atw - International Journal for Nuclear Power | 06/07.2020

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Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information.

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2020

6/7

ISSN · 1431-5254

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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?


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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


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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

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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.

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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

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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.

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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

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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.

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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.

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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.

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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).

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| 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.

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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

atwInternational Journal

for Nuclear Power

nicole.koch@nucmag.com

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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

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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

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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.

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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,

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| 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

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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.

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12469.pdf

Author

James Conca

UFA Ventures, Inc.

jim@ufaventures.com

2801 Appaloosa Way

Richland, WA 99352, USA

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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

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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

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ı Steffen Kanitz


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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

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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

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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

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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

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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

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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

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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


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.)


2

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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

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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

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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

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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

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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

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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

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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).

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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

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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

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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

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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

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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


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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.

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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

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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


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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

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Experimental Investigations into Flow Conditions of Konrad Exhaust Air Channel ı Steffen Wildgrube, Anton Anthofer, Michael Haas, Alexander Kratzsch and Clemens Schneider


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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

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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

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Experimental Investigations into Flow Conditions of Konrad Exhaust Air Channel ı Steffen Wildgrube, Anton Anthofer, Michael Haas, Alexander Kratzsch and Clemens Schneider


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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