogens jump to humans,” Mr
Viljoen said.
Following the One Health concept
for a multidisciplinary collaborative
approach between human and animal
health authorities and specialists,
ZODIAC will benefit from the unique
joint FAO/IAEA laboratories and from
partners such as the World Health
Organization (WHO) and the World
Organisation for Animal Health (OIE).
“We have a unique capacity to
provide laboratory support and
guidance to countries,” said Mr
Viljoen, adding that ZODIAC will, for
example, provide technical know-how
371
NEWS
News

atw Vol. 65 (2020) | Issue 6/7 ı June/July
372
NEWS
and advice to laboratories on test
performance and assist authorities in
the interpretation of results and in
devising containment measures.
ZODIAC will also support R&D
activities for novel technologies and
methodologies for early detection and
surveillance. Under the project, the
IAEA will enhance its capacities to
host scientists and fellows from
Member States at its Seibersdorf
laboratories outside Vienna and to
carry out research on immunological,
molecular, nuclear and isotopic tests,
as well as in the use of irradiation to
develop vaccines against diseases
such as avian influenza.
| www.iaea.org (201741136)
Persons
FRM II: Axel Pichlmaier
new Technical Director
(frm) On 1 July 2020 Dr. Axel
Pichlmaier takes up the post of Technical
Director of the Heinz Maier-
Leibnitz research neutron source. The
51-year-old physicist brings with him
experience from neutron research as
well as from reactor operation and
nuclear supervision.
The initial spark for Axel
Pichlmaier's career was given by
Professor Dr. Klaus Schreckenbach in
1992 in the lecture “Nuclear Solid
State Physics” at the Technical University
of Munich (TUM). Schreckenbach
found a working student in Axel
Pichlmaier among the listening
students in the field of basic research
with neutrons at the Institut Laue
Langevin (ILL) in Grenoble, France.
After his diploma thesis at the prototype
positron source at the Atomic
Egg, Pichlmaier finally did his PhD on
ultracold neutrons within the framework
of a collaboration between the
TUM and the ILL, also under Professor
Schreckenbach.
Now Axel Pichlmaier follows in his
doctoral supervisor’s the footsteps,
who was the first Technical Director of
the FRM II from 1999 to 2005.
Big projects are waiting
According to Pichlmaier, the FRM II
had been handed over by his predecessor
Dr. Anton Kastenmüller “in
perfect condition”. Kastenmüller's second
term of office as Technical Director
of FRM II ended in March 2020 after
five years according to rotation. “In ten
years, Dr. Anton Kastenmüller has rendered
out standing services to the safe
operation of the FRM II,” said the
Scientific Director of the FRM II and
MLZ, Prof. Dr. Peter Müller -Buschbaum.
He also expressly thanked
Dr. Heiko Gerstenberg and Roland
Schätzlein, who had provisionally
taken over the Technical Management
from April until Pichlmaier took
office. In addition to safe operation,
Pichlmaier and his more than 110 employees
are also responsible for the
supply of fresh fuel, the disposal of
spent fuel elements and fuel conversion:
major projects with far-reaching
significance for FRM II.
| www.frmii.tum.de
Company News
Framatome and the
Technical University of Munich
to develop new fuel
for research reactor
(framatome) Framatome and the Technical
University of Munich (TUM)
recently began the commercial development
of uranium- molybdenum fuel
(UMo) for nuclear research reactors.
Framatome and TUM will design
and install the fuel manufacturing
production line, and develop, produce
and irradiate new fuel prototypes. The
project will take place at the CERCA
Research and Innovation Lab (CRIL),
Framatome’s new research and
development laboratory dedicated to
advancing the fabrication of nuclear
fuels for medical, research and
sterilization applications.
“By producing this new fuel for
TUM, Framatome allows research
reactors to maintain performance
while using low-enriched uranium,”
said François Gauché, director of the
CERCA Business Line at Framatome.
“We look forward to advancing this
fuel technology and developing a new
fuel option for research reactors.”
The team will install the operational
manufacturing line in early
2021 with the production of the first
prototypes planned for 2022. TUM,
Framatome, the French Alternative
Energies and Atomic Energy Commission,
the Institut Laue-Langevin
and SCK-CEN (the Belgian nuclear
research center) will be involved in
irradiation activities.
“TUM and Framatome’s collaboration
on the development of this
new fuel guarantees a reliable and
efficient source of neutrons for
research, industry and medicine,” said
Professor Peter Müller-Buschbaum,
scientific director of TUM's FRM II
nuclear research reactor. “This fuel is
an essential tool for the development
of science in Germany.”
The development of this fuel
is a major challenge, which several
international teams are tackling. The
success of this project, in addition to
providing significant support to one
of its main customers, will allow
Framatome to position itself well in
the manufacture of UMo research
fuel.
| www.framatome.com (201711204)
ROSATOM starts life tests
of third-generation VVER-440
nuclear fuel
(rosatom) OKB Gidropress research
and experiment facility, an enterprise
of ROSATOM machinery division
Atomenergomash, has started life
tests of a mock-up of the third-generation
nuclear fuel RK3+ for VVER-440
reactors. The work is carried out
within the contract between TVEL
Fuel Company of ROSATOM and
Czech power company ČEZ a.s., which
includes design and introduction of
this fuel modification at Dukovany
NPP in the Czech Republic.
The difference between RK3+ and
the previous fuel generations for
VVER-440 is the improved structure,
which enabled to advance mechanical
and thermal-hydraulic performance
of the fuel. «Introduction of RK3+ will
make it possible to operate all four
power units at increased thermal
capacity and also to extend the fuel
cycle at Dukovany NPP, which will
improve economic efficiency of the
power plant operation», said
Alexander Ugryumov, Vice President
for Research and Development at
TVEL JSC.
The life tests started after successful
completion of hydraulic tests (hydraulic
filling) of the mock-up with
the aim to determine RK3+ hydraulic
resistance. Life tests are carried out
on a full-scale research hot run-in test
bench V-440 and will last for full
1500 hours.
The aim of tests is to study mechanical
stability of RK3+ components
under thermal-hydraulic and dynamic
conditions, which are close as possible
to full-scale operation.
| www.rosatom.ru ( (201711200)
Westinghouse installs
first-of-a-kind 3d-printed fuel
component inside commercial
nuclear reactor
(west-n) Westinghouse Electric
Company announced today a 3Dprinted
thimble plugging device was
successfully installed in Exelon’s
News

atw Vol. 65 (2020) | Issue 6/7 ı June/July
Uranium
Prize range: Spot market [USD*/lb(US) U 3O 8]
140.00
) 1
Uranium prize range: Spot market [USD*/lb(US) U 3O 8]
140.00
) 1
120.00
120.00
373
100.00
100.00
80.00
80.00
60.00
40.00
20.00
Yearly average prices in real USD, base: US prices (1982 to1984) *
60.00
40.00
20.00
NEWS
0.00
1980
1985
1990
1995
2000
2005
2010
2015
2020
Year
* Actual nominal USD prices, not real prices referring to a base year. Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020 * Actual nominal USD prices, not real prices referring to a base year. Year
Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020
| Uranium spot market prices from 1980 to 2020 and from 2009 to 2020. 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
26.00
) 1 ) 1
160.00
140.00
0.00
24.00
22.00
20.00
Jan. 2009
Jan. 2010
Jan. 2011
Jan. 2012
Jan. 2013
Jan. 2014
Jan. 2015
Jan. 2016
Jan. 2017
Jan. 2018
Jan. 2019
Jan. 2020
Jan. 2021
120.00
18.00
16.00
100.00
14.00
80.00
12.00
10.00
60.00
8.00
40.00
6.00
20.00
4.00
2.00
0.00
0.00
Jan. 2009
Jan. 2010
Jan. 2011
Jan. 2012
Jan. 2013
Jan. 2014
* Actual nominal USD prices, not real prices referring to a base year. Year
Jan. 2015
Jan. 2016
Jan. 2017
Jan. 2018
Jan. 2019
Jan. 2020
Jan. 2021
Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020
Jan. 2009
Jan. 2010
Jan. 2011
Jan. 2012
Jan. 2013
Jan. 2014
* Actual nominal USD prices, not real prices referring to a base year. Year
Jan. 2015
Jan. 2016
Jan. 2017
Jan. 2018
Jan. 2019
Jan. 2020
Jan. 2021
Sources: Energy Intelligence, Nukem; Bild/Figure: atw 2020
| Separative work and conversion market price ranges from 2009 to 2020. 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 2020
Byron Unit 1 nuclear plant during
their spring refueling outage. It is a
first-of-a-kind installation for the
nuclear industry.
Additive manufacturing, also
known as 3D printing, is an innovative
technique that simplifies the manufacturing
process by going directly
from 3D model to an actual part.
Additive manufacturing reduces cost,
improves quality and design flexibility,
and eliminates conventional
manufacturing limitations.
“Westinghouse continues to lead
the way with development of the most
advanced technologies to help the
world meet growing electricity
demand with safe, clean and reliable
energy,” said Ken Canavan, Westinghouse’s
chief technology officer. “Our
additive manufacturing program
offers customers enhanced component
designs that help increase
performance and reduce costs,
as well as provide access to components
that may not be available
using traditional manufacturing
methods.”
“Additive manufacturing is an
exciting new solution for the nuclear
industry,” said Ken Petersen, Exelon
Generation’s vice president of nuclear
fuels. “The simplified approach helps
meet the industry's need for a wide
variety of low-volume, highly critical
plant components. We are proud to
have Westinghouse as a partner on
this industry milestone and to help
further demonstrate the viability of
this technology.”
| www.westinghousenuclear.com
(201711207)
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 to June 2019
p Uranium: 23.90–29.10
p Conversion: 13.50–18.00
p Separative work: 41.00–49.00
July to December 2019
p Uranium: 24.50–26.25
p Conversion: 18.00–23.00
p Separative work: 47.00–52.00
2020
January 2020
p Uranium: 24.10–24.90
p Conversion: 22.00–23.00
p Separative work: 48.00–51.00
February 2020
p Uranium: 24.25–25.00
p Conversion: 22.00–23.00
p Separative work: 45.00–53.00
March 2020
p Uranium: 23.05–27.40
p Conversion: 21.50–23.50
p Separative work: 45.00–52.00
April 2020
p Uranium: 27.50–34.00
p Conversion: 21.50–23.50
p Separative work: 45.00–52.00
| Source: Energy Intelligence
www.energyintel.com
News

atw Vol. 65 (2020) | Issue 6/7 ı June/July
374
‘Green Energy’ Plans Will never Ripen
without Nuclear in the Mix
NUCLEAR TODAY
John Shepherd is a
freelance journalist
and communications
consultant.
Sources:
University
of East Anglia study
https://bit.ly/30l3YKA
Maroš Šefčovič on EU
recovery plan
https://bit.ly/2XHKZrJ
Sizewell C
announcement
https://bit.ly/3h85Pbg
We have had precious little to cheer about this year amidst the impact on all our lives wrought by the coronavirus.
As I write, there continues to be an onslaught of grim statistics from parts of the world still struggling to cope with the
deadly devastation, while much of Europe moves tentatively out of a patchwork of individual ‘lockdown’ regimes.
Only one statistic could be remotely considered as welcome
(a term that has to be seen against the truly awful toll of
deaths)… the impact the lockdowns have had around the
world on the global environment.
According to a study published in the journal ‘Natural
Climate Change’, daily emissions decreased by 17% – or
17m tonnes of carbon dioxide – globally during the peak of
the confinement measures in early April, compared to
mean daily levels in 2019 and dropping to levels last
observed in 2006.
The study, led by the UK’s University of East Anglia, said
emissions from surface transport, such as car journeys,
accounted for almost half (43%) of the decrease in global
emissions during peak confinement on 7 April. “Emissions
from industry and from power together accounted for a
further 43% of the decrease in daily global emissions,” the
study said.
Professor Rob Jackson of Stanford University and chair
of the Global Carbon Project, who co-authored the study,
said the “substantial” drop in emissions highlighted the
need for “systemic change through green energy and
electric cars, not temporary reductions from enforced
behaviour”.
Having limited the belching out of fossil fuels as a result
of the pandemic, our slow sojourn to what is being
described as the ‘new normal’ offers breathing space for
policymakers to reflect on the impact of industrial policies.
However, it seems the opportunity to reboot Europe’s
industrial economy will be done without any encouragement
whatsoever for clean, green, nuclear energy.
The latest announcement from European Union policymakers
– contained in its green recovery plan published
towards the end of May – makes clear that the ‘new normal’
for the energy sector will be a return to business as usual. A
continuation of the political blind spot – at least as far as
the environmental benefits of nuclear power are concerned.
EU vice-president Maroš Šefčovič said the recovery plan
reflected “the new reality and shows we will focus all our
actions on overcoming the crisis, jumpstarting our
eco nomy and putting the European Union firmly on a
resilient, sustainable and fair recovery path”.
Šefčovič is in charge of inter-institutional relations and
foresight. Sadly, there seems very little foresight in the
bloc’s plan as far as a “fair recovery” is concerned. The
vice-president has a track record on not allowing fair play
in the energy industry.
As a former EU energy chief, Šefčovič consistently
supported policies that favoured pouring millions of euros
into developing the lithium-ion battery sector, despite the
fact it struggles with immense environmental issues
including a poor recycling record. On his watch, it is fair to
say Europe’s lead-acid battery industry, with a near 100 %
environmentally-efficient recycling record, enjoyed next to
no recognition or support.
Now nuclear is set to continue as the EU’s bête noire.
According to Foratom, the Brussels-based trade association
for the nuclear energy industry in Europe, the EU’s
green recovery plan “ignores the need for clean, dispatchable
and European sources of energy”.
Foratom has called on EU leaders to focus on solutions
that will help Europe emerge from the pandemic and
fully commit to the EU’s earlier pledge to invest in clean
technologies that creates growth and jobs.
Nuclear is “a European technology, with a European
supply chain, capable of providing Europe with the
low-carbon energy it needs, when its needs it”, Foratom
said. In addition, it is right to stress that nuclear also plays
a key role in medical diagnosis and treatment.
Foratom’s director-general Yves Desbazeille said the
European Commission “has once again ignored Europe’s
largest source of low-carbon dispatchable energy”.
But there are glimmers of hope for sensible planning
and investment in new nuclear. The UK arm of France’s
EDF recently confirmed it had taken the first key steps
toward building the Sizewell C nuclear plant on England’s
eastern Suffolk coast. EDF Energy has applied to the
Planning Inspectorate for a development consent order for
the twin-unit ‘UK EPR’ site.
The application moves forward after four rounds of
public consultation that began eight years ago. EDF said the
“extraordinary circumstances” created by the pandemic had
forced it to delay the application by a further two months –
but Sizewell C now represents a major step towards picking
up the pace for the UK’s nuclear programme.
Sizewell C will be a near replica of the Hinkley Point C
plant that EDF Energy is building in western England.
Even in its nascent stages, Sizewell C is projected to
provide a huge stimulus towards the recovery of the
domestic economy in the wake of the pandemic. According
to EDF, around 25,000 employment opportunities and
1,000 apprenticeships will be created during construction.
In addition to the economic benefits, Sizewell C will
avoid an estimated 9 million tonnes of CO 2 being pumped
into the atmosphere each year, compared to producing the
3.2GW of electricity from gas-powered generation, EDF
said.
These are the statistics that are part of the nuclear story
that cannot be repeated enough. Before first cement is even
poured, nuclear construction projects create thousands of
jobs, support livelihoods, strengthen local businesses and
frequently boost international supply chains. Once operational,
the plants can run safely for decades to come while
contributing to vibrant, clean-energy economies.
Reflecting on this plethora of positive data, what is it that
makes nuclear so objectionable for EU energy planners?
The nuclear industry is not asking for preferential treatment,
only what our antipodean cousins would describe as
“a fair go”.
The facts are clear but, as the saying goes, there are
none so blind as those who will not see.
Author
John Shepherd
Nuclear Today
‘Green Energy’ Plans Will never Ripen without Nuclear in the Mix ı John Shepherd

Kommunikation und
Training für Kerntechnik
Rückbau kerntechnischer Anlagen
In Kooperation mit
TÜV SÜD Energietechnik GmbH
Baden-Württemberg
Seminar:
Stilllegung und Rückbau in Recht und Praxis
Seminarinhalte
Genehmigungen für die Stilllegung und den Rückbau
ı
ı
Gestaltung der Übergänge zwischen der Betriebs- und der Stilllegungsgenehmigung
Gestaltung der Übergänge zwischen Genehmigungs- und Aufsichtsphase
Rechtliche Gestaltung des Genehmigungsverfahrens
ı
ı
ı
ı
Elemente des Umweltbereiches
UVP und Erörterungstermin
Anfechtung von Genehmigungen
Stellungnahmen der EU-Kommission
Reststoffe und Abfälle
ı
ı
ı
ı
Vorstellung von Reststoffkonzepten
Bewertung der Reststoffkonzepte
Neue Regelungen zum Übergang der Entsorgungsverantwortung
Entsorgungsfragen rund um Abfälle aus dem Rückbau
Zielgruppe
Die 2-tägige Schulung wendet sich an Fach- und Führungskräfte, Mitarbeiterinnen und Mitarbeiter
von Betreibern, Industrie und Dienstleistern, die sich mit der Thematik aktuell bereits beschäftigen
oder sich künftig damit auseinander setzen werden.
Maximale Teilnehmerzahl: 12 Personen
Referenten
Dr. Stefan Kirsch
Dr. Christian Raetzke
Wir freuen uns auf Ihre Teilnahme!
ı Abteilung Stilllegung, Entsorgung, Reaktorphysik, TÜV SÜD Energietechnik
GmbH Baden-Württemberg
ı Rechtsanwalt, Leipzig
Bei Fragen zur Anmeldung rufen Sie uns bitte an oder senden uns eine E-Mail.
Termin
2 Tage
23. bis 24. September 2020
Tag 1: 11:00 bis 18:00 Uhr
Tag 2: 09:00 bis 16:30 Uhr
Veranstaltungsort
Kaiserin Friedrich-Haus
Robert-Koch-Platz 7
10115 Berlin
Teilnahmegebühr
1.558,– € ı zzgl. 16 % USt.
Im Preis inbegriffen sind:
ı Seminarunterlagen
ı Teilnahmebescheinigung
ı Pausenverpflegung
inkl. Mittagessen
Kontakt
INFORUM
Verlags- und Verwaltungsgesellschaft
mbH
Robert-Koch-Platz 4
10115 Berlin
Petra Dinter-Tumtzak
Fon +49 30 498555-30
Fax +49 30 498555-18
seminare@KernD.de

DER NUKLEARE TRAUM
DEUTSCHE ATOMGESCHICHTE
ALS 420-SEITIGER BILDBAND
SIEBEN LÄNDER, 60 ORTE, 300 BILDER
Von Otto Hahn bis zu FRM II und EPR,
von Gronau über La Hague bis Onkalo,
vom Bau in Angra bis zur Grünen Wiese.
Mit Gebäude- und Reaktorplänen.
Bestellen Sie bei Interesse gern über
die Buch-Webseite (s.u.). So unterstützen Sie
das Projekt und erhalten als Dank ein Plakat dazu.
www.dernuklearetraum.de