atw - International Journal for Nuclear Power | 11/12.2019

nucmag.com

2019

11/12

Performance Shaping

Factors for Human Error

Reduction

Decommissioning & Dismantling

of the Rossendorf Research

Reactor RFR

First On-site Demonstration

of Laser- based Decontamination

Technology in Germany

ISSN · 1431-5254

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atw Vol. 64 (2019) | Issue 11/12 ı November/December

TMI and Lessons Learned – Afterwards

and for the Future

No technical development is perfect from the outset, each involves specific risks. A responsible approach to

technology therefore requires a responsible and forward-looking approach to its risks, the best possible protection and,

if necessary, further development. This also means that there are no such things as absolutely perfect technology and

absolute safety.

With a view to criticism of technologies, technology sceptics

are provided with a comfortable approach for transporting

their respective concerns by anniversaries. One such event

is the 1979 Three-Mile-Island (TMI 2) accident.

On 28 March 1979, shortly after 4 a.m., the feed water

supply for the steam generators in unit 2 of the Three Mile

Island (TMI 2) nuclear power plant near Harrisburg,

Pennsylvania, USA, failed. Such a malfunction was taken

into account in the operating procedures with a view to

the future and can be easily dealt with by the plant and

operators. The event is one of the design cases. But what

began as a normal incident later developed into the most

serious accident to date in a Western nuclear power plant

due to a combination of technical defects, human error

and organisational inadequacies.

The technical aspects of the TMI-2 accident have been

extensively reviewed and can be traced in detail in the

literature: After the failure of the feed-water supply, the

safety systems of the nuclear power plant reacted as

designed. However, due to valves in the emergency feedwater

system that were closed manually on site, the steam

generators were not supplied with water and a link in the

chain for the residual heat dissipation of the reactor, which

was switched off but still supplies heat due to the decay

heat, was missing. This error was corrected 8 minutes

after the start of the fault. Subsequently, the increase in

pressure in the primary circuit caused the safety relief

valve on the pressure holder to open, the escape of steam

from the primary circuit and the intended pressure

reduction. However, the opened blow-off valve did not

close as designed. This caused large quantities of coolant

to escape from the primary circuit – a loss of coolant

accident “ LOCA” had occurred. Only 2 hours later it was

detected by the operators and corrected by closing other

existing shut-off valves. After further misinterpretations

about the condition of the reactor core and the primary

circuit, the core was no longer cooled by water and the

exothermic reaction between water vapour and the

cladding tube material of the fuel elements then set on,

releasing additional heat. About half of the fuel rod mass

in the core melted or was severely damaged. Core masses

shifted and partially reached the bottom of the reactor

pressure vessel where they solidified. After it had been

recognised that a core melt was in full swing, measures for

core cooling were taken and it was possible to ensure the

heat removal of the TMI-2 core again in the long term.

Of the radioactive substances released in the plant, a

small part was released into the environment. Triggered

by fears of the additional radiation exposure associated

with the release of a calculated average of 0.02 mSv – i.e.

about 1 % of the natural annual radiation exposure – the

Pennsylvania Department of Health kept a register for

18 years containing the data of more than 30,000 persons

who had lived within a five-mile radius of the nuclear

power plant at the time of the accident. This state register

was closed in June 1997 after no unusual health

developments had been identified. More than a dozen

comprehensive studies on the physical health effects of the

accident did not provide evidence of an abnormally high

number of cancer cases in the region around the nuclear

power plant years after the event.

Despite all the shortcomings and shortcomings that

led to the TMI-2 accident, the basic safety concept of

Western nuclear power plants that had been established at

an early stage was confirmed: The amount of radioactive

substances released from the plant was low and there were

no fatalities or injuries. On the other hand, the accident

sequence underlines the suitability of successively

staggered safety barriers, which, taking even serious

accidents into account, aims at the confinement of the

radioactive substances in the plant itself: The reactor

building of TMI-2 withstood, the containment remained

intact.

In addition, research and development for reactor

safety further minimised the risks in an international

context during the past 40 years: Expressed in figures,

even for running plants in Germany and elsewhere, a

safety level was achieved, which is 100 times higher than

the original international reference level. The reactor

disaster at Fukushima in 2011 will not change this point

of view either, as other foreseeable risks and specific

boundary conditions were the causes.

Safety is and remains an international task for all

concerned. Nuclear safety and the maintenance and

promotion of competences, especially when they are

available with excellent know-how in research and

development and are internationally recognised – as in

Germany – are part of the overall social and political

responsibility at the same level as other goals of

environmental protection. Today's and future safety and

the promotion of safety culture and technology, also for

nuclear technology, cannot and must not be the subject

of restricted or restrictive action. Research and

development should always be allowed to advance in

the spirit of freedom of science and technology.

Christopher Weßelmann

– Editor in Chief –

507

EDITORIAL

Editorial

TMI and Lessons Learned – Afterwards and for the Future


508

EDITORIAL

TMI und Lessons Learned – danach

und für die Zukunft

Keine technische Entwicklung ist von vornherein perfekt, jede birgt spezifische Risiken. Ein verantwortungsvoller

Umgang mit Technik erfordert daher einen verantwortungsvollen und vorausblickenden Umgang mit ihren

Risiken, bestmöglichen Schutz und ggf. auch spätere Weiterentwicklung. Dies bedeutet jedoch auch, dass es eine

absolut perfekte Technik nicht gibt, ebenso keine absolute Sicherheit.

Mit Blick auf Kritik an Technologien liefern Jahrestage gerne

Technologieskeptikern einen komfortablen Ansatz, ihre

jeweiligen Anliegen zu transportieren. Ein solches Ereignis

ist auch der Three-Mile-Island (TMI 2)-Unfall von 1979.

Am 28. März 1979, kurz nach 4 Uhr morgens, fiel im

Block 2 des US-amerikanischen Kernkraftwerks Three Mile

Island (TMI 2) nahe Harrisburg, Pennsylvania, die

Speisewasser versorgung der Dampferzeuger aus. Eine

solche Störung ist in den Betriebsabläufen vorausblickend

berücksichtigt und kann von Anlage und Operateuren ohne

Weiteres bewältigt werden. Das Ereignis gehört zu den

Auslegungsfällen. Doch was als normale Störung begann,

entwickelte sich im weiteren Verlauf aufgrund einer Kombination

aus technischen Mängeln, mensch lichen Fehlern und

organisatorischen Unzulänglichkeiten zum bis dahin schwerwiegendsten

Unfall in einem west lichen Kernkraftwerk.

Die technischen Aspekte zum TMI-2-Unfall sind ausgiebig

aufgearbeitet und können in der Literatur detailliert

nachvollzogen werden: Nach dem Ausfall der Speisewasserversorgung

reagierten die Sicherheitssysteme des Kernkraftwerks

auslegungsgemäß. Aufgrund versehentlich

manuell Vor-Ort geschlossener Ventile im Notspeisewassersystem

fehlte aber die Bespeisung der Dampferzeuger und

damit ein Glied in der Kette für die Nachwärmeabfuhr des

zwar abgeschalteten aber durch die Nachzerfalls wärme

immer noch Wärme liefernden Reaktors. Dieser Fehler

wurde 8 Minuten nach Störungsbeginn korrigiert. In

weiterer Folge kam es durch den Druckanstieg im Primärkreislauf

zum Öffnen des für diesen Fall vorgesehenen

Sicherheits-Abblaseventils am Druckhalter, dem Entweichen

von Dampf aus dem Primärkreislauf und der

beabsichtigten Druckabsenkung. Das geöffnete Abblaseventil

schloss aber folgend nicht auslegungsgemäß. Über

diesen Weg entwichen große Mengen Kühlmittel aus dem

Primärkreislauf – ein Loss of coolant accident „LOCA“ war

eingetreten. Erst 2 Stunden später wurde er durch die

Operateure erkannt und durch Schließen weiterer vorhandener

Absperrarmaturen korrigiert. Nach weiteren

Fehlinterpretationen über den Zustand von Reaktorkern

und Primärkreislauf war der Kern nicht mehr durch Wasser

gekühlt und die dann einsetzende exotherme Reaktion

zwischen Wasserdampf und dem Hüllrohrmaterial der

Brenn elemente setzte zusätzliche Wärme frei. Etwa die

Hälfte der Brennstabmasse im Kern schmolz bzw. wurde

schwer beschädigt. Kernmassen verlagerten sich und

erreichten teilweise den Boden des Reaktordruckbehälters,

wo sie erstarrten. Nachdem erkannt worden war, dass eine

Kernschmelze in vollem Gange war, wurden Maßnahmen

zur Kernkühlung ergriffen und es gelang, die Wärmeabfuhr

des TMI-2-Kerns wieder langfristig zu gewährleisten.

Von den in der Anlage frei gesetzten radioaktiven

Stoffen gelangte ein geringer Teil in die Umgebung.

Ausgelöst durch die Ängste der mit der Freisetzung

verbundenen zusätzlichen Strahlenbelastung von berechneten

im Mittel 0,02 mSv – also etwa 1 % der natürlichen

jährlichen Strahlenbelastung – führte das Pennsylvania

Department of Health während 18 Jahren ein Register mit

den Daten von mehr als 30.000 Personen, die zum Zeitpunkt

des Unfalls im Umkreis von fünf Meilen um das

Kernkraftwerk gelebt hatten. Dieses staatliche Register

wurde im Juni 1997 geschlossen, nachdem keine ungewöhnlichen

Entwicklungen bei der Gesundheit festgestellt

worden waren. Mehr als ein Dutzend Studien über die

Auswirkungen des Unfalls auf die physische Gesundheit

gaben auch Jahre nach dem Ereignis keine Hinweise auf

eine abnormal hohe Zahl von Krebsfällen in der Region um

das Kernkraftwerk.

Trotz aller Unzulänglichkeiten und Fehler die zum TMI-

2-Unfall führten, bestätigte sich einerseits das grundlegende

frühzeitig etablierte Sicherheitskonzept westlicher

Kernkraftwerke: Die aus der Anlage freigesetzte Menge an

radioaktiven Stoffen war gering und weder Todesopfer

noch Verletzte waren zu beklagen. Und andererseits

unterstreicht der Unfallablauf den Sinn hintereinander

gestaffelter Sicherheitsbarrieren, die unter Berücksichtigung

auch schwerer Unfälle, den Einschluss der radioaktiven

Stoffe in der Anlage selbst zum Ziel haben: Das

Reaktorgebäude von TMI-2 hielt stand, der Sicherheitsbehälter

blieb intakt.

Zudem haben Forschung und Entwicklung für die

Reaktor sicherheit in den letzten 40 Jahren die Risiken

im internationalen Kontext weiter minimiert: In Zahlen

ausgedrückt wurde selbst für laufende Anlagen in

Deutschland und anderswo ein Sicherheitsniveau erreicht,

das um den Faktor 100 höher liegt, als das ursprüngliche

inter nationale Referenz niveau. Daran ändert auch die

Reaktorhavarie von Fukushima 2011 nichts, da andere

absehbare Ursachen und spezifische Rand bedingungen

Auslöser waren.

Sicherheit ist und bleibt dabei eine international zu

lebende Aufgabe für alle Beteiligten. Kerntechnische

Sicherheit und der Erhalt und die Förderung von

Kompetenzen, vor allem dann, wenn sie mit exzellentem

Know-how in Forschung und Entwicklung vorhanden und

international anerkannt sind – wie in Deutschland –

gehören auf gleicher Ebene zur gesamtgesellschaftlichen

und politischen Verantwortung wie andere Ziele des

Umweltschutzes. Heutige und künftige Sicherheit und

Förderung von Sicherheitskultur und -technik, auch für

die Kerntechnik, können und dürfen nicht Gegenstand

eingeschränkten oder beschränkenden Handelns sein.

Forschung und Entwicklung sollte immer im Geist der

Freiheit von Wissenschaft und Technik vorgetrieben

werden dürfen.

Christopher Weßelmann

– Chefredakteur –

Editorial

TMI und Lessons Learned – danach und für die Zukunft


Did you know...?

Growing Popular Support for Nuclear Power in Belgium

Forum Nucléaire, the association of the nuclear sector in Belgium,

commissioned the 2019 edition of its regular opinion poll

with Kantar TNS. Between 15 July and 6 September 2019

756 Belgians over the age of 16 were interviewed by phone

about their opinion on nuclear power and the nuclear sector

in Belgium. The opinion poll shows a clear increase in support

for nuclear power in Belgium compared to the 2017 edition

of the poll, in particular with regard to the security of supply and

the reduction of CO 2 -emissions. Below you can find a selection

of results. The complete poll was published 18 October 2019

on www.forumnucleaire.be and is available there in Dutch and

French.

For further details

please contact:

Nicolas Wendler

KernD

Robert-Koch-Platz 4

10115 Berlin

Germany

E-mail: presse@

KernD.de

www.KernD.de

DID YOU EDITORIAL KNOW...?

509

For our country the production of nuclear energy in Belgium means …

2017

(n=1027)

49 %

13 %

38 %

2019

(n=756)

57 %

More advantages than inconvenients 9 %

Advantages

and inconvenients equally

32 %

More inconvenients than advantages

Are you favourable to keeping nuclear in Belgium for the production of electricity?

2017

(n=1027)

30 %

50 %

19 % 2 %

2019

(n=756)

46 %

Yes, in long term too

37 %

Yes, but in short term (2025)

16 %

No

1 %

Don't know

The current legislation foresees the closure of all nuclear power plants by 2025.

Do you think that this can be realized without endangering the energy supply? (n=756)

69 %

No

28 %

Yes

3 %

Undecided,

don't know

It is good to replace nuclear power plants with gas fired power plants

even if they emit much more CO 2 . (n=756)

26 %

I don't agree at all

51 %

I rather don't agree

1 %

I don't know

11 %

I neither agree or disagree

7 %

I rather agree

5 %

I completely agree

Figures in percent. Rounded values.

Source:

Forum Nucléaire

Did you know...?


510

Issue 11/12 | 2019

November/December

CONTENTS

Contents

Editorial

TMI and Lessons Learned – Afterwards

and for the Future E/G 507

Did you know...? . . . . . . . . . . . . . . . . . . . . . . . . . . . .509

Inside Nuclear with NucNet

Key to Unlocking Bulgaria’s Belene Project

is Finding Right Financing Package 512

Feature | Environment and Safety

Development of Performance Shaping Factors for Human Error

Reduction during Reactor Decommissioning Activities through

the Task Analysis Method 515

Environment and Safety

Root Causes of the Three Mile Island Accident 521

Spotlight on Nuclear Law

The New Radiation Protection Law –

Protection Against Radon G 525

Research and Innovation

Evaluation of a Double-Ended Guillotine LBLOCA Transient

in a Generic Three-Loops PWR-900 with TRACE Code

Coupled with DAKOTA Uncertainty Analysis 526

Experiment Research on the Insurge Transient Behavior

of Gas-steam Pressurizer under Various Pressure 533

Decommissioning and Waste Management

Decommissioning & Dismantling of the Rossendorf

Research Reactor RFR | Part 1 G 537

First On-site Demonstration of Laser- based Decontamination

Technology in Germany 543

Report

The 25 th Edition of the Frédéric Joliot/Otto Hahn Summer School

on Nuclear Reactors “Physics, Fuels and Systems“ 549

Special Topic | A Journey Through 50 Years AMNT

Protection of Man and Environment – Nuclear Usage

Outside of Energy Sector G 550

KTG Inside . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .555

Cover:

Korea Kori NPP.

Copyright: ©Korea Kori NPP

News . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .556

Nuclear Today

Taking a Leaf out of Greta’s Climate Change Book 562

G

E/G

= German

= English/German

Imprint 542

Insert: INFORUM – Seminarprogramm 2020

Contents


Feature


515 Development of Performance Shaping

Factors for Human Error Reduction

during Reactor Decommissioning Activities

through the Task Analysis Method

511

CONTENTS

Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam


521 Root Causes of the Three Mile Island Accident

Zoltan R. Rosztoczy


526 Evaluation of a Double-Ended Guillotine LBLOCA Transient

in a Generic Three-Loops PWR-900 with TRACE Code

Coupled with DAKOTA Uncertainty Analysis

Andrea Bersano and Fulvio Mascari


537 Decommissioning & Dismantling

of the Rossendorf Research Reactor RFR | Part 1

Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz

543 First On-site Demonstration

of Laser- based Decontamination Technology in Germany

Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann,

Wolfgang Lippmann and Antonio Hurtado

Contents


512

INSIDE NUCLEAR WITH NUCNET

Key to Unlocking Bulgaria’s Belene

Project is Finding Right Financing Package

Last year the government decided to formally revive the nuclear power station project. The process is now

underway to select an investor.

The History of Belene

The project goes back to the mid-1970s when a

two-unit nuclear station at Belene, on the

Danube river in the north of the country, was planned by

Bulgaria’s communist government. Site work began in

1981 and construction of the first Russian-supplied

VVER-1000 began in 1987. This was aborted in 1991 due to

lack of funds and environmental and financial concerns.

The project was revived in 2008, but formally abandoned

in 2012 after failing to find investors and with Sofia facing

pressure from Washington and Brussels to limit its energy

dependence on Russia, which was under contract to build

the facility. The estimated €10bn cost of the project was

too high to be funded by Bulgaria alone. In June 2018, the

government decided to formally revive the project. A call

for interest was published and seven companies, three of

which are Bulgarian, applied to invest in the project.

Russia’s state-owned Rosatom, China’s state-owned CNNC

and state-run Korea Hydro and Nuclear Power Company

all filed applications confirming their interest. Press

reports said that among the three Bulgarian companies

there are two consortia with interests in energy that have

been formed specifically for the Belene project and a third

company, IPK-UP, which is registered in Kozloduy, but

lacks any registered assets or income. The seventh

company that filed its interest is Karlsruhe-registered

Becktron-Liaz Technical Engineering.

Bulgaria’s Nuclear Status

Bulgaria, which joined the EU in 2007, closed four 440-MW

VVER reactors at the Kozloduy nuclear station for safety

reasons as part of its accession treaty with Brussels.

Kozloduy has two newer 1,000-MW VVER reactors in commercial

operation, supplying about 34 % of the country’s

electricity. These two units have been renovated with

financing from Euratom, the US ExportImport Bank,

Citibank and Russia’s Roseximbank. Kozloduy-6 recently

received a licence to operate for another 10 years until

2029; Kozloduy-5 received a 10-year extension in 2017.

Reviving the Belene Project

The idea to revive the Belene project was born during

negotiations with the European Union on Bulgaria’s

accession. The centrist government at the time said Belene

could replace generation that would be lost with the

retirement of the four older Kozloduy units. In 2005, a

socialist-led government revived Belene and a tendering

procedure for a 2000-MW station was approved. A year

later, Bulgaria chose an offer by a proposed consortium

between Atomstroyexport (ASE) of Russia, Areva (now

Framatome) of France and Siemens of Germany for the

deployment of two Generation III VVER-1000 PWRs. In

2008, Bulgaria signed a contract with ASE for the

design, construction and commissioning of the two units.

However, in 2012 a new government headed by current

prime-minister Boyko Borissov cancelled Belene after

failing to find financing. Russia took Bulgaria to international

arbitration over the cancellation.

Bulgaria again began looking for ways to revive Belene

after it lost the arbitration and paid € 600 m in 2016 in

compensation to ASE for components which had already

been ordered. The 2016 decision by the Geneva-based

International Court of Arbitration of the International

Chamber of Commerce said Bulgarian state energy

company Nationalna Elektricheska Kompania (NEK) had to

pay € 620 m to ASE and assume ownership of the components.

The price was reduced after interest adjustments.

By the beginning of 2018 Russia had delivered most of the

equipment that was at the centre of the court ruling,

including two reactor pressure vessels, the RPV heads, the

full steam generator sets and the main pipelines. The

equipment is stored at the Belene site.

A Question Of Financing

According to energy minister Temenuzhka Petkova Bulgaria

wants to build the two units within 10 years and at a cost of

up to € 10 bn. However, the exact nature of how the station

will be financed remains unclear. Ms Petkova told parliament

that the government would like state energy company

NEK to have a “blocking stake” – potentially giving it a veto if

necessary – in the project, but only by contributing existing

assets and infrastructure at the Belene site, valued at

€ 1.5 bn. The government’s policy is to attract private investment

for the project with no state guarantees or long-term

electricity purchase contracts. Critics have said that would

be difficult given the magnitude of nuclear new-build

projects.

The Need For Nuclear

Bulgaria says it needs nuclear because it is heavily dependent

on imported primary energy resources and uses the most

electricity relative to GDP in the EU. Bulgaria imports almost

100 % of its oil and gas, 100 % of its nuclear fuel and about

35 to 40 % of its coal. It exports electricity to neighbouring

countries, but the electricity sector is dependent on imports

of primary energy, mainly from Russia. Russia sees Bulgaria

as a transit point into the rest of Europe for Russian energy

sources because it bypasses Ukraine.

Bulgaria is under pressure by the EU to reduce carbon

emissions in line with the bloc’s climate ambitions. Bulgaria

generates about 40 % of its electricity from coal and new

nuclear could replace a major proportion of this. The government

has also hinted that new reactors might be needed

to replace the two Kozloduy units, which are scheduled to

be retired in 2027 and 2029. A report by the Bulgarian

Academy of Science said electricity demand in Bulgaria is

expected to increase after 2030 and new nuclear could be

economically viable if construction costs are kept in check.

What Happens Next?

Mr Borissov’s ruling party and its allies command a

comfortable majority so plans to build new units at Belene

would not appear to pose a political problem. The

opposition socialists also support nuclear energy. Public

support is high, with almost two-thirds of Bulgarians

voting in support of nuclear in a 2013 referendum,


Key to Unlocking Bulgaria’s Belene Project is Finding Right Financing Package


although the result was not binding because of the low

turnout. The EU and the US have urged Bulgaria to take

stock of its dependence on Russian gas and oil and

encouraged it to seek Western investors for Belene. The

key is to find the right financing package. China is said to

be asking for financial guarantees from Sofia in return for

its participation, but Sofia favours the creation of some

form of public- private partnership. Any financing scheme

will need to be approved by the EU’s state aid watchdog.

A risk analysis of the Belene project, prepared by the

Vienna International Centre for Nuclear Competence for

the Bulgarian Atomic Forum, recommended that the

government reconsider its position of non-participation in

the project. Globally, there is no precedent in which nuclear

projects are implemented without state involvement, the

analysis said. Different forms and levels of participation are

possible, including direct sponsorship, loan guarantees, tax

credits, long-term energy purchase agreements and price

differences, the analysis noted. Another option is to use the

Belene equipment for a new unit at Kozloduy, but this

would incur additional costs. Abandoning the project and

selling the equipment as scrap would result in a loss of

about $ 1.5 m. The sale of the project is also a possibility to

companies from Russia or China.

Whoever the project owner and investors are, the

government has said the main contractor for the nuclear

part of the project will be Russia’s Rosatom, because of the

nature of the technology already paid for by Bulgaria.

Turbine hall equipment and instrumentation and control

systems could be supplied by US-based General Electric

and France’s Framatome. Both companies participated in

the recent tender, but as equipment suppliers rather than

investors.

Author

David Dalton

Advertisement

INSIDE NUCLEAR WITH NUCNET 513

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Key to Unlocking Bulgaria’s Belene Project is Finding Right Financing Package


CALENDAR 514

Calendar

2020

12.01. – 16.01.2020

Power Plant Simulation Conference. Chattanooga,

Tennessee United States, Society for Modeling &

Simulation International, www.scs.org

13.01. – 14.01.2020

ICNPPS 2020 – 14 th International Conference on

Nuclear Power Plants Safety. Zurich, Switzerland,

Waset, www.waset.org

20.01. – 21.01.2020

6 th Central & Eastern Europe Nuclear Industry

Congress 2020. Prague, Czech Republic, SZWGroup,

www.szwgroup.com

22.01.2020

31.03. – 03.04.2020

ATH'2020 – International Topical Meeting on

Advances in Thermal Hydraulics. Paris, France,

Société Francaise d’Energie Nucléaire (SFEN),

www.sfen-ath2020.org

19.04. – 24.04.2020

International Conference on Individual

Monitoring. Budapest, Hungary, EUROSAFE,

www.eurosafe-forum.org

20.04. – 22.04.2020

World Nuclear Fuel Cycle 2020. Stockholm,

Sweden, WNA World Nuclear Association,

www.world-nuclear.org

06.06. – 12.06.2020

ATALANTE 2020. Montpellier, France, CEA,

www.atalante2020.org

08.06. – 12.06.2020

20 th WCNDT – World Conference on

Non-Destructive Testing. Seoul, Korea, EPRI,

www.wcndt2020.com

15.06. – 19.06.2020

International Conference on Nuclear Knowledge

Management and Human Resources Development:

Challenges and Opportunities. Moscow,

Russian Federation, IAEA, www.iaea.org

15.06. – 20.07.2020

Nuclear Fuel Supply Forum. Washington, D.C., USA,

NEI, www.nei.org

WNU Summer Institute 2020. Japan, World Nuclear

University, www.world-nuclear-university.org

10.02. – 14.02.2020

01.09. – 04.09.2020

37 th Short Courses on Multiphase Flow. Zurich,

Switzerland, Swiss Federal Institute of Technology

ETH, www.lke.mavt.ethz.ch

IGORR – Standard Cooperation Event in the International

Group on Research Reactors Conference.

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

10.02. – 14.02.2020

09.09. – 10.09.2020

ICONS2020: International Conference on Nuclear

Security. Vienna, Austria, The International Atomic

Energy Agency (IAEA), www.iaea.org

02.03. – 06.03.2020

International Workshop on Developing a National

Framework for Managing the Response to

Nuclear Security Events. Madrid, Spain, IAEA,

www.iaea.org

08.03. – 12.03.2020

WM Symposia – WM2019. Phoenix, AZ, USA,

www.wmsym.org

08.03. – 13.03.2020

IYNC2020 – The International Youth Nuclear

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

18.03. – 20.03.2020

12. Expertentreffen Strahlenschutz. Bayreuth,

Germany, TÜV SÜD, www.tuev-sued.de

25.03. – 27.03.2020

H2020 McSAFE Training Course. Eggenstein-

Leopoldshafen, Germany, Karlsruher Institute für

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

29.03. – 02.04.2020

PHYSOR2020 — International Conference on

Physics of Reactors 2020. Cambridge, United

Kingdom, Nuclear Energy Group,

www.physor2020.com

31.03. – 02.04.2020

4 th CORDEL Regional Workshop on

Harmonization to support the Operation and

New Build fo NPPs including SMRs. Lyon, France,

NUGENIA, www.nugenia.org

KERNTECHNIK 2020.

Berlin, Germany, KernD and KTG,

www.kerntechnik.com

05.05. – 06.05.2020

10.05. – 15.05.2020

ICG-EAC Annual Meeting 2020. Helsinki, Finland,

ICG-EAC, www.icg-eac.org

11.05. – 15.05.2020

International Conference on Operational Safety

of Nuclear Power Plants. Beijing, China, IAEA,

www.iaea.org

12.05. – 13.05.2020

INSC — International Nuclear Supply Chain

Symposium. Munich, Germany, TÜV SÜD,

www.tuev-sued.de

17.05. – 22.05.2020

BEPU2020, Giardini Naxos. Sicily, Italy, NINE,

www.nineeng.com

18.05. – 22.05.2020

SNA+MC2020 – Joint International Conference on

Supercomputing in Nuclear Applications + Monte

Carlo 2020, Makuhari Messe. Chiba, Japan, Atomic

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

20.05. – 22.05.2020

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

NEI, www.nei.org

31.05. – 03.06.2020

13 th International Conference of the Croatian

Nuclear Society. Zadar, Croatia, Croatian Nuclear

Society, www.nuclear-option.org

VGB Congress 2020 – 100 Years VGB. Essen,

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

09.09. – 11.09.2020

World Nuclear Association Symposium 2020.

London, United Kingdom, WNA World Nuclear

Association, www.world-nuclear.org

16.09. – 18.09.2020

3 rd International Conference on Concrete

Sustainability. Prague, Czech Republic, fib,

www.fibiccs.org

16.09. – 18.09.2020

International Nuclear Reactor Materials

Reliability Conference and Exhibition.

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

28.09. – 01.10.2020

NPC 2020 International Conference on Nuclear

Plant Chemistry. Antibes, France, SFEN Société

Française d’Energie Nucléaire,

www.sfen-npc2020.org

28.09. – 02.10.2020

Jahrestagung 2020 – Fachverband Strahlenschutz

und Entsorgung. Aachen, Germany, Fachverband

für Strahlenschutz, www.fs-ev.org

26.10. – 30.10.2020

NuMat 2020 – 6 th Nuclear Materials Conference.

Gent, Belgium, IAEA, www.iaea.org

09.11. – 13.11.2020

International Conference on Radiation Safety:

Improving Radiation Protection in Practice.

Vienna, Austria, IAEA, www.iaea.org

30.03. – 01.04.2020

INDEX International Nuclear Digital Experience.

Paris, France, SFEN Société Française d’Energie

Nucléaire, www.sfen-index2020.org

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

Calendar


Feature | Environment and Safety

Development of Performance Shaping

Factors for Human Error Reduction during

Reactor Decommissioning Activities

through the Task Analysis Method

Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam

In this study, we present the process of deriving Performance Shaping Factors (PSFs) and developing a classification

system for them to account for human errors that may occur during the decommissioning activities due to the shutdown

of the Kori Nuclear Power Plant Unit 1. Furthermore, Human Reliability Analysis (HRA) for the reduction of human

errors was performed on the Reactor Pressure Vessel Internal (RPVI) cutting work. Task Analysis was conducted on the

RPVI decommissioning activities and possible human errors were identified. The PSF selection criteria that affect the

decommissioning activities were set up based on the human errors identified through the results of the Task Analysis

and review of the PSFs applied to the field of nuclear energy power generation. Finally, the PSFs were derived and their

classification system was developed.

1 Introduction

In recent years, industries have experienced a reduction in

the rate of accidents due to technical problems because of

the development of technologies for accident prevention.

However, it is difficult to evaluate the reliability of such

systems without considering the impact of human errors

by their operators. In particular, according to real statistical

accident data, about 60 to 90 % of all accidents are due to

human errors, while the remainder is due to technical

errors in the system [1-3]; thus, it is evident that even

minor operator errors in real life can severely undermine

the operational performance of the system. Furthermore,

in large or crucial systems, such as large power plants

( including nuclear power plants), air transportation, and

railways, human error can lead to large and more harmful

accidents; therefore, the reduction of human errors in the

case of such systems requires significant attention [4].

At present, Korea is preparing to safely and economically

decommission Unit 1 of the Kori Nuclear Power Plant,

whose operation was permanently suspended in June

2017. Various technological developments are being

conducted for the decommissioning of Kori Unit 1 with the

primary objective of ensuring the safety of human life and

property and improving economic efficiency. Therefore, it

is important to establish countermeasures to decrease

human errors during this decommissioning, because these

human errors not only hinder the safety of the operators,

but also affect the economics of the project by causing

delays in the decommissioning schedule. In order to do so,

human error trends and influencing factors need to be

identified through Human Reliability Analysis (HRA) [1];

in particular, HRA is used to identify, model, and quantify

the likelihood of human error [3]. Figure 1 shows the

flowchart for HRA that needs to be applied for the cutting

of the Reactor Pressure Vessel Internal (RPVI) components,

which is the decommissioning activity targeted in this

study.

It is important to conduct an HRA for the cutting

operation of RPVI components, because they have the

highest levels of radioactivity among all the other parts of

the unit, and this operation is highly susceptible to human

errors because of the use of remote equipment, among

others. Therefore, in particular, the objective of this study

is to obtain the Performance Shaping Factors (PSFs) that

| Fig. 1.

Procedure to perform the HRA in the decommissioning activity [5].

affect RPVI cutting and develop a classification system

for these PSFs.

Task characteristics, procedures, and information on

the decommissioning of RPVI components were identified

by reviewing related domestic and foreign literature as

well as by examining previous cases of decommissioning of

overseas nuclear power plants. In addition, task analysis

for the decommissioning of the RPVI was conducted

to identify probable human errors. First, potential PSFs

were identified for this study based on a combination of

the expected human errors from the detailed task

analysis and the general PSFs from the evaluation of

PSFs in nuclear facilities. Then, the selection criteria for

the PSFs affecting the decommissioning activities were

established. Finally, the PSFs were derived to construct

the classification system.

FEATURE | ENVIRONMENT AND SAFETY 515

Feature

Development of Performance Shaping Factors for Human Error Reduction during Reactor Decommissioning Activities through the Task Analysis Method ı Byung-Sik Lee, Hyun-Jae Yoo and Chang-Su Nam



Task

Task Information

and Requirements

Task to Consider

Working Time

Teamwork and

Communication

Workload

Task Support

Workplace Factors

Hazard Identification

Expected PSFs

2 Task Analysis

Task Analysis (TA) is the initial step in the case of human

error assessment. It provides the characteristics of different

tasks, their vulnerabilities, and properties by understanding

the objectives, performance methods, scopes of

the tasks, and procedures involved [6]. In addition, TA

helps eliminate the conditions that can cause errors before

they occur by providing detailed information about the

tasks as well as other information for predicting and

preventing errors.

2.1 TA Method

Although several TA methods exist, considering the

physical characteristics (complexity) and Human Machine

Interface (HMI) aspect (using remote devices) of the

decommissioning activities, we utilized the Hierarchical

TA (HTA) method.

HTA is a systematic and detailed TA method that is used

to achieve task objectives. It is appropriate for not only

identifying detailed task configuration and conditions, but

also expressing complex task steps as a simple, hierarchical

structure. Furthermore, the HTA method was used because

it can easily describe the characteristics of the work

involved and identify significant information about the

Title

| Tab. 1.

Details of tasks to be performed for TA.

• Task influencing factors

(cutting size, number of cutting operations, precision, etc.)

• Task output requirements

• Record feedback to indicate adequacy of action taken

• Alarms and warnings

• Actions to be taken

• Equipment needed (type, size, performance,

required utility, equipment usage constraint, etc.)

• Task frequency and required accuracy

• Physical position of the operator

(standing, sitting, squatting, etc.)

• Biomechanics

• Movement

(lifting, pushing, rotating, pulling, swaying, etc.)

• Force required

• Unit working time based on “work contents”

• Additional work hours taking into account “work support”

and “workplace environmental conditions”

• Number of work shifts and workers per work shift

• Cooperation required between the teams performing

the work

• Personal communication for monitoring or

taking control actions

• Cognitive workload

• Physical workload

• Overlap of task requirements (serial versus parallel task

elements)

• Special and protective clothing for work

• Job aids, procedures, or reference materials needed

• Required auxiliary tools and equipment

• Ingress and egress paths to the work site

• Workspace required to perform the task

• Typical workplace environmental conditions

(e.g. lighting, temperature, noise, etc.)

• Work breaks taking into account “work support” and

“ workplace environmental conditions”

• Identify work-related hazards,

e.g. potential personal injury hazards

Examples include:

• Stress

• Time pressure (critical path operations)

• Extreme environmental conditions

• Reduced staffing

HMI, communication and decision making processes, as

well as possible accidents. The HTA method involves

describing the manner in which tasks need to be performed

after establishing their overall objectives and classifying

them into their sub-tasks [6].

Using a tabular format to perform HTA allows one to

express complex tasks that require significant skill in a

suitable manner, because one can include detailed notes, if

necessary. In this study, we comprehensively reviewed

various items, such as HMI, Communication, Time, and

Accident, for the decommissioning activity in a tabular

format; this is shown in Table 1 [5].

2.2 Target Decommissioning Activity for TA

In general, one of the most challenging tasks during plant

decommissioning is believed to be the removal of the

highly radioactive internal components of the reactor pressure

vessel (RPV); this is true for Kori Unit 1 as well. In

addition, another reason that this is one of the most

difficult activities is because these radioactive components

must be dismantled and cut underwater owing to the

severe radiological conditions of the RPVI components

[7-8]. Therefore, it is recommended that the reactor

internals be removed as early as possible in the plant

dismantling sequence, so that these water systems and

their associated support systems can be released for

decommissioning, which minimizes the costs associated

with maintaining these systems in operation after

permanent plant shutdown [8].

The cross-section of the RPV with the primary internal

components at Kori Unit 1 is shown in Figure 2. As can be

seen from the figure, the internal structures adjacent to the

core barrel active region are the most highly activated, and

in most cases, include intermediate level waste components

that might require removal prior to the disposal of the

remainder of the RPV and reactor internal components

[7]. Thus, in this study, this RPV internal segmentation

activity is selected from among the various dismantling

activities in Kori Unit 1.

Furthermore, in this study, the TA was performed for

the most complex and highly radioactive RPVI cutting task

among various disassembly activities by using the HTA

method. The sequence of operations for each sub-activity

in this target task is listed in Table 2.

2.3 TA Results

The TA for the RPVI Dismantling Activity was performed as

per the activities listed in Table 2 based on the items listed

in Table 1. In our study, this analysis was performed for

each of the 10 sub-activities. The summary of the TA for

the RPVI Dismantling Activity is given as follows.

p In general, in the decommissioning of nuclear power

plants, the cutting of parts in the RPVI is the most

complicated and difficult task during the dismantling

process. Therefore, the influence on the internal factors

of the workers was evaluated in order to have a relatively

high value in terms of operator internal response.

p This kind of dismantling operation, which is complex

and requires accuracy and reliability, is significantly

influenced by the internal and external characteristics

of the worker. Therefore, sufficient education and

training is required. However, as the driving principle

and operation method of these cutting equipment and

accessories ( cutting equipment, remote control device,

display, auxiliary equipment, etc.) are not complicated

and operation is relatively simple, an operator is not

required to have considerable experience in using them.

Feature



No Sub activity Activity Description

1 Control Rod Guide Tube Upper Area Cutting

and Packaging

2 Control Rod Guide Tube Lower Area Cutting

and Packaging

1) Lift up the Upper Internal Assembly and fix it to the turntable

2) Cut the control rod guide tube upper area

3) Transfer and package the cut section to a storage container

1) Cut the control rod guide tube lower area

2) Transfer and package the cut section to a storage container

3 Upper Plate Cutting and Packaging 1) Move and release the Upper Support Plate in an empty space in the reactor tank

2) Cut the Upper Core Plate fixed to the turntable

3) Fix the Upper Core Plate to the turntable and cut the Upper Support Plate

4) Transfer and package the cut pieces to a storage container

4 Baffle Fixed Bolt Head Cutting 1) Install the mechanical drill for Baffle Separation

2) Cut the Baffle Fixing Bolt Head by placing the mechanical drill inside the RPV

5 Baffle Cutting and Packaging 1) Lift the Baffle and remove it from the Former

2) After fixing it to the turntable, cut the Baffle

3) Transfer and package the cut Baffle pieces to a storage container

6 Core Barrel Lower Area Cutting 1) Lift up the Lower Internal Assembly and fix it to the turntable

2) Rotate the turntable to cut the Lower Internal Assembly

3) Lift the upper area of the Lower Internal Assembly into the Vessel

7 Lower Internal Structure Assembly Cutting

and Packaging

8 Thermal Shield Separation, Cutting and

Packaging

1) Cut the Instrument Nozzle from the Core Support Structure Assembly

2) Transfer and package the cut nozzle to a storage container

3) Cut and package the tie plate fixed to the Turntable

4) Fix the Lower Core Plate to the Turntable and cut it


1) Lift the upper area of the Lower Internal Assembly and fix it to the turntable

2) Separate the Thermal Shield by cutting the Bolt Head

3) Release the removed Core barrels from the Thermal Shield inside the vessel

4) After fixing the Thermal Shield to the Turntable, remove the Irradiation Specimen Guide

5) Cut the Turntable Thermal Shield Upper and Lower Panels

9 Former Separation 1) Lift the Core Barrel to the turntable and fix it

2) Cutting the Former fixing bolt head outside the Core Barrel

3) Separate the Former from the Core Barrel

4) Transfer and package the separated Former to a storage container

10 Core Barrel Cutting and Packaging 1) Fix the Core Barrel to the turntable and cut it

2) Temporarily release the cut Upper Core Barrel in the Vessel

3) Segment the Lower part of the Core Barrel fixed to the turntable


5) Repeat the procedure for the cutting the remaining Core Barrel


| Tab. 2.

Task Description of RPVI Dismantling Activity.

p Considering the characteristics of the work (equipment

and facilities, the object to be cut, and clothing for

radiation protection in the work environment), a

detailed work plan must be established in advance.

Further, as this cutting work is time-consuming, the

psychological and physical influences that the supervisor

and the worker can receive are considerable.

p If the cutting activity is dangerous, takes a long time,

and has associated time constraints, it should be

performed during the day/night time. In this case,

various difficulties (such as break time, clothing

discomfort, and physio logical factors) are generated.

Because these difficulties have a significant impact on

the internal and external factors affecting the worker,

much cooperation and communication is required

between the worker and the supervisor in this working

environment.

p The radiation and the physical environment of the

workplace are the major risk factors for the workers,

and the influence of these working environments on the

internal and external factors of the workers was

con siderable.

p Nuclear decommissioning work is not a frequently

performed task. Therefore, workers may have insufficient

experience and education/training. Therefore,

after the decommissioning activity has been carried

out, it is necessary to feedback the results of the work to

be reflected in the necessary work procedures and to be

managed as experience data.

| Fig. 2.

Internal components for the RPV in Kori Unit 1 [7].

3 PSFs in the Decommissioning Activities

Because the workers’ activities are the fundamental factors

that renders the system vulnerable, it is necessary to

identify, model, and quantify the possibility of human

error using HRA [3]. In particular, the nominal Human

Error Probability (HEP) used in the HRA is estimated

Feature




based on the workers’ activity, most often based on the

PSFs. This is because PSFs characterize the important

aspects of human error and provide numerical criteria for

adjusting the average HEP level [9]. Therefore, the key

step in the HRA for decommissioning activities of a nuclear

power plant is to select appropriate PSFs, develop the

classification system, and quantify their impacts on the

decommissioning activity.

3.1 Criteria for selecting PSFs

When selecting PSFs that affect the decommissioning

activities of nuclear plants, it is important to carefully

select all the important situational factors that affect the

hazard level so as to not miss any of the factors. In addition,

it should be ensured that there is no overlap in the meaning

and scope of the PSFs; furthermore, the factors should be

selected based on their actual effect on human error

analysis. Therefore, the PSFs are selected for the HRA of

the decommissioning activities of the nuclear power plant

considering the following criteria.

p The PSFs should be selected to appropriately reflect the

work characteristics. Therefore, the TA results of the

work should be fully reflected by the PSFs.

p The internal and external factors that might affect

the workers’ performance should be comprehensively

considered.

p When a worker encounters an abnormal event, the

internal factors should be included in the PSFs

appro priately to account for the reactions that occur

naturally in workers, including physical, cognitive, and

emotional factors.

p The external factors should comprehensively include

those that directly or indirectly affect the workers’

response to performance, i.e., the business organization

and work environment factors, among others.

p Based on previous research results, in our case, the

PSFs will consist of three levels (namely, Level 1, Level

2, and Level 3). Though there are direct and indirect

dependencies between the PSFs in Level 1 and Level 2,

the PSFs are selected in a manner that there is no

dependency between the Level 3 PSFs.

p In practice, the PSFs are selected based on factors

that directly affect the trends of the ongoing events.

Therefore, factors such as workers’ training and work

shifts, among others, indirectly affecting the work

performance of the workers, are excluded from the PSF

Classification Direct effect Indirect effect

Personal Factors

System factors

Task factors

| Tab. 3.

Derived PSF factor.

• Duration of mental stress

• Mental tension

• Pain or discomfort

• Hunger or thirst

• Emotional state

• Duration of physical stress

• Disruption of circadian rhythm

• Lack of sleep

• Work hours

• Shift rotation

• Suddenness of onset

• State of current practice or skill

• Motivation and attitudes

• Personality and

intelligence variables

knowledge of required

performance previous

training/experience

• Sensory deprivation

• Distractions (noise, glare,

movement, flicker. color)

• Complexity

• Movement constriction

• Workplace layout

• Threats of failure

• Lack of physical exercise

• High jeopardy risk

• Conflicts of motives

about job performance

selection because they are considered to be inherent in

the other PSFs themselves.

p The PSFs should have clear definitions so that their

meanings and roles do not overlap. Therefore, as much

as possible, the scale of the PSFs should be reduced by

grouping all the PSFs and reducing the number of tasks

involved.

3.2 Selection and definition of PSFs

Based on our review of existing literature related to the

work performances of individual workers from various

industrial fields, we observed that it is not easy to find

a consensus on the factors influencing the workers’

performance. However, considering the results of these

previous studies, it is deemed that the possibility for human

error can be determined based on the degradation of the

workers’ human error factor, i.e., their task performance.

The Institute of Nuclear Power Operations (INPO)

presents 85 error precursors that can lead to possible

human errors considering the business requirements, personal

abilities, work environments, and human nature in

terms of the operation of the nuclear power plants [10].

These precursors are considered as the risk variables for

human errors made by workers.

Therefore, it is realistic to utilize an Error Precursor,

i.e., PSF, that affects work performance. While estimating

the probability of human error, it is desirable to add

psychological factors that cannot be directly managed

after the possibility of human error for manageable

factors is reduced through safety management or accident

prevention activities. The PSFs extracted for use in the

human error assessment model in this study are listed

in Table 3 below.

3.3 Development of a classification system

for PSFs

TA was conducted to derive the PSFs for RPVI cutting,

which is one of the primary tasks involved in the

decommissioning of nuclear power plants. Because it is

difficult to select PSFs such that their meanings and roles

are clearly delineated, the PSFs are classified into three

major categories, which are defined by their task analysis

results. In other words, the human factors of the workers

themselves, the operating system factors related to the

work, and the ergonomic factors linking the worker and

the dismantling activities were suitably classified.

In particular, an important aspect of the TA is “Human,”

i.e., a worker who performs the task of decommissioning

the RPV and can cause human errors. It is noteworthy

that human factors are important to consider not only

in the field of nuclear power generation, but also, in

other industrial fields, such as the railways and aviation

[11-16].

p Supervisors and workers might be affected psychologically

while performing the tasks described above.

Because the cutting process takes considerable time, it

is important to consider the various stresses and

emotional conditions that supervisors and workers

might experience.

p In particular, the work environment and conditions for

the cutting task in normal or abnormal situations affect

the workers’ physical and physiological factors. Thus, to

perform the RPVI cutting tasks, wearing the appropriate,

protective work clothes and protective cap

and having good physical and health conditions to

carry or handle the equipment are considered important.

Furthermore, while performing the tasks, the

Feature



physiological factors of the workers, such as fatigue,

hunger, and excretion, is equally important.

p When the nuclear power plant is dismantled for the

first time, there might be a lack of personal skills and

experiences, because the workers are not accustomed

to the decommissioning work. Therefore, information

about the complete decommissioning task and equipment

used for each task should be taught through

training and education.

Next, there is a need for an operating system that can be

managed, supervised, and overseen to efficiently perform

the decommissioning tasks. Therefore, the “operation”

elements, such as organization, including the supervisors

and workers; task management for managing work

schedules and workload efficiently; and documents

include detailed information on decommissioning, such as

procedure and scenarios, are important.

p In order to ensure that the cutting work is performed

smoothly, the supervisors and workers need to form

appropriate teams. Organizational factors, such as the

objectives of these team, necessary team-building

elements (cooperation, role, etc.), and precise and

detailed communication of the supervisors’ judgement/decisions

to the workers are important.

p Next, the tasks should be well-designed including the

preparation, methods and strategy, as well as processes

involved in the tasks. Because the work is actively

con ducted throughout the day, the work assignment

Level 1 Level 2 Level 3

Human

Operation

Ergonomic

System

Psychological

State

Physical

State

Performance

Capability

Organizational

Factors

Task Management

Procedure and

Information

HMI

(Human Machine

Interface)

Workplace Design

Workplace Physical

Environment

Stress

and shift work schedule should be coordinated and

managed.

p Finally, because supervisors and workers understand

and perform their work based on the procedures and

scenarios related to the decommissioning of the RPVI

components, information on regulations, equipment

and facilities, as well as accuracy and details of the

procedures are important.

Furthermore, because the cutting work is human- centric,

the surrounding situations and conditions must be

designed in a manner that is suitable for human beings.

Therefore, “ergonomic system” factors, such as worker

interaction with the necessary equipment and facilities,

which are based on the decommissioning task characteristics

and requirements for RPVI components, workplace

design based on the path of the supervisor and worker and

task types, and optimal workplace environment for

workers, are important points to consider.

p The systems and equipment the supervisors and

workers operate must be interactive to ensure that they

are not difficult to use to ensure the convenience and

safety of the workers.

p Because the RPV is located under water and is not

directly visible and accessible to humans, the cutting

processes under the RPVI decommissioning task

are generally performed using remotely controlled

devices. Therefore, the configuration, condition,

and per formance of these remote systems, such as

Emotional State (Excitement, Boredom, Accomplishment, Frustration, Dissatisfaction)

Safety Awareness

Fatigue

Physical Capability

Discomfort

Task Knowledge and Memory (Diagnoses, Action, Results, Information)

Experience

Personal Capability

Overall Planning

Supervision of Management including Decision Making

Team Factors(Roles, Coordination, Risk Management, Atmosphere, Assisting Other)

Work Process Design (considering Task Characteristic)

Workload Management

Problem Identification and Solution

Communication Availability and Quality

Procedure and Information Availability

Procedure and Information Complexity

Procedure and Information Accuracy and Completeness

Procedure and Information Feedback and Recency

Interaction Element (Menu, Push Button, Direct Manipulation, Special Symbol, Shape Type)

Familiarity of Equipment and Facility

Complexity of Equipment and Facility

Maintenance

Physical Access to Work Items

Warning Sign (Alarm Location, Quantity, Intensity, Importance and Easy of Identification)

Arrangement of Functional Areas

Safety Device

Noise

Lighting

Temperature

Radiation Level


| Tab. 4.

PSF Classification System in the Reactor Decommissioning Activity.

Feature




remote control, camera, and display should be well

established, and designers should consider functional

spatial arrangements and accessibility in order to use

them.

p An optimal working environment is required to ensure

the safety of workers during the decommissioning

work. In particular, it is important to establish an

optimal physical environment for the workers by

checking for factors, such as the noise of the equipment,

temperature experienced by the workers wearing

protective clothing, level of radiation exposure, cleanliness

of the workplace, and turbidity of the water in

which the cutting operation is being performed.

In particular, the PSFs are classified into three categories

per the level of detail and importance required for the

decommissioning activities. Level 1 includes “Human

Factors,” “Operation Factors,” and “Ergonomic System

Factors.” Next, Level 2 of Human Factors includes the

basic elements responsible for human error, namely

the psychological and physical state and performance

capability of the supervisors and workers. Level 2 of

Operation Factors is composed of Organizational Factors,

Task Management, Procedures, and Information. Finally,

Level 2 for human-oriented Ergonomic Factors is composed

of HMI, Workplace Design, and Workplace Physical

Environment. The PSFs under Level 3 are selected in a

manner so as to not overlap with each other; in particular,

these include a total of 44 factors as shown in Table 4.

4 Concluding Remarks

In this work, we study and present the methodology for

HRA for the cutting task of the Nuclear RPVI components

in order to reduce human errors during the nuclear decommissioning

activities. In order to do so, the HRA implementation

procedures were reviewed to identify the components

of the HRA and the most important TA was performed

in preparation for the HRA. The task characteristics,

procedures, and information on the decommissioning

of the RPVI components were identified by reviewing relevant

literature in the field and examining experiences in

decommissioning at overseas nuclear power plants.

In addition, PSFs required for the nuclear decommissioning

activities were identified using TA results and

PSFs selection criterion for RPVI cutting (10 tasks in total).

The results of the PSF review applied to nuclear facilities

and types of human errors identified through the detailed

TA were synthesized. Furthermore, the selection criteria

for the PSFs affecting the decommissioning activities were

set. Finally, PSFs were selected and classified using a

classification system.

In a future study, we will examine the interrelationship

among the PSFs and consider methods for assessing

the PSFs. Moreover, we will develop a framework to model

the mutual influences that exist among the PSFs with

appropriate consideration of the relationships and

dependencies among them. The collection of experience

data for decommissioning nuclear power plants will also

reduce the uncertainty in the information used to perform

HRA.

References

[1] Madonna, M., et al.: Il Fattore Umano Nella Valutazione Dei Rischi: Confronto Metodologico Fra

Le Tecniche Per L’analisi Dell’affidabilità Umana. Prevenzione Oggi. 5 (n. 1/2), 67–83 (2009).

[2] Hollnagel, E.: Cognitive reliability and error analysis method (CREAM). Elsevier, (1998).

[3] Griffith, C.D., Mahadevan, S.: Inclusion of fatigue effects in human reliability analysis. Reliability

Engineering & System Safety, 96 (11), 1437–1447 (2011).

[4] Lee, Y., et al.: Research Activities and Techniques for the Prevention of Human Errors during the

Operation of Nuclear Power Plants. Journal of the Ergonomics Society of Korea, 30 (1), 75-86

(2011).

[5] O’Hara, J. M., et al.: Human Factors Engineering Program Review Model, US NRC, NUREG-0711,

Rev. 3, 2012

[6] Center for Chemical Process Safety : Guidelines for preventing human error in process safety

(1994)

[7] Byung-Sik Lee: Optimization of reactor pressure vessel internals segmentation in KoreaATW

Vol.62 (2017), Issue 11, 654~658, November, 2017

[8] Boucau, J., et.al.: Best practices for preparing vessel internals segmentation projects. No.

NEA-PREDEC--2016. February 16–18, Lyon, France (2016).

[9] Boring, R.L.: Modelling human reliability analysis using MIDAS. In: International Workshop on

Future Control Station Designs and Human Performance Issues in Nuclear Power Plants (2006).

[10] Human performance reference manual, Institute of Nuclear Power Operations, INPO-06-003,

2006

[11] Chang, Y. H. J., Mosleh, A.: Cognitive modeling and dynamic probabilistic simulation of operating

crew response to complex system accidents. Part 2: IDAC performance influencing factors model.

Reliability Engineering & System Safety, 92(8), 1014-1040 (2006).

[12] Blackman, H. S., Gertman, D. I., Boring, R. L.: Human error quantification using performance

shaping factors in the SPAR-H method. In Proceedings of the human factors and ergonomics

society annual meeting (Vol. 52, No. 21, pp. 1733-1737). Sage CA: Los Angeles, CA: SAGE

Publications (2008).

[13] Li, P. C., Chen, G. H., Dai, L. C., Zhang, L.: A fuzzy Bayesian network approach to improve the

quantification of organizational influences in HRA frameworks. Safety science, 50(7), 1569-1583

(2012).

[14] Jung, K., Byun, S., Kim, J., Heo, E., Park, H.: An empirical study on evaluation of performance

shaping factors on AHP. Journal of the Ergonomics Society of Korea, 30(1), 99-108 (2011).

[15] Mindock, J.: Development and Application of Spaceflight PSFs for HRA. PhD Thesis , University of

Colorado at Boulder (2012).

[16] Baek, D., Koo, L., Lee, K., Kim, D., Yoon, W., Jung, M.: Taxonomy of PSFs for human error analysis

of railway accidents. Business Administration, Hanyang University (2007).

Authors

Byung-Sik Lee

Hyun-Jae Yoo

Chang-Su Nam

Dankook University

119, Dandae-ro, Dongnam-gu, Cheonan-si

Chungnam, 31116

Republic of Korea

Acknowledgments

This work was supported by the National Research

Foundation of Korea (NRF), granted financial resource

from the Ministry of Science, ICT and Future Planning

(No. 2017M2A8A5015148 and No.2016M2B2B1945086),

Republic of Korea.

Feature



Root Causes of the

Three Mile Island Accident

Zoltan R. Rosztoczy

The accident at Unit 2 of the Three Mile Island nuclear power plant, at that time operated and partly owned by

Metropolitan Edison Company, occurred 40 years ago, on March 28, 1979. Following the accident, two major

investigations were conducted, one by the President’s Commission on the Accident at Three Mile Island [1], appointed

by President Carter, and the other by the Nuclear Regulatory Commission’s Special Inquiry Group. [2] The investigations

documented the timeline of the accident and the availability and failure of equipment, and addressed operator actions

during the accident, the training of operators, and NRC procedures that applied to the event. The design process for the

plant and the designer’s responsibilities, including the plant’s safety analysis, were not addressed. Many additional

studies and papers have been published over the past 40 years, none of which have addressed the design process or the

safety analysis of the plant. The only effort specifically addressing the design of the plant and responsibility for the

accident was Metropolitan Edison’s lawsuit against Babcock & Wilcox (B&W), the designer of the plant. A trial began

but was terminated, and the case was settled out of court. The court records are sealed; information is not available.

More than 10 years prior to the TMI-2

accident, B&W was designing its first

nuclear power plant. In the designation

of safety systems and in the safety

analysis of the plant, there were two

relatively minor but important omissions.

These omissions turned out to

be the root causes of the accident. If

just one of them had been corrected

during the intervening years, the

accident would have been avoided.

The TMI design was reviewed by

utilities purchasing plants from B&W

and by the NRC. The omissions

remained undetected. The safety role

of the pilot-operated relief valve

(PORV) and the PORV block valve

were not fully appreciated. The manufacturer

of the PORV was not notified

of the valve’s safety function, namely

that it has to be able to close after

being exposed to accident loads. [2]

Also, the plant’s safety analysis report

(SAR) did not address loss-of-coolant

accidents (LOCA) initiated by very

small breaks. Unfortunately, the plant

responds very differently to an event

initiated by a stuck PORV than to the

small-break events presented in the

SAR. At the time, this was unknown.

Lessons learned from the omissions

in the TMI design are timely today,

when new types of reactors, such

as small modular reactors, are on the

drawing board. The designers of these

new systems can learn from the TMI

experience.

The initiating event

Operators attempting to clean a

condensate polisher tripped the steam

generator feedwater pumps. Then,

the plant safety system tripped the

turbine. The turbine was no longer

removing heat from the reactor coolant

system (RCS), the temperature

and pressure of the RCS started rising

rapidly, and the PORV opened, as

designed.

Upon shutdown of the feedwater

pumps, the plant’s safety system

turned on the emergency feedwater

pumps. Due to a maintenance error,

both emergency feedwater block

valves, which are supposed to be open

when the plant is operating, were

closed, so no emergency feed-water

reached the steam generators. The

closed valves caused the RCS to heat

up faster than in the case of a normal

turbine trip, and the PORV was

exposed to a larger load than normal,

most likely a heavy two-phase flow

(steam and water mixture) or water

discharge. Thus, the closed valves

could have played a role in causing

the accident. This possibility is not

addressed in the literature.

As RCS pressure increased, the

reactor protection system shut the

reactor down, after which the RCS

pressure dropped. The PORV should

have closed, but instead it stuck open,

and the plant faced a LOCA. The

obvious question is, “Why did the

PORV fail to close?”

Designers of nuclear power plants

have a dual responsibility. They must

design the plant not only for normal

operation of generating electricity,

but also for safe performance in case

of events that might occur during the

lifetime of the plant and in case of

postulated accidents.

Components of systems that have

both an operating function and a

safety function have to be identified

and designed to perform both functions

in a reliable manner. The PORV

is a good example of such a component.

During normal operation, the

PORV maintains RCS pressure below

specified limits by opening and closing

and by discharging steam from the

pressurizer. During abnormal events,

as in the TMI-2 case, the PORV could

be discharging two-phase flow or

water. The valve must be designed to

perform its safety function – namely,

to close following a two-phase flow or

water discharge. Apparently, this was

not the case at TMI. The PORV was

not designed to perform its safety

function. The purchase order failed to

specify this requirement, and the

supplier of the valve did not know that

the valve had a safety function and

that it had to close following twophase

flow or water discharge. [2]

Designers are also responsible for

incorporating operating experience

into their design. Prior to the TMI-2

accident, PORVs failed to close seven

times at B&W plants. [2] Despite

this record, the PORV itself was not

modified or replaced. Instead, an

indicator light was installed to show

whether the block valve upstream of

the PORV had received a signal to

close, but there was no indication in

the control room that the valve had

actually closed.

Reprinted with

permission from the

March 2019 issue of

Nuclear News

Copyright © 2019 by

the American Nuclear

Society

| From right to left: President Jimmy Carter, Pennsylvania Gov. Richard

Thornburgh, and the NRC’s Harold Denton tour the TMI-2 control room on

April 1, 1979. Photo: The Jimmy Carter Presidential Library and Museum

521

ENVIRONMENT AND SAFETY


Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy


ENVIRONMENT AND SAFETY 522

PORVs opened relatively frequently

on B&W-designed pressurized

water reactors. The thermal hydraulic

design of the reactor core was closer

to acceptable limits than other PWR

cores, and the amount of water

contained in the secondary side of

the steam generators was very small –

only 25 percent of some other PWRs’

water content. [2] These differences

made the system react faster to

changes. With quick plant response,

the PORV came into action relatively

frequently. More frequent use of the

PORV led to more frequent failures.

The PORV failure at TMI-2 was the

eighth at a B&W plant, more than

an order of magnitude higher than

PORV failures with other reactor

designs. [2]

The NRC has specific requirements

for equipment related to safety. Equipment

essential to accident mitigation

and equipment whose failure can

cause or aggravate an accident are

considered “safety related.” A stuckopen

PORV causes a breach in the

boundary of the RCS, creating a

LOCA. Among postulated accidents,

LOCAs are considered to be the

most serious, and therefore they

receive special attention. Nuclear

power plants are designed with three

barriers to protect the public from

radioactive material release: The

fuel is enclosed in a sealed cladding,

the reactor core is within the closed

RCS, and the RCS is covered by a

containment building. Among all

postulated accidents, there is only

one – the LOCA – where two of the

barriers are predicted to be damaged.

In the case of a LOCA, the event itself

breaches the RCS, and the predicted

consequences of the accident are

expected to damage some of the

fuel cladding. Protection of the public

is reduced to a single barrier, the

containment building. Furthermore,

valve failures are more likely than

pipe breaks. Thus, the most likely

LOCA is the stuck-open PORV.

Surprisingly, the PORV was not

identified by the designer as safetyrelated

equipment. The design was

reviewed by Metropolitan Edison

and evaluated by the NRC. Neither

objected to the PORV not being

designated as safety related, and

the NRC approved the construction

permit application. Had the PORV

been designated as safety-related

equipment, it would have had to

meet reliability requirements and be

tested under accident conditions. If

the TMI-2 PORV had been tested, it

most likely would not have passed.

Following the accident, the manufacturer

of the valve stated that the

TMI-2 PORV was not qualified to close

following a two-phase flow or water

discharge. [2] If the PORV had been

designated as safety related, it would

have been replaced or modified.

The reason given for not designating

the PORV as safety related was

the presence of a block valve upstream

of the PORV. If the PORV is stuck

open, the block valve can be closed,

terminating the accident. Thus, the

block valve is essential for the mitigation

of a PORV failure accident, and

it is also considered safety-related

equipment. It must have automatic

safety-grade actuation initiated from

the stuck-open PORV or, if the initiation

is manual, safety-grade position

indication must be available in the

control room with sufficient time

for operator action. Neither of these

conditions existed at TMI-2.

Consequences of PORV failure

Part of the designer’s responsibility is

to conduct a complete and detailed

safety analysis of the plant. The

analysis must include transients that

might occur in the plant. The analysis

of transients must show that continued

operation of the plant following

these events is justified. The

plant’s safety analysis also has to

address all potential accidents, both

system failures and operator errors

that the plant could be subject to,

unless they are considered to be

extremely unlikely (severe accidents).

It is the designer’s responsibility to

identify all accident types specific

to the design of the plant. In the case

of water-cooled reactors, one of these

accident types is a breach in the RCS

– a LOCA.

For PWRs such as TMI-2, it is an

NRC requirement that a complete

spectrum of breaches in the RCS be

analyzed, starting from the doubleended

break of the largest pipe in the

RCS down to the break size that the

makeup water system can keep up

with. Unfortunately, it was not emphasized

that a breach in the system

includes stuck-open valves if the

valve’s discharge area is within the

size range of the postulated accident.

The PORV falls within the size range.

Complete spectrum also means all

possible break locations. The consequences

of a stuck-open valve on

the top of the pressurizer could be

different from a same-size break at a

lower elevation.

Typically, prior to the TMI accident,

the large-break LOCA analysis

| A six-page special report – Nuclear News’s

initial coverage of the TMI accident –

was mailed separately to subscribers and

ANS members in April 1979.

included break sizes in both the hot

and cold legs of the RCS, starting

from a double-ended break down

to a 0.5 square-foot break. Usually,

the consequences were most severe

at one of the larger breaks. From

there on, smaller sizes resulted in

more favorable consequences. The

small-break LOCA analysis ran from

0.5 square foot down to about

0.1 square foot. The trend was the

same; smaller breaks had less severe

consequences. Breaks even smaller

were not analyzed for two reasons:

(1) the calculations ran long on the

computer and the analyses were

expensive, and (2) the trend was

already established. Instead, the

assumption was made that the

trend would continue down to the

smallest required size. Also, smallbreak

LOCA analysis was assumed

to be independent of break location.

Thus, breaks less than 0.5 square

foot were not analyzed at different

locations, and breaks less than

0.1 square foot were not analyzed

at all.

The safety analysis of the plant

serves many purposes. It provides

both the designer and the operator of

the plant with an understanding of

how the plant responds to a specific

event or accident, indicates potential

damage if mitigating actions are

not taken, guides the designer in the

design of the needed safety systems,

and provides information for training

the operating staff. The analysis

shows how reactor operators can

recognize a specific event and what

actions they must take and provides

the needed information for the

preparation of emergency procedures.

The results of the analysis also show




compliance with applicable regulations.

It shows that potential damage

has been mitigated and the safety

of plant personnel and the public is

ensured.

The TMI-2 LOCA analysis was

performed by the designer. It was

rather elaborate but was incomplete

in the sense that it failed to show that

very small breaks behave differently

and have more serious consequences

than small breaks. As part of the

TMI-2 licensing process, no LOCA

analysis was performed by the

designer, and no LOCA analysis was

submitted to the NRC by the utility

for a break size anywhere close to

the size of the discharge opening of

the PORV (about 2 square inches). No

analyses were performed for any size

break at the top of the pressurizer or

for a LOCA caused by a stuck-open

PORV. Unfortunately, as was learned

from the TMI accident and from

analysis performed after the accident,

a stuck-open PORV.

LOCA is very different from the

small-break LOCA analyses presented

for TMI-2. Analysis of smaller breaks

showed that the trend reverses

and the consequences increase with

decreasing break size. The plant

responds differently, reliance on

safety systems and instrumentation

changes, and different operator actions

are required. The consequences

of a PORV failure are even more

different. RCS pressure can drop while

the water level in the pressurizer is

rising. Void can form in the reactor

core and accumulate at high locations

of the RCS while the pressurizer water

level is high. Furthermore, the water

level can drop below the top of the

core, resulting in core damage, while

the pressurizer water level is still high.

Obviously, pressurizer water level

indication is not a useful tool for the

handling of this accident.

A special design feature of B&W

plants further aggravates this effect.

The pressurizer surge line is designed

with a loop seal to prevent steam from

entering the pressurizer. Eliminating

steam flow to the pressurizer prevents

water level drops in the pressurizer,

keeping the water level high, while

void is accumulating in the RCS.

Due to the lack of analysis, the

consequences of a PORV failure

were unknown. As it turned out, the

actual consequences without proper

mitigation were a lot worse than

one would expect. The assumption

that consequences get better with

decreasing break size was incorrect.

The actual consequences of the

accident equally surprised the designers,

the owner/operator of the

plant, and the regulators. The plant’s

response to the PORV failure was

totally unexpected.

Accident management

Early in the morning of March 28,

1979, four young operators at

TMI-2 realized that something had

happened, but they had no idea what

it was. The turbine shut down, the

reactor scrammed, and a cascade of

alarms sounded and flashed. The

plant was acting strangely. RCS

pressure was decreasing while the

pressurizer water level was increasing.

The operators had not faced this

situation before. It was not covered in

their training. They did not know

what to do.

The event facing the operating

crew was a stuck-open PORV and a

very small LOCA. They did not know

that was the case. There was no direct

indication of PORV position in the

control room. They could not see that

the PORV was stuck open.

Not knowing what was going on

and not having familiarity with the

event, the operators were improvising,

trying to maintain water level

in the RCS within prescribed limits.

They relied on the pressurizer water

level reading, as they were trained to

do. Unfortunately, they took a few

inappropriate actions, which included

turning off the high-pressure emergency

core cooling system, opening

the letdown line, ignoring signs of

overheating of the reactor core, and

pumping radioactive water to the

auxiliary building. All of this occurred

before they learned – two hours and

20 minutes into the accident – that the

PORV was stuck open. Then they

took corrective action and closed

the block valve.

The obvious question is, “Why

were the operators in the dark, and

why did they lack familiarity with

this event?” Their training covered

mitigation of postulated accidents, including

LOCAs. There was only one

set of accidents missing, very small

LOCAs, including PORV failure. Since

the designer did not analyze this

event, it was not included in operator

training. Not knowing the plant’s

response to a PORV failure, the

designers and the training staff

instructed the operators to always

rely on the pressurizer water level

indication for water level measurements

in the RCS. The operators

followed their training on that

morning.

Despite the total lack of training

for a stuck-open PORV event, could

the operators have realized what

was going on and taken appropriate

action? The answer is yes. [1] The

temperature of the PORV drain pipe

was monitored and showed high

readings, an alarm signaled high

water level in the containment building

sump, high neutron level indications

were observed in the reactor

core, temperature and pressure were

rising in the containment building,

and the reactor coolant pumps were

vibrating. Any of these observations,

typical of a LOCA, could have brought

attention to a stuck-open PORV. The

remedy should have been obvious:

Close the block valve.

Once the block valve was closed,

the LOCA was terminated. The next

step was to cool the core by natural

circulation of the water in the RCS.

This was not possible, however, due to

the large amount of void that had

accumulated in the RCS. The operating

staff had to improvise again to

reduce the void and the bubble in

the RCS, and then to establish neutral

circulation. It took a couple of days’

work for them to accomplish this.

Both the plant’s designer and

operator lacked the knowledge of how

the plant would respond to a stuckopen

PORV. What they did not know,

they could not pass on to the operators.

The operators’ training was

misleading, and the emergency procedure

was incorrect for the incident

they were facing.

Industry practice and

oversight

B&W’s two omissions – safety-related

classification of the PORV and the

PORV block valve, and the lack of

PORV failure analysis – were not

unique to B&W. The other U.S. PWR

designers, Westinghouse and Combustion

Engineering, made the same

omissions. How could three independent

sets of engineers make the

same mistake? Licensing of the plants

was a major consideration. The SAR

was the centerpiece of the licensing

review. Precedent provided guidance

for the preparation of the report.

Analyses presented in previous

applications were included in the

report; analysis that was not

required was ignored. Dozens of

utilities received SARs with the

same omissions. The omissions

had a direct and major effect on

the training of reactor operators.

The operators received training on

a plant simulator, with postulated






| TMI II – Late accident phase.

Zoltan R. Rosztoczy

was Manager of the

Safety Analysis

Department of

Combustion Engineering’s

Nuclear Division

in its formative years.

He later joined the

Nuclear Regulatory

Commission and was

a charter member of

the commission’s

Senior Executive

Service.

accidents programmed into the

simulator. One accident, the PORV

failure, was missing. Nobody noticed

it or took corrective action. After

PORVs failed seven times at B&W

plants, this accident was still missing

from the operator training program

and from the simulator.

In the case of PWR evaluations,

the NRC had the distinct advantage

of reviewing SARs from three

independent designers. Comparisons

among the three designs frequently

helped in the reviews. The NRC,

however, failed to recognize its own

effect on plant design and analysis.

A nuclear power plant is a complex

system. A regulatory review and

evaluation cannot address all aspects

of the design, and priorities have to

be set. There was a tendency not to

require more from an applicant than

was required from previous ones.

Spending time reviewing areas of the

design that weren’t reviewed in the

past was discouraged. Consequently,

regulators and designers addressed

the same areas of the design and the

safety analyses over and over again

and ignored other areas.

Conclusions

Failure to incorporate the safety

function of the PORV and the block

valve in the design of the plant created

the condition for the TMI accident.

With no positive indication in the

control room of an open PORV and no

positive position indication of the

block valve, the operators were left to

guess what was going on and what

needed to be done.

Not having addressed PORV failure

in the plant safety analysis, the

designers, as well as the training and

operating staff, were unfamiliar with

the plant’s response to this type of

accident. They did not know that

| TMI II – Late accident phase.

the plant conditions the operators

were facing were possible, and as

a result, training and instructions

were inadequate. When similar plant

designs are being reviewed or evaluated

one after the other, there is a

tendency to address the same issues

in each case. Plants are very complex,

and not everything can be evaluated

as part of one review. It is appropriate

to shift emphasis in subsequent

reviews and to address issues previously

not covered.

Appropriate NRC regulations relative

to LOCAs to control the design

and operation of the plants’ safety

systems and develop operator training

programs and emergency procedures

were evolving when B&W designed its

first plants, but they were in place at

the time of TMI-2’s licensing. The

problem was that some of the

regulations were not followed.

The two omissions – not recognizing

the safety function of the PORV

and the block valve, and the failure to

analyze the stuck-open PORV event –

were the root causes of the TMI-2

accident. Correcting the first omission

would have prevented the accident.

Correcting the second omission would

have resulted in prompt and effective

mitigation of the accident.

Lessons learned

Understanding the root causes of

the TMI accident provides valuable

guidance for nuclear power plant

designers, especially for designers

of new plant types, such as small

modular reactors. The recognition

of safety-related components and

design-specific accidents is more

complex and more difficult than it

appears to be. It is the designer’s

responsibility to identify all safetyrelated

systems and components

and to analyze all accident types.

Many systems and components

of a plant have both an operational

function and a safety function. In the

design of every system, the question

must be raised as to whether a system

or component has a safety function.

Then, if applicable, it must be

designed for both the operational

function and the safety function.

Plant response during accidents

can be abnormal and never seen

during normal operation. The plant’s

safety analysis must be complete, and

it must describe all potential plant

responses.

Designers cannot depend on

utilities’ reviews and regulatory

evaluations to correct shortcomings.

The design must be done right in the

first place, and the quality assurance

process should guarantee perfection

of the design.

Acknowledgment

I am grateful to Sheldon Trubatch for

his valuable suggestions, review of

this article, and thoughtful comments

and insight into the era from the legal

perspective surrounding the accident.

I have derived great benefit from our

stimulating discussions.

References

1. John G. Kemeny, et al.: Report of the President’s Commission on

the Accident at Three Mile Island (October 30, 1979).

2. Mitchell Rogovin, George T. Frampton Jr.: Three Mile Island: A

Report to the Commissioners and to the Public (January 1980).

Author

Zoltan R. Rosztoczy




Das neue Strahlenschutzrecht (IV) – Schutz vor Radon

Christian Raetzke

Den vorläufigen Abschluss der kleinen Reihe zum neuen Strahlenschutzrecht sollen heute die neuen Regelungen zum

Schutz vor Radon bilden. Das ist zwar ein Thema, das in der Kernenergie und Kerntechnik nicht unbedingt im

Vordergrund steht; dafür dürfte es künftig in Teilen der Wirtschaft und der Bevölkerung um so mehr für Aufmerksamkeit

sorgen und ist sicherlich von allgemeinem Interesse.

Der Schutz vor Radon wird im neuen Strahlenschutzrecht

(im Strahlenschutzgesetz – StrlSchG – vom 27.06.2017

und in der neuen Strahlenschutzverordnung – StrlSchV –

vom 29.11.2018), das an vielen Stellen und besonders

auch hier die Euratom-Grundnorm zum Strahlenschutz

(Richtlinie 2013/59/Euratom) umsetzt, einschneidend

neu und viel umfassender als bisher geregelt. In der alten

StrlSchV gab es bei „Arbeiten“ (so die damalige Terminologie),

die im Zusammenhang mit natürlich vorkommender

Radioaktivität standen, bestimmte Regelungen für

Arbeitsplätze bis hin zu einer Anzeigepflicht (§ 95 StrlSchV

a.F.); dazu gehörte in Anlage XI Teil A eine kurze Liste

„Arbeitsfelder mit erhöhten Radon-222-Expositionen“.

Die Regelungen im neuen Strahlenschutzrecht sind viel

umfassender aufgestellt. Sie betreffen nicht nur Arbeitsplätze,

also die berufliche Strahlenexposition, sondern

auch Aufenthaltsräume, also die Exposition der Bevölkerung.

Einige Regelungen gelten speziell für eine der beiden

Expositionskategorien, andere sind allgemeiner, übergreifender

Art. Am besten findet man sich wohl zurecht,

wenn man die Regelungen entsprechend in drei Gruppen

betrachtet: allgemeine, auf Arbeitsplätze bezogene und

auf Aufenthaltsräume bezogene Bestimmungen. So sind

sie auch im Gesetz und in der Verordnung eingeteilt.

Bei den allgemeinen Regelungen ist besonders die

Pflicht der zuständigen Länderbehörden zu erwähnen, bis

Ende 2020 sog. Radonvorsorgegebiete auszuweisen. Dieser

Begriff (der so im Gesetz nicht auftaucht) be zeichnet

Gebiete gem. § 121 StrlSchG, in denen die Radon-222-

Aktivitätskonzentration in der Raumluft „in einer beträchtlichen

Zahl von Gebäuden“ die Referenzwerte für Aufenthaltsräume

oder Arbeitsplätze überschreitet. Die beiden

Referenzwerte für Aufenthaltsräume und Arbeitsplätze sind

im Gesetz formal getrennt festgelegt (§§ 124 bzw. 126

StrlSchG), lauten aber übereinstimmend 300 Becquerel

(Bq) je Kubikmeter (m 3 ). Die Bestimmung der Vorsorgegebiete

erfolgt „auf Grundlage einer wissenschaftlich

basierten Methode“, die „Vorhersagen“ ermöglicht (so § 153

StrlSchV); geeig nete Daten hierfür können auf tatsächlichen

Messwerten, aber auch auf geologischen Daten, also

letztlich auf Berechnungen beruhen. Eine „beträcht liche

Zahl von Gebäuden“ soll erreicht sein, wenn auf mindestens

75 % des auszuweisenden Gebiets der Referenzwert in

min destens zehn Prozent der Gebäude überschritten wird.

Sind erst einmal diese Radonvorsorgegebiete ausgewiesen,

so ergibt sich aus § 123 StrlSchG die Pflicht, bei

Neubauten in diesen Gebieten „geeignete Maßnahmen zu

treffen, um den Zutritt von Radon aus dem Baugrund zu

verhindern oder erheblich zu erschweren“. Das gilt einheitlich

für Gebäude mit Arbeitsplätzen und/oder mit Aufenthaltsräumen,

also auch für Wohngebäude oder Bauten

für Schulen und Kindergärten. § 154 StrlSchV zählt fünf

Arten von Maßnahmen auf und lässt „mindestens eine“

davon genügen; damit soll offenkundig die Anforderung

handhabbar gemacht und eine Übersteigerung vermieden

werden. Zu den Maßnahmen zählen etwa die Verringerung

der Radon-222-Aktivitätskonzentration unter dem

Gebäude, die Begrenzung der Rissbildung in Wänden und

Böden mit Erdkontakt und die Absaugung von Radon an

Randfugen oder unter Abdichtungen.

Der Bund ist nach § 122 StrlSchG verpflichtet, einen

Radonmaßnahmenplan zu erstellen; das ist mittlerweile

geschehen (siehe www.bmu.de/publikation/radonmassnahmenplan).

Das Dokument nennt, neben den gesetzlich

geforderten Maßnahmen, auch und vor allem Informationskampagnen,

Förderung der Allgemeinbildung

zu Radon und der beruflichen Qualifikation und Weiterbildung,

Entwicklung einheitlicher Messstrategien, Vergabe

weiterer Studien und Ähnliches.

Bei Arbeitsplätzen besteht dann eine Handlungspflicht,

wenn sie entweder einem der bereits in der alten StrlSchV

genannten „Arbeitsfelder“ angehören (vor allem Berg werke

und Anlagen der Wassergewinnung und -ver teilung), und

zwar unabhängig vom geografischen Gebiet, oder wenn sie

sich im Erd- oder Kellergeschoss eines Gebäudes in einem

Radonvorsorgegebiet befinden. §§ 127 bis 131 StrlSchG

ordnen dann ein abgestuftes Verfahren an. Zeigt eine erste

Messung, die innerhalb von 18 Monaten nach Einrichten

des Arbeitsplatzes oder nach Ausweisung des Gebietes vorzunehmen

ist, eine Überschreitung des Referenzwerts,

muss der für den Arbeitsplatz Verantwortliche Maßnahmen

zur Reduzierung der Radonkonzentration ergreifen und

erneut messen. Liegt der Wert dann immer noch über dem

Referenzwert, muss er den Arbeitsplatz bei der zuständigen

Behörde anmelden und die Exposition abschätzen; vom

Ergebnis dieser Abschätzung hängen dann bestimmte Maßnahmen

des beruflichen Strahlenschutzes ab. Der Arbeitgeber

wird damit nicht zum Strahlenschutzverantwortlichen,

da es sich – in der neuen Terminologie des StrlSchG

– nicht um eine geplante Expositionssituation, sondern um

eine bestehende Expositionssituation handelt; er hat aber

teils vergleich bare Pflichten.

Für Aufenthaltsräume (also Wohnungen, Schulen,

Kinderg ärten etc.) enthalten §§ 124 und 125 StrlSchG

Regelungen, die keinen verpflichtenden Charakter haben,

sondern auf Unterrichtung der Bevölkerung und auf das

„Anregen“ von Messungen und ggf. von Schutzmaßnahmen

hinauslaufen. Wie oben erwähnt, gibt es aber für

Neubauten in Radonvorsorgegebieten, die Aufenthaltsräume

enthalten, eine Pflicht, Maßnahmen zum Schutz

vor Radon zu treffen; insofern sind nur Bestandsbauten

vorerst von verbindlichen Maßnahmen verschont.

Wie geht es weiter? Das Thema Radon steht schon seit

einigen Jahren im Fokus der Strahlenschützer, im allgemeinen

Bewusstsein dürfte es aber erst dann wirklich ankommen,

wenn ab 2021 die Radonvorsorgegebiete ausgewiesen

sind, Informationskampagnen richtig anlaufen und echte

Pflichten für viele Arbeitgeber und Bauherren entstehen. Es

steht zu hoffen, dass die neuen Regelungen dann die

gewünschte Steuerungswirkung entfalten und die findige

Bevölkerung nicht einfach nur gemäß einem Spruch handelt,

den der Verfasser von einem geschätzten Kollegen aus

dem (radonmäßig stark betroffenen) Vogtland gehört hat:

„Die Zahnarztpraxis unters Dach, die Oma in den Keller“.

Author

Rechtsanwalt Dr. Christian Raetzke,

Beethovenstr. 19, 04107 Leipzig

525

SPOTLIGHT ON NUCLEAR LAW

Spotlight on Nuclear Law

The New Radiation Protection Law – Protection Against Radon ı Christian Raetzke


526

RESEARCH AND INNOVATION

Evaluation of a Double-Ended Guillotine

LBLOCA Transient in a Generic

Three-Loops PWR-900 with TRACE Code

Coupled with DAKOTA Uncertainty

Analysis

Andrea Bersano and Fulvio Mascari

In the present study, the model of a generic three-loops PWR-900 western type reactor has been developed and a

double-ended guillotine break on the cold leg has been simulated by TRACE code. Through the SNAP graphical

interface, a DAKOTA uncertainty analysis, based on the probabilistic method to propagate input uncertainty, has been

performed by selecting uncertain parameters related to the safety injection system and to the initial plant status. In

particular, six uncertain input parameters have been considered: the accumulators’ initial pressure and temperature,

the safety injection system temperature and flow rate, the reactor initial power and the containment initial pressure.

The main figure of merit selected for the application of regression correlation is the hot rod cladding temperature. Both

Pearson and Spearman’s correlation coefficients have been computed for the cladding temperature of the hot rod to

characterize its correlation with the input uncertain parameters in the different phases of the transient. In addition, the

dispersion of the calculated data have been discussed for selected relevant thermal-hydraulic parameters, such as the

primary pressure, the core mass flow rate and the water collapsed level in the vessel.

1 Introduction

Nuclear energy is part of the energy

mix of several countries considering

its important role in the reduction of

air pollution and CO 2 emissions

caused by energy pro duction [1]. Due

to the complexity of nuclear plants,

caused by their complex geometry

where multicomponent and twophase

thermal hydraulic phenomena

take place, their thermal hydraulic behavior

is characterized with computational

tools to assess their operating

conditions and to evaluate their safety

(both for Design Basis Accidents and

Beyond Design Basis Accident) [2, 3].

The computational tools used in

the nuclear sector, also called codes,

undergo a process of Verification and

Validation (V&V) [4]. It is part of this

rigorous process the use of an experimental

“assessment database” [3];

results from Separate Effect Test

Facilities (SETF) and Integral Test

Facilities (ITF) are used to evaluate

the qualitative and quantitative code

accuracy in the prediction of the

phenomena of interest. The codes

that use more realistic information

concerning phenomena and plant

behavior are often referred to as Best

Estimate (BE) codes [5].

Though the high level of maturity

reached by BE thermal hydraulic

system codes in the last decades, in

their application there are still some

sources of uncertainty ( uncertainty is

used as a measure of the error made

with the code in predicting the plant

behavior) affecting the calculation

results [3]. In general the sources of

uncertainty can be grouped as a) code

uncertainty (e.g. approximations in

the onservation equation and in the

closure models and correlations)

b) representation uncertainty (nodalization

effect), c) scaling issue (codes

validated against scaled-down facilities),

d) plant uncertainty (e.g. initial

and boundary conditions), e) user

effect. For this reason, providing the

result of a BE calculation alone may be

not sufficient and the evaluation of

the uncertainty on the results is

required. Several methodologies have

been developed in the past to perform

Uncertainty Analysis. In particular

these uncertainty methodologies can

be grouped in a) methods to propagate

input uncertainty, divided in

probabilistic (e.g. CSAU, GRS, IPSN,

etc.) and deterministic methods (e.g.

AEAW, EDF-Framatome, etc.); and b)

method to extrapolate output uncertainty

(e.g. UMAE) [6].

In this framework, the target of the

present paper is to use the probabi listic

method to propagate the input

uncertainty in the calculation of a cold

leg double- ended guillotine break

transient of a generic three-loops PWR-

900 western type reactor with the

avail ability of active and passive Emergency

Core Cooling Systems (ECCS).

The Uncertainty Quanti fication (UQ)

application of this methodology is

based on a set of statistical techniques

for the evaluation of the number of

needed code runs and for the correlation

of the output results with the

uncertain inputs. The method has been

applied using DAKOTA toolkit [7],

developed by Sandia National Laboratories,

and has been used in the SNAP

(Symbolic Nuclear Analysis Package)

environment/architecture [8] with the

related DAKOTA uncertainty plug-in

[9]. The calculations have been performed

with TRACE (TRAC/RELAP

Advanced Computational Engine) v5

patch 4 BE thermalhydraulic system

code developed by USNRC [10]. The

hot rod cladding temperature has been

selected as a figure of merit and a

limited number of input un certainty

parameters, mostly related to plant

initial conditions, have been chosen for

the UQ application.

2 Description of probabilistic

method to propagate

input uncertainty,

Dakota Toolkit, and UQ

hyphothesis

2.1 Probabilistic method

to propagate input

uncertainty

The probabilistic method to propagate

the input uncertainty [11] is, in brief,

based on a random sampling of the

input uncertain parameters selected

by the user; a set of N code runs having

in the input a combination of the uncertain

parameters is created and

solved with the selected code. Then,

by using regression analysis, the effect

of the input parameters uncertainty

on the results is computed, in terms of

selected Figure of Merits (FOMs). The


Evaluation of a Double-Ended Guillotine LBLOCA Transient in a Generic Three-Loops PWR-900 with TRACE Code Coupled with DAKOTA Uncertainty Analysis ı Andrea Bersano and Fulvio Mascari


main advantages of this method are

that the number of code runs does not

depend on the number of input

uncertain parameters and it is not

necessary a prior development of

a Phenomena Identification and

Ranking Table (PIRT), which is a long

process. The number of code runs N

depends on the probability content

and on the confidence level set by the

user and on the number of FOMs

selected for the analysis and it is computed

using the Wilks formula [12,

13]. Each uncertain input parameter

should be defined by its range of

variation and its Probability Density

Function (PDF); this means that even

if the exact value of a parameter is

uncertain, some values are more likely

to be close to the real one than others.

This is one of the main task to be

tackled because the evaluation of the

correct PDF for each parameter is not

a trivial task and a careful study is

required. The selection of the most

suitable PDF for each parameter depends

mainly on the physical quantity

and on the availability of reducedscale

or full-scale measured data. If

data are available, it is possible to

build a histogram and to derive the

PDF. A comprehensive overview of

PDF type and derivation techniques is

provided in [14].

2.2 DAKOTA toolkit in the

SNAP environment/

architecture

DAKOTA [7] is an open source software

written in C++ and developed by

Sandia National Laboratories to perform

parametric and uncertainty

analysis in a fast and automatic way.

The aim of this toolkit is to bridge

simulation codes and analysis methods

for parametric evaluation, uncertainty

quantification and system optimization

[15]. The DAKOTA toolkit is also

provided as a plug-in [9] for the

Symbolic Nuclear Analysis Package

(SNAP), which is a graphical user

interface designed to support the use

of USNRC nuclear codes (e.g. TRACE,

RELAP, MELCOR, etc.). Using SNAP, it

is possible to build the input deck in a

graphical environment and to have a

direct visualization of the code calculated

data by using its animation

capability. Through SNAP it is possible

to set up the DAKOTA uncertainty

analysis [16, 17] and to perform automatically

all the steps qualitatively

described in the previous section.

Figure 1 shows a schematic representation

of DAKOTA uncertainty analysis

workflow in a SNAP environment/

architecture.

In particular, DAKOTA plugin allows

to:

1) Enter the uncertain input parameters

with their range and PDF;

2) Select the sampling method

( Monte Carlo or Latin Hypercube);

3) Enter the desired FOMs for the

analysis;

4) Set the final report that contains

the results of the uncertainty

analysis application; the report is

auto matically generated at the end

of the uncertainty quantification

analysis.

DAKOTA is used at the beginning of

the analysis to sample the uncertain

input parameter values and to

generate the set of code inputs. Then,

after the solution of the set of code

inputs and the extraction of the

desired data, DAKOTA performs the

uncertainty analysis and apply

regression techniques to evaluate the

correlation between input and output

parameters selected as a FOM.

The required number of code runs

can be found solving the Wilks

formula with respect to N for a probability

α and a confidence level β [11]:

(1)

With α=0.95 and β=0.95, for one

FOM the required number of code

runs is 59.

2.3 UQ application hypothesis

The target of this analyses, is not to be

a detail uncertainty study in term of

input uncertainty parameters as

presented in [18], but a) to develop a

full UQ application with TRACE and

DAKOTA toolkit in a SNAP environment/architecture

and b) to have

some insights about the degree of

corre lation between the input parameters

selected and the FOM chosen

for this analysis. Six uncertain parameters

have been selected for this

uncertainty application based also on

BEMUSE program results [18] and

through SNAP have been implemented

in the DAKOTA and TRACE

input: the Safety Injection System

(SIS) temperature, the SIS characteristic,

the accumulator initial temperature

and pressure, the initial core

power and the initial con tainment

pressure. The SIS characteristic is a

value that multiply the default injected

flow rate curve as function of the

primary pressure. Table 1 summarizes

the uncertain input parameters,

their mean value used for the reference

calculation, the range of variation

and the adopted PDF.

One FOM was selected for the

analysis, the cladding temperature of

the hot rod; therefore, with a probability

of 95% and a confidence level of

95%, a total of 59 calculations were

required based on Wilks formula as

previously described. Latin Hypercube

sampling [19,20] has been used

for this analysis. It is a stratified

sampling method that, with respect to

a pure Monte Carlo sampling, allows

to achieve the target statistical

| Fig. 1.

DAKOTA uncertainty analysis workflow for TRACE code in a SNAP environment/architecture.

RESEARCH AND INNOVATION 527




Parameter Average value Range of variation PDF type


Safety injection system (SIS) temperature [K] 285 [-10,+10] K Normal

Safety injection system (SIS) characteristic [-] 1 [0.95,1.05] Normal

Accumulator initial temperature [K] 325 [-10,+10] K Normal

Accumulator initial pressure [bar] 40.8 [-2,+2] bar Normal

Initial core power [MW] 2785 [0.98,1.02] Normal

Initial containment pressure [bar] 1.013 [0.85,1.15] Uniform

| Tab. 1.

Input uncertain parameters selected for the analysis.

| Fig. 2.

TRACE nodalization of the primary system and of the containment of the generic three-loops PWR developed with SNAP.

accuracy with a minor number of

samples. All 59 runs were correctly

executed and they reached the end of

calculation without any failure. To

analyze the transient progression, the

primary system pressure, the core

flow rate, the cladding hot rod temperature

and the vessel collapsed level

have been plotted for the complete set

of 59 runs, while the application of

regression analysis has been carried

out only for the cladding temperature.

3 Generic PWR-900 TRACE

model and steady state

operation

3.1 TRACE nodalization

description

TRACE, developed by USNRC, is a

component-oriented code designed

to perform best-estimate thermalhydraulic

analysis for LWR. It is a

finite volume, two fluid, code with 3D

capability and it is based on two fluid,

two-phase field equations. This set of

equations consists of the conservation

laws of mass, momentum and energy

for the liquid and gas phases [10, 21].

The code version adopted in this

analysis is TRACE code v5 patch 4 and

the input deck has been developed

with SNAP.

The nodalization of a generic threeloops

PWR-900, shown in Figure 2,

has been developed to perform the

large break LOCA analysis. In order to

minimize the computational time (in

view of input uncertainty propagation

with probabilistic method) and maintain

an accurate prediction of target

phenomena, the nodalization strategy

used follows the general nodalization

approach of the TRACE W4loops

samples input-deck distributed with

SNAP. Starting from that nodalization

approach sample and considering the

level of detail target of this analysis

(e.g. modeling the three loops separately,

modeling the interaction containment/primary

coolant system,

etc…), more details have been considered

in the input-deck development.

In the nodalization of the generic

three-loop PWR-900 no lumped loops

have been considered and the three

loops have been modeled separately,

one simulates the broken one (loop A)

and two simulate the intact loops

(loop B and C). The model is composed

of 69 Hydraulic Components

(HC) and 45 Heat Structures (HS).

Power is provided to the three heat

structures simulating the core through

one power component, the total

thermal power is around 2800 MW; in

one of the HS that compose the core is

inserted a supplemental rod that simulates

the hot rod in the reactor with a

total peaking factor of 2.278 [22]. The

pressurizer is connected to the hot leg

of an intact loop (loop B). The break

has been modeled with a set of three

valves; at the break opening one valve

interrupts the connection between the

two sections of the guillotine break on

the cold leg of loop A; simultaneously,

the other two valves connect the two

closed sections of the leg to the components

simulating the containment.

Particular attention has been paid

to the containment building modeling

in order to simulate the interaction

containment/primary system during

the LOCA. In particular, the containment

has been modeled with two connected

hydraulic regions thermally

coupled to the heat structure simulating

the containment solid structure

and the thermal interaction with

the environment. Figure 2 shows

the TRACE nodalization of the

generic PWR primary system and the




| Fig. 3.

TRACE nodalization of

the RPV of the generic

three-loops PWR developed

with SNAP.

| Fig. 4.

SNAP animation showing the fluid condition of the primary system during the steady state (at 900 s).


containment (the containment drawing

is scaled down with a factor 0.2 in

comparison with the other reactor

components for a better visualization

of the complete plant layout). The

ECCS are connected on the cold leg of

each loop; for each loop, the ECCS

consist of an accumulator and the

high pressure and low pressure safety

injection systems (HPIS, LPIS). The

HPIS and LPIS are modeled by a single

“fill” component with a table that

controls the injected flow rate as

function of the pressure in the primary

system. The Reactor Pressure Vessel

(RPV) has been modeled by using the

3D Vessel component available in

TRACE. The 3D vessel nodalization

has been divided into: 2 radial sectors,

the inner one for the core and the

outer one for the downcomer, 3 angular

sectors, one for each primary loop

and 7 axial sectors, 2 for the vessel

lower plenum, 3 for the core and 2

for the vessel upper plenum. The 3D

vessel nodalization representation

made with SNAP is shown in Figure 3,

with the core region highlighted.

3.2 Steady state

characterization

Before the beginning of the LOCA,

1000 s of steady state operation have

been simulated to test the nodalization

in steady state condition. The

steady state parameters are reported

in Table 2 and they have been checked

for consistency with public available

information [23, 24].

Figure 4 shows the fluid con ditions

in the primary system during the

steady state; the color legend refers to

the fluid status, ranging from blue

(subcooled liquid) to red (superheated

gas/steam). Figure 5 shows

the axial pressure profile in the

primary system in steady state conditions,

to support the steady state

qualification of the nodalization. The

normalized pressure profile is consistent

with other similar data available

in the public scientific literature

[25].

4 Cold leg lbloca transient

and uncertainty analysis

4.1 LBLOCA transient analysis

The analyzed transient is a doubleended

guillotine break (200 %) in the

cold leg of loop A. After 1000 s of

steady state simulation the break is

opened (start of the transient: t = 0 s)

and 500 s of transient are simulated.

The reactor SCRAM is supposed after

0.5 s from the break. From Figure 6 to

Figure 9 the results for the 59 runs are

shown for the primary system pressure,

the core flow rate, the cladding

hot rod temperature and the vessel

collapsed level. Figure 6 shows the

behavior of the primary system pressure;

after the LOCA initiation, water

flows from the primary system to the

containment and the primary system

pressure drops significantly in few

tens of milliseconds in agreement

with the publically available technical

scientific literature [26]; after this

first drop of pressure and the phase

of subcooled depressurization, the

saturation con dition is reached and

saturated depressurization starts at a

reduced rate. The maximum flow rate

through the break is limited by the

critical velocity at the break. After the

transient initiation, the flow rate in

the core (Figure 7) drops from the

nominal value and it is reversed since

the flow is directed to the break

location; therefore, water flows downwards

in the core region and then

upward in the downcomer to reach

the break in the cold leg.

The high voiding in the core and

the subsequent SCRAM stop the chain

reaction and, due to a sensible

reduction of heat removal in the

core, the heat stored in the fuel is

Thermal power [MW] 2785

Primary system pressure [bar] 155

Total core flow rate [kg/s] 13,947

Core inlet temperature [K] 558.4

Core outlet temperature [K] 594.1

Secondary system pressure [bar] 58

Steam generator feedwater temperature [K] 440

Steam generator feedwater flow rate [kg/s] 512

| Tab. 2.

Steady state parameters of the reference calculation.

| Tab. 5.

Normalized axial pressure profile along the primary system in steady state

condition (at 900 s).





redistributed, leading to the first

cladding temperature peak, as shown

in Figure 8. The cladding temperature

passes from the steady state value

(around 617 K) to 955 K. The result

of the core voiding is a sudden drop

of the water collapsed level in the

reactor vessel, as shown in Figure 9,

under the Bottom of Active Fuel

(BAF). When the pressure in the

primary system is lower than the

accumulator initial pressure, the

accumulator check valves open and

water is discharged in the primary

system; this happens around 11 s after

the transient initiation. The water

initially injected in the cold legs by the

accumulators bypasses the vessel

lower plenum through the upper

down comer region and it is directed to

the break without penetrating the

core. After the first fast depressurization

phase, the primary pressure

continues to reduce at a lower rate,

equalizing the containment pressure

(around 0.45 MPa after around 40 s)

and ending the blowdown phase.

Water is injected in the primary

system by the accumulators and the

low pressure safety injection system

start to inject water after around 30 s

from the break initiation. The refill

phase starts around 40 s, when the

emergency core coolant water reaches

the vessel lower plenum and the

collapsed level in the vessel starts to

rise. During this phase the core is

mainly uncovered and heat is not

removed from the fuel rods, with the

exception of a small amount of heat

removed by thermal radiation and

natural convection of the steam

present in the core. For this reason

during the refill period the cladding

temperature increases (Figure 8) due

to the quasi adiabatic heating of

fission product decay. When, around

50 s from the LOCA initiation, the

water level reaches the core bottom

(Figure 9) the refill period ends

and the reflood phase starts. Water

collapsed level rises quickly up to

around 65 s (time of end of accumulators

injection), and it continues at a

lower rate due to the LPIS. In this

phase the net core flow rate is positive,

even if very small and with many

oscillations. The water entering the

core is heated up, starts to boil and

entrains water droplets that help the

cooling of the hottest parts of the core.

With the rising of the water level, the

cooling is increased and the cladding

temperature starts to decrease. The

complete rewetting of the cladding

surface caused by the rising water

level produces a strong temperature

drop (core quenching). This happens

around 125 s after the beginning of

the LOCA.

Analyzing the dispersion of the

results, the primary pressure (Figure

6) presents an almost negligible

dispersion during the blowdown

phase and the predictions of the 59

runs are very similar; after the blowdown

the dispersion band width is

always lower than 0.1 MPa with a final

average value of 0.475 MPa. As

regards the core mass flow rate

(Figure 7), cladding temperature

(Figure 8) and vessel water collapsed

level (Figure 9), the results dispersion

is very limited during the blowdown

phase, while it is more noticeable in

the refill phase, especially for the

vessel water collapsed level, and it is

higher during the reflood phase.

In particular, the refill initial time

shows a dispersion band width of 12 s

(33 – 45 s); during the reflood the

collapsed level dispersion band width

is around 1 m and the Top of Active

Fuel (TAF) is reached in a time band

of 33 s (126 – 159 s). The cladding

temperature has, at the first peak, a

low dispersion band width of 14 K

and the peak has almost the same

timing for all runs; instead, the dispersion

band width is higher for the

second peak (50 K) and with a time

dispersion band width of 10 s. The hot

rod cladding quenching time is also

affected by a dispersion band width

of 20 s (115 – 135 s).

| Tab. 6.

Primary system pressure predicted by TRACE code.

| Tab. 7.

Core mass flow rate predicted by TRACE code.

| Tab. 8.

Cladding temperature of the hot rod predicted by TRACE code.

| Tab. 9.

Reactor vessel water collapsed level predicted by TRACE code.




| Tab. 10.

Pearson’s simple regression correlation coefficients.

4.2 Uncertainty analysis

response correlations

4.2.1 Description of regression

correlations in DAKOTA

As a result of the uncertainty analysis,

DAKOTA [7, 8, 20] computes four

response correlation coefficients:

simple, partial, simple rank and

partial rank. The simple coefficient is

related to the actual input and output

data. The simple coefficient r between

an input variable x and an output

variable y, in n samples, is computed

using the Pearson’s correlation. It is a

measure of the degree of linear correlation

between the two variables and

its value is comprised between -1 and

1. If r0 the

correlation is positive (an increment

of x leads to an increment of y). The

simple correlation coefficient between

x and y is obtained by dividing the

covariance of the two variables by the

product of their standard deviations

[27]:

(2)

The partial correlation coefficient is

computed similarly to simple one but

taking into accounts the effects of

the other variables. This is useful, for

example, if there is a strong correlation

between two inputs; in this way

the correlation of the second input on

the output may be adjusted after

having considered the correlation between

the first input and the output

[7]. Rank correlation coefficients use

the ranked data instead of the actual

ones. Ranks are obtained by ordering

the data in ascending order, and are

more convenient to be used when

inputs and outputs are characterized

by sensible difference in magnitude; it

is possible to understand if the input

sample with the lower rank is associated

to the output with the lower

rank and so on [7, 20]. To compute the

rank correlation, DAKOTA uses the

Spearman’s rank correlation that is

similar to Pearson’s one but with the

ranked data instead of the actual

values. If two variables are monotonically

related, without repetitions,

the Spearman coefficient is -1 or +1

(depending if the function is monotonically

decreasing or increasing),

since the ranked values are used.

Moreover, Spearman’s correlation is

less sensitive to possible outlier values

of the variables than Pearson’s one.

4.2.2 Results of response correlation

coefficients for the

cold leg LBLOCA transient

The time dependent computation of

response correlation coefficients has

been performed extracting the FOM

value at different selected instant of

the transient evolution. Figure 10

shows the Pearson’s simple correlation

coefficient for the six input

uncertain parameters from the LOCA

beginning to the complete core

quenching (after 130 s). Figure 11

shows the Spearman’s rank correlation

coefficient for the same parameters.

On both graphs the values 0.2

and 0.5 (and -0.2 and -0.5) have been

highlighted as measure of the correlation

between the input parameter

and the FOM. As indicated in the [20],

for the Spearman coefficient, if the

coefficient is higher than 0.5 (or lower

than -0.5) the correlation is significant,

if it is between 0.2 and 0.5 (or

-0.2 and -0.5) the correlation is

moderate, otherwise it is low [20]. In

| Tab. 11.

Spearman’s simple rank regression correlation coefficients.

the following analysis the same

threshold values have been adopted

also for the Pearson coefficient. In this

application, the two response correlations

show similar trends; the main

advantage of this time dependent

analysis is the possibility to have a

measure and characterize the correlation

of the different input parameters

on the uncertainty of the

selected FOM in all the phases of the

transient.

In the blowdown phase (0-40 s) the

effect of the initial core power on the

hot rod cladding temperature is

positive due to the heat stored in the

solid structures. Both Pearson’s and

Spearman’s coefficients are higher

than 0.8 for this parameter so the

correlation is significant and almost

monotonically linear. In the remaining

part of the transient both coefficients

are close to 0.2 so the corre lation is

much weaker than in the initial phase.

After around 20 s from the LOCA

initiation the accumulators initial

pressure has a negative Pearson’s and

Spearman’s coefficients around -0.3;

therefore the lower is the accumulator

pressure the higher is the cladding

temperature since the accumulator

starts to inject water later. In the refill

phase (40-50 s), also the SIS characteristic

has both response correlation

coefficients around -0.3; in fact, with a

lower injection flow rate the hot rod

cladding temperature is higher. In the

reflood phase (after 50 s) also the

initial containment pressure has negative

Pearson’s and Spearman’s coefficients

around -0.35. This is due to the

fact that with a lower pressure in the

containment a greater amount of

coolant water is expelled through

the break and the cladding temperature

is higher. The uncertainty in the

accumulators and SIS temperature

have a low correlation with the FOM.






5 Conclusions

Uncertainty analysis is used in the

nuclear sector to evaluate uncertainties

still present in the results of code’s

calculations. Among the several

methodologies developed in the past

to perform Uncertainty Analysis, the

probabi listic method to propagate the

input uncertainties has been selected

for this analysis considering its

suitability to be coupled with simulation

codes; moreover, several

toolkits, some integrated in computational

platforms, have been developed

for this purpose. In the present

activity, a double-ended cold leg

LBLOCA transient has been simulated

with the BE thermal hydraulic system

code TRACE for a generic three-loops

PWR-900 reactor. Using DAKOTA

toolkit, in a SNAP environment/architecture,

an uncertainty analysis has

been carried out by selecting six

uncertain input parameters and the

hot rod cladding temperature as the

main figure of merit. In addition, the

primary pressure, the core flow rate

and the pressure vessel collapsed level

have been analyzed to evaluate the

transient progression and the results

dispersion. The aim of this analysis is

not to be a detail and exhaustive

uncertainty study in term of input

uncertainty parameters but to develop

a complete uncertainty quantification

application with DAKOTA in a SNAP

environment/architecture and to have

some insights characterizing the

correlation between the input uncertainty

parameters and the selected

FOM. Pearson’s and Spearman’s

response correlation coefficients have

been computed between the LOCA

initiation and the complete core

quenching. In the blowdown phase,

the hot rod cladding temperature has

a significant correlation with the

initial core power; the accumulators’

initial pressure has a moderate correlation

with the FOM only in the period

of water injection from the accumulators.

The time dependent response

analysis, adopted in this application,

is very useful since it could be used to

characterize the effect of the uncertain

input parameters on the output

global uncertainty in the different

phases of a transient. In a future

follow-up, additional uncertain input

parameters and FOMs can be introduced

in the analysis in order to have

a more complete evaluation of the

results uncertainty.

Acknowledgement

The authors are grateful to Ms Cristina

Bertani for the review of the manuscript.

Nomenclature

N

n

p

r

x

y

α

β

Subscripts

m

Number of code runs

Number of samples of a certain variable

Number of Figure of Merit

Pearson’s correlation coefficient

Input variable in the computation of response correlations

Output variable in the computation of response correlations

Probability

Confidence level

mean value of the related parameter

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Authors

Andrea Bersano

Energy Department (DENERG)

Politecnico di Torino

Corso Duca degli Abruzzi 24

10129, Turin

Italy

Fulvio Mascari

Nuclear Safety

Security and Sustainability Division

(FSN-SICNUC), ENEA

Via Martiri di Monte Sole 4

40129, Bologna

Italy




Experiment Research on the Insurge

Transient Behavior of Gas-steam

Pressurizer under Various Pressure

Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun

For small and medium sized reactors with gas-steam pressurizer, the transient behavior of gas-steam pressurizer plays

a vital role on the safety of the nuclear reactor operation. This paper focuses on the transient behavior of gas-steam

pressurizer. The pressure response was investigated under various insurge experimental conditions, and the influence

factors of pressure changes were analyzed.

Research shows initial pressure of the pressurizer and the presence of a non-condensable gas with varying mass

fraction will all have some kind of effect on the transient behavior of gas-steam pressurizer. Initial pressure decides initial

magnitude of liquid temperature of the pressurizer, and the presence of a non-condensable gas with varying mass

fraction can greatly affect the heat and mass transfer process both wall and interface, and further affect the system

pressure variations. This paper focus on the effects on pressurizer insurge transients under various initial pressure of the

pressurizer and the presence of a non-condensable gas with varying mass fraction.

Introduction

A transient can be caused from a

simple loss of secondary steam flow to

more complicated accidents. After

the Three Mile Island accident [1],

Chernobyl accident [2] and Fukushima

nuclear accident [3]. Many people

realized the importance of small break

LOCA’s and the necessity for having

reliable physical models for all of the

components in the loop.

In view of increasing pressure on

energy restructuring and the serious

environmental pollution concerns,

the integrated natural circulation

reactor is receiving a great attention

for its ability to provide energy that is

clean, safe and economic. And the integrated

reactor is going to be one of

the best option for small and medium

sized reactors (SMRs) in lots of countries

with the high level of great safety

and reliability. SMRs have been developed

in many countries for small-scale

power generation, district heating,

and seawater desalination. Argentine

CAREM reactor, Russian VVER-300/

VK-300 reactor, Korean REX-10/

SMART reactor, Japanese IMR reactor,

American NuScale reactor, Chinese

NHR reactor and so on, these are the

typical representatives of the SMRs

[4]. In order to simplify the structure

and design and enhance safety, the

steam-gas pressurizer is generally

utilized in the integrated reactor. The

non-condensable gas is used to keep

the pressure stable in the steam-gas

pressurizer and the transient behavior

of gas-steam pressurizer plays a vital

role on the safety of the nuclear reactor

operation. As so far, early pressurizer

transient models were generally

developed under certain conditions,

slow insurge velocity or low pressure

for example. And these models may

not be applicable to various velocity or

create a greater risk of inaccurate

results under the high pressure.

Westinghouse [5] developed the

TOPS pressurizer model included

the effects of wall condensation by applying

Nusselt Laminar film theory

to estimate a wall heat transfer

coefficient. Saedi [6] investigated the

relative magnitudes of the physical associated

with insurge transients, and

initiated a data base (at low pressure)

for a model, developed

by Kim [7], at MIT. In some other

pressurizer analysis, Mark [8] came

up with the effects of the presence of a

non-condensable gas on insurge

transient. Leonard [9] performed the

experiment on the pressure behavior

of steam-gas pressurizer during the

insurge and he focused on the various

non-condensable gas, various gas

content and stratification in his study.

Kim [10] found the condensation heat

transfer at wall is an important physical

phenomenon during the insurge

transient. Paulsen [11] improved and

developed the theoretical modeling

for RELAP5. Wu lei [12, 13] and Ma

Xizhen [14, 15] established the nonequilibrium

gas-steam pressurizer

model and improved the steam condensation

heat transfer model in

presence of non-condensable gas.

In order to understand and model

an accident, one should recognize the

processes that take place during a

transient. In general, these processes

include insurges, outsurges, and combined

insurges and outsurges. Early

pressurizer transient models were

generally developed under certain

conditions, slow insurge flow rate or

low pressure for example, and these

models may not be applicable under

various pressure. At present, the

transient behavior of steam-gas pressurizer

under various pressure need to

be further studied in the integrated

natural circulation reactor with

steam-gas pressurizer. Through the

design and establish of the experiment

system, the combination of

theoretical and experimental which

can provide the data support for the

design and operation of the integrated

natural circulation reactor.

1 Description of

Experimental System

To find out the experimental physical

processes occurs in a pressurizer, an

experimental apparatus has been

built (Figure 1).

The primary tank (2420 mm high

and 450 mm ID) which models the

pressurizer volume equipped with a

magnetic level gauge to accurately

measure the initial liquid level and

the change of liquid level during the

insurge. The storage tank (2490 mm

high and 400 mm ID) is filled with

cold water to a level also measured

with a magnetic level gauge and

pressurized with nitrogen. This tank

| Fig. 1.

Schematic diagram of apparatus.

Nomenclature

p t system pressure

T w temperature at each

measured point of gasphase

space

p N2

nitrogen partial pressure

p s steam partial pressure

n N2

nitrogen amount of

substance

n s steam amount of substanc

V g gas-phase space volume

Z N2

nitrogen compressibility

factor

Z s

steam compressibility

factor

x N2

nitrogen mole fraction

T g average temperature in

gas-phase space



Experiment Research on the Insurge Transient Behavior of Gas-steam Pressurizer under Various Pressure ı Wang Bolong, Li Weihua, Jia Haijun, Li Jun and Zhang Yajun



| Fig. 2.

Heating rod installation method and voltage regulator.

| Fig. 3.

Data acquisition system and computer data network.

serves as a reservoir for “cold” injection

water.

Two tanks are both equipped with

nine heating rods totaling 36 kW and

the electric heating system utilizes the

TSGC2J-40 type voltage regulator to

adjust the load voltage of the heating

rod, and then control the heating rod

power to meet the requirements

of each experimental condition

(Figure 2).

Besides, the experimental system

also consists of electric heating system

and data measurement and acquisition

system (Figure 3).

2 Experimental conditions

and parameter control

Several precautionary procedures were

listed as followed prior to initiating

each of the transient experiments. The

primary and storage tanks were filled

to the required levels with deionized

water. Steam was bubbled through the

primary tank to assist the heating rods

in bringing the tank up to saturation

conditions. The relief valve was opened

and the water allowed to boil for a

while when the primary tank reached

saturation. This process was intended

to rid the system of dissolved gases in

the water. For the experiments with an

initial non-condensable gas fraction in

the vapor, the gas was injected into the

vapor space after the tank had been

purged of dissolved gases, and before

the system had reached the desired

operating pressure.

The partial pressure of the steam

and the temperature of gas-phase

space, which are directly determined

by nitrogen mole fraction in the

pressurizer. Therefore, the experiment

should put strict controls on its

initial value. Based on the real gas

equations, the nitrogen mole fraction

can be obtained by measuring the

temperature of gas-phase space and

system pressure. It should be noted

that the calculation leaves out of

account stratification of hot gas and

steam superheating phenomenon.

The concrete calculating methods are

as follows: measurement parameters:

system pressure p t ,the temperature at

each measured point of gas-phase

space in the pressurizer T w1 , T w2 … T wn .

Based on the real gas equations,

the nitrogen mole fraction can be

obtained by measuring the temperature

of gas-phase space and system

pressure. It should be noted that the

calculation leaves out of account

stratification of hot gas and steam

superheating phenomenon. The

concrete calculating methods are as

follows:

Measurement parameters: system

pressure, the temperature at each

measured point of gas-phase space in

the pressurizer T w1 , T w2 … T wn .

Middle parameters: nitrogen

partial pressure p N2

, steam partial

pressure p s , nitrogen amount of

substance n N2

, steam amount of substance

n s , gas-phase space volume V g ,

nitrogen compressibility factor Z N2

,

steam compressibility factor Z s .

The average temperature T g in gasphase

space can be obtained by

measuring the temperature at each

measured point of gas-phase space

in the pressurizer.

The real gas equations of steam and

nitrogen are as follows where R indicates

gas constant:

Then, it can come to the nitrogen mole

fraction X n2

.

=

| Fig. 4.

Computing framework.

The computing framework is demonstrated

by Figure 4.




Parameter Value Unit

Non-condensable gas species Nitrogen /

Non-condensable gas mass fraction 0, 10%, 20% /

Regulator space initial pressure 0.5, 1.0, 1.5 MPa

Initial height of liquid level 800 mm

Final height of liquid level 1200 mm

Insurge time 20 s

| Tab. 1.

Specific experimental conditions.

The data acquisition system was

initiated prior to beginning the

insurge to allow some steady data to

be taken, the specific experimental

conditions are shown in Table 1.

3 Experimental results and

analysis

There are two main reasons affected

the pressure in a transient experiment,

the first part is the increase of

liquid level plays a role that compresses

gas-phase space and the

pressure is going up, the second part is

a higher initial pressure leading to a

higher initial liquid temperature of

the pressurizer and affect the heat and

mass transfer process both wall and

interface, and further affect the system

pressure variations. The pressure

variation is a combination of the two

aspects.

The pressure and temperature

histories for the experiments with

20 % mass fraction nitrogen are

shown in Figure 5 and Figure 6. To

make conclusions of trials more

comparable, this analysis applies

dimensionless method to the system

pressure, and pressure variation

was represented by the ratio of

system pressure to initial pressure.

Figure 7 shows the dynamic change

process of dimensionless pressure

corresponding to different initial

pressures.

The system pressure growth rate

and the initial pressure of the pressurizer

have positive correlation. The

higher the initial pressure, the faster

the system pressure growth rate in the

transient.

The presence of a non-condensable

gas with varying mass fraction will all

have some kind of effect on the

transient behavior of gas-steam

pressurizer. At the same system pressure,

the main effects of the presence

of a non-condensable gas are given as

follows:

I

Affect the nitrogen partial pressure,

the steam partial pressure

and the vapor temperature of the

pressurizer.

II Affect the compressibility of the

vapor space in the pressurizer.

III Affect the heat and mass transfer

process of the vapor space in the

pressurizer.

The pressure histories for the experiments

with various mass fraction

nitrogen are shown in Figure 8,

Figure 9 and Figure 10. The initial

primary tank non-condensable gas

mass fraction for the transient is

labeled in the figure. The most obvious

difference in the results is the large

variation in peak pressure, the higher

initial non-condensable gas mass

fraction in the primary tank, the

higher the peak pressure in the

transient. And the specific values are

shown in Table 2.


| Fig. 5.

The system pressure histories with 20 % mass fraction nitrogen.

| Fig. 6.

The temperature histories with 20 % mass fraction nitrogen.

| Fig. 7.

The dimensionless pressure histories with 20 % mass fraction nitrogen.

| Fig. 8.

The pressure histories with various mass fraction nitrogen under 0.5MPa pressure.





Initial

system pressure

| Tab. 2.

Specific experimental values.

| Fig. 9.

The pressure histories with various mass fraction nitrogen under 1.0 MPa

pressure.

Non-condensable gas

mass fraction

Peak

pressure

0.5 0 0.55

0.5 10 % 0.59

0.5 20 % 0.61

1.0 0 1.08

1.0 10 % 1.21

1.0 20 % 1.25

1.5 0 1.69

1.5 10 % 1.87

1.5 20 % 1.92

In particular, the research demonstrated

that when the pressurizer is

free from non-condensable gas,

may present the twice peak pressure

phenomenon during the transient,

and the first peak pressure will present

about the same time. While only

one peak pressure will present during

the transient containing the noncondensable

gas.

There are two main reasons

affected the pressure in a transient

experiment, the first part is the increase

of liquid level plays a role that

compresses gas-phase space and the

pressure is going up, the second part is

steam condensation leading to a decrease

pressure at some degrees. The

peak pressure is combination of the

two aspects. In the early stage of the

transient, the compression of the

vapor space plays leading roles and

the pressure is going up. At this stage,

there was no distinct effect on the temperature

of the interface by incoming

the cold water during the transient. In

the second stage, steam condensation

leads to a decrease pressure at some

degrees due to the temperature reduction

of the interface. And in next stage,

the compression of the vapor space

also plays leading roles comparing

with the steam condensation and the

pressure is going up.

Conclusion

Based on the proceeding analysis and

experiments, an experimental study

of the pressure response during an

insurge transient under various pressure

has been performed. The following

conclusions are drawn based upon

the experimental and analytical

results:

I

The system pressure growth rate

and the initial pressure of the pressurizer

have positive correlation.

The higher the initial pressure, the

faster the system pressure growth

rate in the transient.

II The changes in system temperature

coincided well with those of

system pressure.

III The higher initial non-condensable

gas mass fraction in the primary

tank, the higher the peak pressure

in the transient.

IV In particular, when the pressurizer

is free from non-condensable gas,

may present the twice peak pressure

phenomenon during the transient,

and the first peak pressure

will present about the same time.

While only one peak pressure will

present during the transient containing

the non-condensable gas.

Acknowledgments

Special thanks should go to Mr Ma

Xizhen who have put considerable

time and effort into the designing and

constructing work of the experiment

plant.

References

[1]. Bot P L. Human reliability data, human error and accident

models—illustration through the Three Mile Island accident

analysis[J]. Reliability Engineering & System Safety, 2004,

83(2):153-167.

[2]. Jr H T P. Summary report on the post-accident review meeting

on the chernobyl accident : International Nuclear Safety Advisory

Group. International Atomic Energy Agency (IAEA) Safety Series

No. 75-INSAG-1 (STI/PUB/ 740), Vienna, IAEA, 1986. 260 Austrian

Schillings[J]. Journal of Environmental Radioactivity, 1987,

5(5):403-404.

| Fig. 10.

The pressure histories with various mass fraction nitrogen under 1.5 MPa

pressure.

[3]. Kinoshita N, Sueki K, Sasa K, et al. Assessment of individual

radionuclide distributions from the Fukushima nuclear accident

covering central-east Japan[J]. Proceedings of the National

Academy of Sciences of the United States of America, 2011,

108(49):19526-19529.

[4]. International Atomic Energy Agency: Advances in small modular

reactor technology developments. Vienna: the IAEA in Austria,2014.

[5]. Redfield, J. A., Prescop, V., & Margolis, S. G. (1968). Pressurizer

performance during loss-of-load tests at shippingport: analysis

and test. , 4(3), 173-181.

[6]. Saedi, H. R. (1982). Insurge pressure response and heat

transfer for PWR pressurizer. Massachusetts Institute of Technology.

[7]. Kim, S. N. (1984). An experimental and analytical model of a

pwr pressurizer during transients. British Journal of Surgery,

87(12), 1615–1616.

[8]. Leonard, M. T., & Griffith, P. (1983). The effects of a

noncondensable gas on pressurizer insurge transients. Trans.

Am. Nucl. Soc.; (United States), 46(6), 844-845. NOMURA Katsuya,

K.S.O.Y., Numerical analysis of droplet breakup behavior using

particle method. 1999. 38(12): p. 1057-1064.

[9]. Leonard M T, Griffith P. The effects of a non-condensable gas

on pressurizer insurge transients[J]. Trans. Am. Nucl. Soc.; (United

States), 1984, 46(6):844-845.

[10]. Kim, S.N., Griffith, P., 1987. PWR pressurizer modeling. Nucl.

Eng. Des.102, 199–209.

[11]. Paulsen, M.P., et al., 1996. RETRAN-3D—a program for transient

thermal–hydraulic analysis of complex fluid flow systems.

Electric Power Research Institute, NP-7450.

[12]. WU Lei JIA Hai-jun LIU Yang MA Xi-zhen. Transient Characteristics

of Integrated Non-condensable Gas-steam Pressurizer [J],

Atomic Energy Science and Technology, 2014, 48(s1):200-207.

[13]. WU Lei LIU Yang JIA Hai-jun YANG Xing-tuan. Research on

Steam Condensation Heat Transfer Model in Presence of Noncondensable

Gas at High Pressure [J]. Atomic Energy Science and

Technology, 2016, 50(2):261-266.

[14]. MA Xi-zhen JIA Hai-jun LIU Yang WU Lei. Numerical

Simulation Study on Steam Condensation in Presence of Noncondensable

Gas [J]. Atomic Energy Science and Technology,

2015, 49(s1):265-269.

[15]. MA Xi-zhen JIA Hai-jun LIU Yang WU Lei. Effect of Noncondensable

Gas on Steam-gas Pressurizer [J]. Atomic Energy

Science and Technology, 2016, 50(9):1586-1591.

Authors

Wang Bolong

Li Weihua

Jia Haijun

Li Jun

Zhang Yajun

Institute of Nuclear and New Energy

Technology

Tsinghua University

Beijing 100084, China

Collaborative Innovation Center of

Advanced Nuclear Energy Technology

Tsinghua University


Key Laboratory of Advanced Reactor

Engineering and Safety of Ministry of

Education

Tsinghua University





Stilllegung und Rückbau des

Rossendorfer Forschungsreaktors RFR

Teil 1: Objektbeschreibung, Genehmigungsverfahren, Ausgangssituation,

Planungskonzept und Meilensteine

Reinhard Knappik, Klaus Geyer, Sven Jansen und Cornelia Graetz

Mit der Entlassung des Rossendorfer Forschungsreaktors im September 2019 aus dem Geltungsbereich des Atomgesetzes

(AtG) sind die Stilllegung und der Rückbau der alten kerntechnischen Anlagen des ehemaligen Zentralinstituts für Kernforschung

(ZfK) der Akademie der Wissenschaften der DDR bis auf ein ca. 50 m langes Reststück einer Rohrleitung der

Speziellen Kanalisation abgeschlossen.

In der zweiteiligen Veröffentlichung werden die Arbeiten zur Stilllegung und zum Rückbau des Rossendorfer

Forschungsreaktors (RFR) übersichtsartig vorgestellt, wobei im Teil 1 (atw 11-12 2019) anknüpfend an die Objektbeschreibung

die Genehmigungsverfahren, die Ausgangssituation radiologisch und konventionell, das realisierte Planungskonzept

sowie die Meilensteine vor gestellt werden. Im Teil 2 (atw 1/2020) wird auf ausgewählte Aspekte der Stilllegung- und

Rückbaudurchführung, zum Strahlenschutz, zur Freigabe sowie zum Reststoff- und Abfallmanagement eingegangen.

1 Einleitung Die Geschichte des Rossendorfer Forschungsreaktors und damit des Forschungsstandortes

Rossendorf bei Dresden (jetzt Dresden-Rossendorf) begann mit dem „Abkommen über die Hilfeleistung der Union der

Sozialistischen Sowjetrepubliken an die Deutsche Demokratische Republik bei der Entwicklung der Forschung auf dem

Gebiet der Physik des Atomkerns und der Nutzung der Kernenergie für die Bedürfnisse der Volkswirtschaft“, das am

28. April 1955 abgeschlossen wurde. Auf dieser Grundlage erfolgte die Lieferung zweier Großgeräte, ein 2 MW

Forschungsreaktor vom Typ WWR-S (Abb. 1) und ein 25 MeV-Zyklotron, an das am 1. Januar 1956 gegründete

Zentral institut für Kernphysik (später Zentralinstitut für Kernforschung).

Am 14. Dezember 1957 wurde der

erste Forschungsreaktor der DDR

nach einer Bauzeit von nur 21

Monaten erstmals kritisch und am

16. Dezember 1957 offiziell im Rahmen

eines Staatsaktes in Betrieb

genommen. Der RFR war ein leichtwassermoderierter

und -gekühlter

Tankreaktor. Über einen Zwischenschritt

(von 2 MW auf 4 MW und

5 MW im Jahre 1965) erfolgte zuletzt

eine Leistungs erhöhung auf 10 MW

(1967), die ab 1981 im Dauerbetrieb

realisiert wurde. Die Leistungserhöhung

auf 10 MW konnte nur durch

Einsatz neuer Brennstäbe mit einer

Anreicherung von 36 % U-235 gegenüber

vorher von 10 % U-235 erreicht

werden. Die Energieabgabe des RFR

betrug insgesamt rund 28.000 MWd

bei einem Leistungseinsatz von rund

105.000 Stunden. Es gab während der

Betriebszeit des RFR kein Ereignis,

welches strahlenschutzrelevant war.

Interessante Details über die RFR-

Betriebszeit enthält das Buch „Beiträge

zur Geschichte des Rossendorfer

Forschungsreaktors RFR“ [1].

Im Zuge der weiteren Entwicklung

des Forschungsstandortes entstanden

eine Vielzahl von weiteren Anlagen

und Einrichtungen in denen eine kerntechnische

Nutzung bis 1991 erfolgte,

wobei hier der RFR nur bis zum

27. Juni 1991 betrieben wurde. Mit

der Neuordnung des Forschungsstandortes

nach der Wiedervereinigung und

der Auflösung der Akademie der

Wissenschaften zum 31. Dezember

1991 wurden dann auf der Grundlage

mehrerer in den Jahren 1993 und

1996 vollzogener Kabinettsbeschlüsse

mit der Stilllegung und dem Rückbau

der kerntechnischen Anlagen am

Forschungsstandort begonnen. Der

Freistaat Sachsen beauftragte den

Verein für Kern verfahrenstechnik und

Analytik Rossendorf e. V. (VKTA, Umbenennung

in „VKTA – Strahlenschutz,

Analytik & Entsorgung Rossendorf

e. V.” im Dezember 2014) mit der vollständigen

Beseitigung der nuklearen

Altlasten des Forschungsstandortes

Rossendorf. Im Folgenden werden die

wichtigsten Rückbauobjekte mit

Termin der Ent lassung (in Klammern)

aus dem Geltungsbereich des Atomgesetzes

aufgeführt:

p die Rossendorfer Anordnung für

kritische Experimente (RAKE,

1998)

p das Urantechnikum (2000)

p der Rossendorfer Ringzonenreaktor

(RRR, 2000)

p die Spezielle Kanalisation am

Forschungsstandort (AtG-Teilentlassungen

2010, 2013 und 2018)

p die Objekte und Anlagen der Isotopenproduktion

(von 2009 bis

2014, [2])

p die Anlagen und das Gelände der

Entsorgungs wirtschaft des ehemaligen

ZfK (2005, 2011 und 2018)

p der RFR (2009, 2010, 2019)

| Abb. 1.

Ansicht des Forschungsreaktors während der Fertigstellungsphase.

Die Kosten für Stilllegung, Rückbau

und Entsorgung werden vollständig

vom Freistaat Sachsen getragen,

sodass die Mittelbereitstellung auch

Einfluss auf die terminlichen Arbeitsabläufe

hatte. So mussten in den

Jahren 2005 und 2006 die Rückbauarbeiten

aufgrund der zu knapp bemessenen

Finanzmittel durch den

Freistaat Sachsen eingestellt werden,

was zu entsprechend längeren Rückbauzeiten

führte, da z. B. Mitarbeiter

nicht mehr zur Verfügung standen.

2 Objektbeschreibung

Zu den RFR-Anlagen (Abb. 2) gehörten

der „Rückbau komplex RFR“

mit einer Fläche von ca. 9.000 m 2 und

den Objekten:

p Labortrakt (1) mit Reaktorwarte

(2)

537

DECOMMISSIONING AND WASTE MANAGEMENT


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


DECOMMISSIONING AND WASTE MANAGEMENT 538

| Abb. 2.

Territoriale Lage der RFR-Anlagen.

p Reaktorhalle mit Vorhalle und

Anbau (3)

p Pavillon und einen weiteren

Anbau (4)

p Ventilations- und Filtergebäude

mit Fortluftschornstein (5)

p Schauer (6)

sowie die außerhalb des Komplexes

liegenden Objekte:

p Trafostation (7)

p Notstromgebäude (8)

p altes Armaturenhaus (9)

p Armaturenhaus (10)

p Trockenkühlturm 1 und 2 (11)

p Rohrleitungen des 2. Kühlkreislaufes

(12)

Weiterhin gab es außerhalb noch die

Sammelbehälter anlage mit der Pumpbedienstation,

in der die radioaktiven

Abwässer des RFR aufgefangen

wurden. Diese Anlagen befanden sich

im Kontrollbereich der Entsorgungswirtschaft

und wurden im Rahmen

dieses Rückbaukomplexes aus der

atomrechtlichen Aufsicht entlassen.

Die territoriale Lage der nummerierten

RFR-Anlagen ist schematisch in der

Abb. 2 dargestellt. Im Betriebshof befanden

sich befestigte Verkehrsflächen,

Rohre der Regen- und Schmutzwasserkanalisation

sowie Leitungen für

konta minationsverdächtige Abwässer

bis zu einer Tiefe von 5 m, unterirdische

Abluftleitungen, Heizungskanäle,

Schächte sowie Medienleitungen.

Im Folgenden werden wichtige

Objekte kurz charakterisiert:

Labortrakt mit Reaktorwarte

(1 und 2)

Der Labortrakt war ein viergeschossiger

Ziegelbau mit Stahlbetondecken,

dessen Grundfläche ca. 960 m 2

betrug. In diesem Gebäude waren

ursprünglich Laboratorien, die Reaktorwarte

und Büroräume untergebracht.

Im Zuge der Erneuerung

des RFR wurde eine neue Warte in

Stahlbeton- Skelettbauweise mit einer

Fläche von ca. 260 m 2 an der NO-Seite

des Labortraktes angebaut.

Reaktorhalle (3)

Die Reaktorhalle war ein unterkellerter

Ziegel-Stahl- Skelettbau mit

einer Gesamthöhe von 24,5 m und

einer Grundfläche von ca. 700 m 2 .

Darin befanden sich neben dem

Reaktor unter anderem Brennelemente-Lagerbecken

(AB 1 und AB 2)

sowie im Keller geschoss vier Heiße

Kammern.

Pavillon (4)

Der Gebäudebereich „Pavillon“ mit

einer Grundfläche von ca. 170 m 2

bildete einen stark gegliederten einbzw.

zweigeschossigen Anbau, unter

dem Abluftkanäle zum Filter- und

Ventilationsgebäude sowie Rohrleitungen

für kontaminierte Abwässer

verliefen. Im Gebäudebereich waren

ein radiochemisches Labor und eine

Gamma- Bestrahlungsanlage untergebracht.

Ventilations- und Filtergebäude (5)

Das Filter- und Ventilationsgebäude

mit einer Grundfläche von ca. 350 m 2

war ein anderthalbgeschossiger

Ziegel bau mit einem Flachdach, auf

dem sich der ca. 33 m hohe Fortluftschornstein

befand. Im Inneren

waren die Venti lationskammern mit

den dazugehörigen Filteranlagen.

Altes Armaturenhaus (9)

Das Gebäude mit einer Grundfläche

von ca. 22,7 m x 7,2 m und einer Höhe

von ca. 4,5 m hatte über zwei Drittel

der Grundfläche eine ca. 3,5 m tiefe

Grube mit dem Pumpenfundament.

Das Gebäude bestand aus Mauerwerk;

die Fundamente und Teile der

aufgehenden Wände aus Stampfbeton.

Armaturenhaus (10)

Das Armaturenhaus mit einer

Abmessung von ca. 18,4 m x 12,4 m x

6,0 m wurde im Rahmen der Gesamtrekonstruktion

RFR im Jahr 1985 neu

errichtet und beinhaltete Pumpen,

Armaturen und eine Reinigungs anlage

für das Sekundärkühlsystem. Es

handelte sich um eine Stahl betonskelettkonstruktion

aus Fertigteilen,

| Abb. 3.

Schnittdarstellung RFR-Gebäudekomplex.




wobei die Außenwände aus Gasbetonfertigteilen

bestanden.

Trockenkühltürme (11)

Die Trockentürme hatten eine Abmessung

von ca. 23,3 m x 13,2 m x 8,0 m.

Sie bestanden aus einem Traggerüst

aus Stahl, das mit Leichtmetallelementen

verkleidet war.

Die umbaute Fläche des RFR ohne

Filter- und Venti lationsgebäude betrug

ca. 2.100 m 2 , wobei der Kontrollbereich

die Reaktorhalle mit Kellergeschoss,

Teile des Kellergeschosses des

Anbaus der Reaktorhalle und Teile des

Kellergeschosses des Labortraktes

umfasste. Die Schnittdarstellung in

der Abb. 3 vermittelt einen Eindruck

von der baulichen Anordnung.

Neben dem Reaktor (Detaildarstellung

nach Re konstruktion in Abb. 4)

waren die vier Heißen Zellen, das

Lagerbecken AB 2 und der Pumpenraum

(vgl. Abb. 5) hinsichtlich des

Rückbaus schwierige Teilobjekte.

Aus der Betriebshistorie waren für

die ersten Schritte des Rückbaus und

der Entsorgung vor allem zu berücksichtigen:

p betriebliche und strahlenschutzrelevante

Ereignisse

p der Einsatz von Brennelementen

mit einem hohen Anreicherungsgrad

p die nahezu vollständige Erneuerung

des RFR in den Jahren 1986

bis 1989 mit der Wiederinbetriebnahme

am 27. Januar 1990 sowie

die endgültige Einstellung des

nuklearen Betriebes des RFR am

27. Juni 1991 bedingt durch eine

befristete Genehmigung zum

Versuchsbetrieb.

3 Genehmigungsverfahren

Im Jahre 1991 wurden Aufsichtliche

Anordnungen gemäß § 19 Absatz

3 AtG durch das Sächsische

Staatsministerium für Umwelt und

Landesentwicklung (SMU) erlassen,

um bedingt durch die Auflösung der

Institute der Akademie der Wissenschaften

der DDR zum 31. Dezember

1991 letztendlich einen ungeregelten

Zustand für den RFR zu vermeiden

sowie am 19. Dezember 1991 den

Betreiberwechsel zum VKTA vorzunehmen.

1993 fasste die Sächsische

Staatsregierung den Kabinettsbeschluss,

den Forschungsreaktor endgültig

stillzulegen und zurück zubauen.

Um das Genehmigungsverfahren

nach § 7 Absatz 3 AtG einzuleiten,

stellte der VKTA bereits im Dezember

1994 beim Sächsischen Staatsministerium

für Umwelt und Landesentwicklung

einen Antrag auf Genehmigung

zur Stilllegung und zum Abbau

| Abb. 4.

Schnittdarstellung RFR.

des RFR. Von 1993 bis 1998 wurden

technische, sicherheitstech nische und

strahlenschutztechnische Maßnahmen

zur Anpassung an den bundesdeutschen

Standard durch geführt.

Hervorzuheben ist die Erstellung

eines brandschutztechnischen Gutachtens

durch eine Fremdfirma, auf

deren Grundlage bautechnische Maßnahmen,

die Reduzierung der Brandlast

sowie die Inbetriebnahme einer

neuen Brandschutzanlage erfolgten.

Weiterhin mussten am Forschungsstandort

Rossendorf erst die Voraussetzungen

geschaffen werden, um die

Kernbrennstoffe bzw. Kernmaterialien

verwahren, die radioaktiven Abfälle

und Reststoffe zwischenlagern,

behandeln, analysieren und freigeben

zu können. Dies erforderte zahlreiche

atom- und strahlenschutzrechtliche

Genehmigungen sowie deren bauliche

Umsetzung, die in den Jahren

1992 bis 1999 ebenfalls stattfanden.

Hervorzuheben sind hierbei die Einrichtung

zur Entsorgung von Kernmaterial,

die Transportbereitstellungshalle

für die CASTOREN, das

Zwischenlager, die Reststoffbehandlungsanlage,

das Freimesszentrum,

das Analytiklabor sowie die notwendigen

Einrichtungen für den

Strahlenschutz, wie die Inkorporationsmessstellen

sowie für die

| Abb. 5.

Schnittdarstellung RFR.

Umgebungsüberwachung. Nach Erhalt

der ersten beiden Genehmigungen

zur RFR-Stilllegung [3, 4] wurden

von 1998 bis 2001 die Anlagen kernbrennstoff-

und kernmaterialfrei [5,

6, 7] gefahren und für den VKTA mit

dem Rückbau des 2. Kühlwasserkreislaufes

der Abschluss der Stilllegungsphase

erreicht. Von 2001 bis 2018

erfolgten dann der Rückbau sowie die

Geländesanierung mit abschließender

Flächenprofilierung im Rahmen






der Dritten und Vierten RFR-

Stilllegungsgenehmigungen.

Die Grundlage für die Stilllegung

und den Rückbau des RFR waren insgesamt

vier Genehmigungen mit ihren

Änderungen. Die Erste Genehmigung

45-4653.18 VKTA 01 zur Stilllegung

des Rossendorfer Forschungsreaktors

RFR [3] wurde am 30. Januar 1998

durch das SMU erteilt. Deren 1. Änderungsgenehmigung

45-1653.18 VKTA

01/1, erteilt vom Sächsischen Staatsministerium

für Umwelt und Landwirtschaft

(SMUL) am 06. November

2000, bein hal tete vor allem das Innehaben

der endgültig abgeschalteten

Anlage, die sichere Betriebsführung

der abgeschalteten Anlage zum Zwecke

der Stilllegung, die Überführung

von Brennelementen aus der Spaltzone

in das Brennelement lagerbecken

AB 2 und den innerbetrieblichen

Transport.

Mit der Zweiten Genehmigung

45-4653.18 VKTA 02 des SMUL vom

30. Oktober 1998 zur Stilllegung des

Rossendorfer Forschungsreaktors

RFR [4] (eingereicht im Oktober

1997) konnte der Rückbau des 2.

Kühlkreislaufes realisiert werden,

deren 1. Änderung vom 11. Februar

beinhaltete die Erweiterung des

Genehmigungsumfanges bzgl. eines

Raumes im RFR-Labortrakt.

| Abb. 6.

Überblick zum Aktivitätsinventar.

Die Dritte Genehmigung 4653.18

VKTA 03 zur Still legung und zum

Abbau des Rossendorfer Forschungsreaktors

RFR des SMUL vom 3. April

2001 [8] ermög lichte die Entsorgung

der Betriebsmedien sowie die Außerbetriebnahme

und den Rückbau der

nicht mehr benötigten Systeme und

Komponenten des RFR. Dazu zählte

z. B. der Rückbau des 1. Kühlkreislaufes,

des Speisewassersystems und

des Reaktorbehälters. Insgesamt gab

es 14 Vorhaben, wobei mit Erteilung

der Genehmigung bereits für vier

Vorhaben die Zustimmungen des

SMUL vorlagen. Die restlichen Vorhaben

wurden abbaubegleitend mit

dem SMUL abgestimmt.

Mit der Vierten Genehmigung

4653.18 VKTA 04 des SMUL vom

1. Februar 2005 [9] konnte schließlich

der Abbau der Restanlage des RFR vorgenommen

werden. Der Änderungsbescheid

4653.18 VKTA 04/1 vom

9. November 2010 beinhaltete die Freigabe

und Ent lassung des Raumes 01 im

Gebäude 103 (Notstrom gebäude) und

mit der 2. Änderungsgenehmigung

4653.18 VKTA 04/2 des SMUL vom

9. Januar 2014 wurden die Änderungen

des räumlichen Geltungsbereiches

sowie des Genehmigungsumfanges

beschieden. Hintergrund war die Entscheidung

des VKTA aufgrund von

radiologischen Voruntersuchungen sowie

durch technologisch bedingte

Änderungen der Planungen des Abbaus,

einen Total abbruch der RFR-

Restanlage unter Strahlenschutzbedingungen

vorzunehmen, wobei

hierfür die notwen digen Erläuterungsberichte

benötigt wurden wie ebenfalls

für die Baufreiheit die Erweiterung des

räumlichen Geltungsbereiches.

Zur jeweiligen Genehmigungsplanung

erfolgte die grundlegende

Beschreibung des Gesamtvorhabens

durch Erläuterungsberichte. Im

Rahmen der Vierten Genehmigung

gab es beispielsweise 18 Vorhaben,

wobei das Vor haben 11 und das Vorhaben

15 jeweils in drei Teile

gegliedert war. Es wurden folglich

22 Erläuterungs berichte erstellt, von

denen sieben bereits bei Erteilung der

Genehmigung von der zuständigen

Behörde bestätigt wurden. Die weiteren

Erläuterungsberichte erhielten im

Zuge des Aufsichtsverfahrens die

Zustimmung. Nach Abschluss eines

Vorhabens wurde ein Abschlussbericht

erstellt und an die zuständige

Behörde übergeben.

4 Ausgangssituation

(radiologisch,

konventionell)

Die radiologische Ausgangssituation

war geprägt durch das Vorhandensein

von be- und unbestrahltem Kernbrennstoff,

Kernmaterial sowie von

Aktivierung und Kontamination

unterschiedlichster Stoffe, die lokal

und aktivitäts mäßig sehr differenziert

waren. Einen Grobüberblick zum

Aktivitätsinventar gibt die Abb. 6.

Die Herstellung der Kernbrennstoff-

Freiheit der RFR- Anlage war ein

wichtiger Schritt für den Start des

Rückbaus. Dazu mussten sowohl die

951 bestrahlten Brenn elemente mit

einer Gesamtaktivität von 8,91 E+15

Bq als auch einige Posten kernbrennstoffhaltige

Abfälle aus der RFR-

Anlage entfernt werden. Diese Vorhaben

wurden im Februar 2001 abgeschlossen.

Weitere Einzelheiten sind

im Abschnitt 7.2 zu entnehmen (s. atw

1/2020).

Hinsichtlich der Aktivierung waren

alle reaktornahen Baustrukturen

und Anlagenteile, wie z. B. die Inneneinbauten

des Reaktors, die Bestrahlungskanäle,

das Biologische Schild,

die Thermische Säule zu beachten.

Wichtige Aktivierungsnuklide waren

beispielsweise Eu-152, Ba-133, H-3,

Co-60 (Beton/Bauschutt), Fe-55,

Co-60, Ni-63, H-3 (Eisenteile), Fe-55,

Co-60, Ni-63, H-3 (Stahlguss-Reaktordeckel),

H-3, Ni-63, Co-60, Be-10




(Be-Reflektorelemente), Ni-63, Ni-59,

H-3, C-14, Co-60 (Borcarbid-

Absorberstäbe), C-14, H-3, Eu-152,

Co-60 (Graphit). Unerwartet wurde

bei der radiologischen Voruntersuchung

festgestellt, dass anstatt

des in den auunterlagen ausgewiesenen

Barytbetons das Biologische

Schild aus Beton mit Eisenteilen

bestand.

Kontaminationen resultierten vor

allem aus flüssigkeitsgetragenem

Transport nach Freisetzung sowie aus

einem Am-241-Ereignis (Freisetzung

aus einem Target, 1969). Die Hauptnuklide

in der RFR-Anlage waren

H-3, Co-60, Sr-90+, Cs-137, Eu-152,

U- und Pu-Nuklide und Am-241. Zur

Erfassung der jeweiligen Ausgangssituation

wurde nach der historischen

Erkundung die radiologische Voruntersuchung

nach Möglichkeit mit

konventionellem Schadstoffuntersuchungen

gekoppelt. Dies reduzierte

später den zeitlichen und finanziellen

Aufwand und so konnten z. T. potentielle

Verdachtsflächen besser beurteilt

und die Entsorgung schneller

voran gebracht werden.

Das Auftreten von chemotoxischen

(konventionellen) Schadstoffen resultiert

sowohl aus Betriebsabläufen

als auch im Wesentlichen aus schädlichen

Bestandteilen der Baustoffe.

Zu den Gebäudeschadstoffen zählen

beispielsweise Polyzyklische Aromatische

Kohlenwasserstoffe (PAK,

insbesondere in Teeranstrichen),

Polychlorierte Biphenyle (PCB, Oberflächenbeschichtungen),

künstliche

Mineralfasern (KMF) und Asbest (in

Dämmstoffen). In Ausrüstungen

und Anlagen fand man z. B. PCB

(Transformatoren, Kondensatoren),

Mineral kohlenwasserstoffe (MKW,

Pumpen) oder Asbest (Dichtungen).

Aus Betriebsabläufen stammen u. a.

Kontaminationen von Schwermetallen

und MKW. Das frühzeitige

Erkennen der Schadstoffsituation

ermöglichte die planerische Einarbeitung

in die Rückbauprozesse

sowie vor allem einen zeitlichen Vorlauf

für eine sachgerechte Entsorgung

des betreffenden Materials zu erhalten.

Zu der Schadstoffproblematik

veröffentlichte der VKTA einige

Beiträge [10–12]. Bezüglich des RFR-

Rückbaus sind insbesondere festgestellte

PAK-Kontaminationen im Beton

und Erdreich, das Vorfinden von

Asbest hinter den Stahlbecken im

Lager becken AB 2 (nicht in Bauzeichnung

erwähnt) sowie lokale Kontaminationen

mit MKW, Quecksilber

und Schwermetallen im Beton hervorzuheben.

5 Planungskonzept

Grundsatz für die Planung war zum

einen das Ziel, bei Stilllegung und

Rückbau die Entsorgung von freigegebenen

Stoffen parallel vorzunehmen

und die radio aktiven Abfälle im

1999 errichteten Zwischenlager

Rossendorf, welches 2000 erweitert

wurde, ordnungs gemäß für die spätere

Endlagerung zu lagern. Zum

anderen sollte das Betriebspersonal

des RFR so weit wie möglich eingebunden

werden und der VKTA nicht

nur die Planungshoheit innehaben,

sondern auch möglichst viele kostengünstige

Beiträge zur Aufgabenerfüllung

leisten. Da der VKTA planungstechnisch

nicht alles abdecken konnte,

wurde ab 2004 mit der heutigen

Siempelkamp NIS Ingenieurgesellschaft

mbH (NIS) eine Fachfirma eingebunden.

NIS war, unterstützt durch

Fachplaner in den Gebieten Lüftung,

Tragwerksplanung und Schadstofferkundung,

mit der Planung, der

Erstellung und Bewertung von Ausschreibungsunterlagen

sowie der

Bau überwachung beauftragt. Ebenso

wurden für spezielle Aufgaben weitere

Fachfirmen gebunden. Der VKTA

selbst, zeitweilig unterstützt durch

einen Projektsteuerer, stellte die

Rückbauleitung, bestehend aus:

p Rückbauleiter

p Gebäudeverantwortlicher,

Strahlen schutzbeauftragter

p Strahlenschutzingenieur/in

p Strahlenschutzfachkraft

p technischem Personal für strahlenschutztechnische

Messungen,

Transporte und kleineren technischen

Aufgaben

Die personelle Absicherung der Rückbauleitung

in der Rückbau-Etappe bis

2007 wurde dabei ausschließlich

durch das ehemalige Betriebspersonal

des RFR sicher gestellt. Die ab 2006

auszuführenden Arbeiten wurden

strukturiert, geplant, ausgeschrieben

und an externe Dienstleister im

Rahmen von Verträgen nach VOL

oder VOB vergeben. Hierzu wurde

eine Struktur verwendet, die sich

einheitlich in Arbeits-, Termin- und

Finanzpläne gliederte mit den

Baulosgruppen:

p Vorbereitende Maßnahmen

p Bereitstellung von Ausrüstungen

p Dienstleistungen Abbau

p Abbaubegleitende lufttechnische

Maßnahmen

p Dienstleistungen Dekontamination

p Arbeitsbegleitender Strahlenschutz

p Mess- und Freimessprogramme in

Planung und Durchführung

p Abbaubegleitende sonstige

Maßnahmen

p Abbrucharbeiten bis zur Geländeprofilierung

Die Leistungen des arbeitsbegleitenden

Strahlenschutzes, der Planung

und Durchführung von radiolo gischen

und konventionellen Messungen/Analysen

und von (Frei)Mess programmen

übernahm der VKTA weitestgehend

selbst. Dazu nutzte der VKTA u. a. sein

nach DIN EN ISO/IEC 17025 akkreditiertes

Labor für Umwelt- und Radionuklidanalytik.

Andere auszuführende

Arbeiten wurden in rund 200 Losen

i. d. R. ausgeschrieben und den Losgruppen

zugeordnet. Die bewusst

kleinteilige Vergabe von Losen erlaubte

es dem VKTA, mit seinen Planern

auf besondere und erst im Verlauf des

Abbaus erkennbare neue Situationen

zügig zu reagieren, ortsnahe Firmen

einzubinden und letztlich Kosten zu

sparen.

Der Arbeitsumfang des Gesamtprojektes

entsprechend der jeweiligen

Genehmigung wurde, wie bereits

erwähnt, in Einzelvorhaben gegliedert,

die in Erläuterungsberichten

der Genehmigungsbehörde dargelegt

wurden. Die zeit liche Abarbeitung der

einzelnen Vorhaben erfolgte entsprechend

eines Rahmenplanes, der

mehrmals den objektiven Umständen

angepasst werden musste. Für jedes

Einzelvorhaben lag ein Detailplan vor.

Die Durch führung der Arbeiten erfolgte

auf der Grundlage von Arbeitsanweisungen

und Ablaufplänen,

wobei u. a. vor jedem Vorhaben ein

Rückbauerlaubnis- sowie ein Arbeitserlaubnisverfahren

gemäß der entsprechenden

VKTA- Regelung durchzuführen

war.

6 Meilensteine Stilllegung

und Rückbau

Nach dem Erhalt der jeweiligen

Genehmigung konnten die geplanten

Arbeiten durchgeführt werden, wobei

nach folgend wichtige Meilensteine

des Rückbaufortschrittes beim RFR

chronologisch (Durchführungszeitraum

in Klammern) aufgeführt

werden:

p Kabinettsbeschluss des Freistaates

Sachsen zur „Endgültige Stilllegung“

(13. Juli 1993)

p Antrag auf Stilllegung

(21. Dezember 1994), Erhalt der

Ersten Genehmigung (30. Januar

1998, vgl. Abschnitt 3) und somit

ab 2. Februar 1998 RFR-Betriebsführung

nicht mehr auf der

Grund lage der Aufsichtlichen

Anord nungen des SMU, u. a.

vom 28. Juni 1991

p Herstellen der Kernbrennstofffreiheit

des Reaktor behälters






durch Umlagerung der Brennelemente

aus dem Reaktorkern ins

Lagerbecken AB 2 und dem Ausbau

der kernbrennstoffhaltigen

Neutro nen detektoren (6. April

1998)

p Antrag auf Zweite Stilllegungsgenehmigung

(10. Oktober 1997);

Erhalt der Zweiten Genehmigung

(30. Oktober 1998) mit

1. Änderung (11. Februar 1999)

p Rückbau des 2. Kühlkreislaufes

mit Entkernung und Abriss des

Armaturenhauses, der Trockenkühltürme,

des Pumpenhauses

sowie der Verbindungsleitungen

zwischen den Gebäuden und dem

Reaktorgebäude; mit SMUL-

Schreiben vom 16. August 1999

waren diese Anlagen nicht mehr

Bestandteil der RFR-Anlage

p Überführung der Brennelemente

aus dem Lagerbecken AB 2 in

CASTOR® MTR 2-Behälter, Endabfertigung

mit anschließendem

Transport in die Transportbereitstellunghalle

des VKTA

(24. November 2000)

p RFR-Anlage kernbrennstofffrei

nach Abgabe/Konditionierung der

restlichen Posten Kernbrennstoffabfälle

(26. Februar 2001)

p Antrag auf Dritte Stilllegungsgenehmigung

(29. Dezember

1998); Erhalt der Dritten

Genehmigung (3. April 2001)

p Ausbau des Reaktorbehälters,

Arbeiten zur Transportbereitstellung

und Transport zum

Konditionierer (2001/2002)

p Antrag auf Vierte Stilllegungsgenehmigung

(31. Juli 2003);

Erhalt der Vierten Genehmigung

(1. Februar 2005) mit 2. Änderungsgenehmigung

(9. Januar

2014)

p Transport der RFR-Brennelemente

mittels CASTOREN ins Zwischenlager

Ahaus (3 Konvoitransporte

vom 30. Mai 2005 bis 13. Juni

2005)

p Abbau des RFR-Baukörpers einschließlich

der Aus kleidung des

Lagerbeckens AB 1 (2008/2009)

p Ausräumen, Dekontamination und

Abbruch der Heißen Zellen

(2009/2010)

p Ausbau des Deaerators

(2010/2011)

p Ausräumen, Dekontamination und

Entkernung des Kellergeschosses

(2011/2012)

p Abbau, Dekontamination und

Entsorgung der Teile des

frei gegebenen Fortluftschornsteines

(2013/2014)

p Dekontamination, Entkernung

und Abriss des Filter- und

Ven tilationshauses (2014/2015)

p Dekontamination, Entkernung

und Abriss des Labortraktes mit

Reaktorhalle und -warte (2016)

p Abschluss des Ausbaus aller

tiefliegender Bau strukturen und

Bodensanierung (2016)

p Abschluss der Baugrubenverfüllung

sowie Beendigung der

restlichen bautechnischen

Maß nahmen (Profi lierung,

Oberflächengestaltung

(2017/2018)

Autoren

Reinhard Knappik

Klaus Geyer

Sven Jansen

Cornelia Graetz

VKTA Rossendorf

Bautzner Landstraße 400

01328 Dresden

Deutschland

Imprint

| Editorial Advisory Board

Frank Apel

Erik Baumann

Dr. Erwin Fischer

Carsten George

Eckehard Göring

Dr. Florian Gremme

Dr. Ralf Güldner

Carsten Haferkamp

Christian Jurianz

Dr. Anton Kastenmüller

Prof. Dr. Marco K. Koch

Ulf Kutscher

Herbert Lenz

Jan-Christan Lewitz

Andreas Loeb

Dr. Thomas Mull

Dr. Joachim Ohnemus

Olaf Oldiges

Dr. Tatiana Salnikova

Dr. Andreas Schaffrath

Dr. Jens Schröder

Norbert Schröder

Prof. Dr. Jörg Starflinger

Dr. Brigitte Trolldenier

Dr. Walter Tromm

Dr. Hans-Georg Willschütz

Dr. Hannes Wimmer

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ISSN 1431-5254




First On-site Demonstration of Laser- based

Decontamination Technology in Germany

Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann

and Antonio Hurtado

1 Introduction Several ongoing dismantling and decommissioning (D&D) projects of nuclear facilities in

Germany produce wide experiences in terms of the regulation framework and the practical implementation. The D&D

of commercial and research reactors require a comprehensive adaption of the available technologies to meet the

demands of the associated health and safety requirements. This causes a rising monetary and human effort.

To face this challenge the German

federal government focuses the

research and development activities

to:

a- protect human beings and the

environment during nuclear

decommissioning,

b- educate staff and young academics

and

c- develop and optimize decommission

methods.

The dismantling and decommissioning

of the various nuclear facilities

located at Karlsruhe Institute of

Technology (KIT) Campus North, the

Kerntechnische Entsorgung Karlsruhe

GmbH (KTE) faces several technical

challenges.

On the one hand, metallic surfaces,

e. g. in the Hot Cells Facility (HZ),

must be decontaminated radiologically,

on the other hand, paint layers

as well as radiological contamination

must be removed from metallic components

during post treatment, e. g.

in the Waste Treatment Department

(EB).

Furthermore, large surfaces of

concrete structures, e. g. in the

Reprocessing Building (PG) of the

Karlsruhe Reprocessing Plant (WAK),

in the Multi-Purpose Research Reactor

(MZFR) and in the Research Reactor

(FR-2) must be cleared from paint

layers, containing sometimes high rate

of polychlorinated biphenyls (PCB).

The overriding task in these

processes is to enclose the radioactive

substances and limit the radiation

exposure of personnel and environment.

Furthermore, the volume of

waste should be minimized. The work

situation for the operating personnel

must be constantly monitored and

the workload should be reduced

wherever possible.

Sometimes, the currently available

chemical and mechanical processes,

reach their limit to comply with the

objects mentioned before.

Abrasive blasting of metallic structures

results in the production of

secondary waste, which has to be

further processed in case of PCB. In

chemical decontamination processes,

pickling acids also must be specially

treated or stored.

Furthermore, all mechanical processes,

such as shaving or sandblasting,

also cause high workloads

for the operating personnel due to

vibrations, restoring forces and noise.

Chemical processes, however, can

cause damage through skin contact

and inhalation.

Decontamination by laser beam

represents an interesting alternative.

This technology can expand the

repertory of currently available decontamination

tech nologies. Being a

contact-free procedure it comes along

with the advantage of waste minimization

and reduces restoring force for

hand-held as well as semi-automatic

application, following NEA recommendations

[1].

At TU Dresden laser-based decontamination

tech nology has been developed

during the last years. The

research project LaPLUS has been

aimed at the optimi zation of the

chemical-toxic decontamination of

concrete surfaces and the technology

transfer for the radiologic decontamination

of metal surfaces. For that

purpose, special hand-held laser tools

for the use in nuclear sites were

designed and tested on a laboratory

scale. An essential part of the project

was the technology transfer from

laboratory scale to prototype status.

For that purpose, the laser-based zi.

decontamination was demonstrated

in a realistic environment at the

Multi-Purpose Research Reactor

( MZFR).

Under project leadership of TU

Dresden, TU Berg akademie Freiberg

(Development of process analysis

tools) and IABG mbH (Design of laser

tools) participated in the project. The

KTE accompanied the project as an

associated partner and supported the

technology transfer to the MZFR.

This paper presents the major

results achieved inside the project. It

covers the laser-based decontamination

of PCB containing coatings on

concrete and that of radiologic contaminated

metal surfaces, the design

and the test of special laser tools

developed for the application in

nuclear facilities.

2 Application of laser

systems for decontamination

When a laser beam hits a surface,

absorption of the energy leads to rapid

heating of the substrate. The resulting

spatial and temporal temperature

distribution depends on optical and

thermal characteristics of the substrate

as well as on the parameters of

the laser process [2]. Heating of the

surfaces can result in melting, evaporation

or sublimation. All of those

mechanisms can be used for the

removal of unwanted species, as

necessary for surface decontamination.

Additionally different

practical requirements defined by the

decontamination task require a

careful selection of a suitable laser

system.

Rapid heating of the treated

substrate surface above evaporation

temperature can be achieved using

pulsed lasers. The pulse duration is directly

proportional to the volume of

the heat affected zone [3], which

means that shorter pulses lead to

smaller heat affected volume. The

possibilities and requirements of

radiologic decontami nation using

short pulsed lasers is treated in

section 3.

A different result can be achieved

applying a con tinuous working (cw)

laser beam, which continuously heats

the surface. In case of interaction with

paint, e.g. epoxy based coatings, the

volatile components of that coating

will evaporate and incinerate after

reaching the ignition temperature. A

subsequent combustion process of the



First On-site Demonstration of Laser- based Decontamination Technology in Germany ı Torsten Kahl, Georg Greifzu, Carsten Friedrich, Christian Held, Marion Herrmann, Wolfgang Lippmann and Antonio Hurtado



coating is established. Chemical-toxic

decontamination of PCB-containing

paint using a continuous wave laser is

treated in section 4. Application of

high power diode lasers allows rapid

treatment of surfaces with economically

desirable laser equipment.

Apart from beam parameters,

further practical requirements have to

be met for successful application of

lasers in nuclear facilities.

High mobility and toughness is

required from the beam shaping optics,

as opposed to regular industrial

application. This demand is coped

with using simplified 1D scanning

optics.

For the cost-efficient use of lasers,

the application of optical fibres for

beam transport is mandatory. This

allows to arrange the laser-system

outside the contaminated area, thus

eliminating the risk of contamination

for the laser- system. Only part of the

fibre and the beam-shaping laser

optics will be placed inside the control

area and can still be reused on several

decommissioning sites, as explained

in section 6.

3 Decontamination

of metal surfaces

Laser-based cleaning of metal surfaces

has been established within the last

decade in applications like pre-treatment

for welding, restauration of art

and decoating of paint. In these cases,

a laser beam is scanned and moved

over the soiled surfaces and removes

adhering unwanted species. High

process selectivity can be achieved

on metal surfaces, because a large

fraction of the beam is reflected on

blank metals as opposed to higher

absorption on organics and oxides.

Laser-based cleaning substitutes

chemical as well as mechanic abrasive

processes and results in a reduction of

waste of factor 2.6 J/cm² for

austenitic steel, >3.5 J/cm² for ferritic

steel and >5.3 J/cm² for zinc plated

Wavelength

Average Power

Pulse-Energy

Peak Power

Pulse length

Pulse frequency

Scanner

1064 nm

150 W

11.5 mJ @ 12 kHz

112 kW @ 12 kHz

102 ns@ 12kHz

12 - 40 kHz

Spot diameter 472 µm

Scanning width

Mass of optics

60 mm

1,7 kg

| Tab. 1.

Specifications of Nd:YAG Laser CL150.

| Fig. 1.

Idealized scenarios of contamination, blue color symbolizes contamination:

contamination of uncovered surfaces (A),

on top of covered surface (B) and

on as well as under covering layer (C),

contamination within oxide layer (D).

| Fig. 2.

Ablation depth as function of cumulative energy with a pulse duration

of 105 ns.

| Fig. 3.

Concentration of surrogates on sample surface prior and after

decontamination.




steel. Tested paint layers can be

ablated at fluence above 1.8 J/cm²,

while oxide layers can be removed at

even lower fluence.

The achieved ablation depth for

the above mentioned materials is

plotted in Figure 2.

It can be seen that the removal

speed is about 100 times higher for

paint than for metals, what supports

the selective removal of the paint on

the metal ground. A variation of laser

fluences between 1.8 – 3.5 J/cm² has

been tested for its decontamination

effect on blank austenitic steel. Single

scans resulted in a contamination

reduction of 50 – 90 % for all surrogates.

Similar tests have been conducted

on painted surface applying a

laser fluence of 4.4 J/cm², to prove

complete removal of the covering

layer (Figure 3).

Both tested cases result in a

decreased contaminant concentration

and decontamination factors up to

98.9 have been achieved.

4 Laser-based PCB

degradation

The TU Dresden has developed a PCB

decontamination process by utilising

a continuous wave diode laser.

Laboratory experiments on concrete

surfaces coated with epoxy paint

demonstrated a reduction of 96.83 %

of the PCB value. The PCB decomposition

rate on the surface and in

the exhaust gas before filtering is

88.75 % [16].

PCB decomposes at temperatures

above 800 °C. At optimized process

parameters the temperature in the

laser spot is much higher [17], enable

the full decomposition of the PCB.

Rapid cooling of the exhaust gas is

required to prevent the formation of

toxic polychlorinated dioxins and

furans (PCDD/PCDF) which can be

ensured by the laser processing. The

application of a fabric filter in the

extraction and filtration unit limits

the temperature of filtered exhaust

gas to



| Fig. 6.

Overview of tested samples as supplied by KTE (A– lid of a 200 liter drum, B– steel sheet, C– coated u

profile, D– locker door part).

| Fig. 7.

Components for decontamination of complex metal surfaces, (A) laser tool for the decoating of flat

surfaces and (B) laser tool for the decontamination of edges.

transported using optical fibres to

reach the experimental area. This

feature allows for the protection of

costly laser equipment from contamination

in any case, as described in

section 2. The user can simply equip

the required laser tool and start the

decontamination process. Suction

and filtration of mobilized contamination

at high efficiency is facilitated

applying a commercial filtration and

suction unit by ULT GmbH. The

localized suction of aerosols leads to

prevention of recontamination and

guarantees clear vision for the user at

all times.

Restricted access to the experimental

area is beneficial to simplify the

laser safety requirements, even in case

of a technology demonstration.

A full-fledged risk assessment

according to §5 of ArbSchG [18] and

§3 of ArbStättV [19] has been filed to

scan for any arising risks for operators

coming along with the application of

the whole set up. All requirements of

the occupational safety have been

tested and proven. The risk assessment

covers the following subject

area:

p General occupational safety

p Laser safety

p Safety from toxic substances (PCB)

p Radiation safety

p Respiratory protection

The risk assessment complies with the

laws, guidelines (ASR), workers compensation

board rules and provisions.

The risk assessment was finalized by

TU Dresden and IABG and checked by

KTE and has been the fundamental

milestone for approval of the on-site

demonstration at MZFR.

5.3 Decontamination of

complex metal geometries

The decontamination of metal surfaces

at MZFR was conducted on

paints and metals prior unknown, to

check the ability to transfer results

from laboratory scale to on-site tests.

To test limits of the laser tool, samples

of varying geometries have been

supplied by KTE, e.g. painted steel

sheets, bracket steel or complex surfaces

like ridges/spline/serration, as

shown in Figure 6.

Different laser tools were designed

to ensure the decontamination at

different surface geometries, e.g. in

case of tighter angles the tool was

adapted to maintain the complete

suction of particles and aerosols

during the laser process. A tool change

can be completed within 1.5 min,

Figure 7.

Similar ablation characteristics

were found for all tested samples as

compared to the laboratory results

(Figure 2). This implies that generalized

material characteristics for paint

layers and metals can be applied.

From practical point of view, the

flexibility of the laser ablation process

was verified, as all tested geometries

were completely cleaned without

residues (Figure 8).

5.4 Decontamination of

concrete walls and floor

at MZFR

Decontamination of concrete demands

solutions for a multitude of

demands, e.g. uneven surfaces, varying

thicknesses of paint, different

shapes of surfaces and unknown PCB

concentration in the decontamination

paint. The practicability of the

developed laser system was tested on

walls and floors at the MZFR on

sample surfaces sized 30 x 30 cm².

The decontamination paint on the

wall was thin and exhibited a high

PCB concen tration (see Table 3). The

floor featured lower PCB concentration

and very thick paint (approx.

1.5 mm).

In Figure 9 the side view of the

laser head is shown during the decontamination.

During the process all

arising by-products, like visible particles

and gases, are extracted by the

extraction and filtration unit and do

not endanger the working staff.

Figure 10 displays the result

of decontamination on a floor

| Fig. 8.

Steel sheet partly decoated.

| Fig. 9.

Laser head during the decontamination of the floor.




| Fig. 10.

Demonstration field floor (A) before and (B) after the decontamination

with multiple scans.

Sample number Position Before decontamination

PCB in mg/kg

demonstration field before (A) and

after (B) the laser process.

In Figure 11 the laser head is

displayed with different adaptions

for flat areas (picture A and B) and

internal corners (picture C). The

modular set-up enables a fast change

of the adaptions omitting any changes

to the leaser heads basic components.

Using such adaptions the decontamination

of a wall with adjoining

corners was demonstrated, illustrating

the clear advantage compared

to shaving equipment, where usually

boundary areas need an additional,

time- consuming handling step.

The removal of thick layers and/or

remaining soot can be realized by

multiple passes. Both have been

demonstrated in Karlsruhe.

During the demonstration an

average laser power of 3.7 kW was

applied, resulting in a practical

determined ablation speed of 6 m²/h

5.5 PCB sampling at the MZFR

Within the demonstration PCB

sampling was performed to prove the

decontamination rate of the laserbased

paint removal. This study aimed

at the evaluation of the PCB concentration

of the primary source. To

release material the PCB-concentration

must amount less than 50 mg/

kg. Samples with mass > 50 mg were

extracted using an electric scraper.

AGROLAB Labor GmbH performed

the analysis of the samples according

to DIN EN 15308. The samples taken

before the decontamination consisted

mainly of decontamination paint.

After the decontamination no paint

remained on the surface anymore,

therefore the now surfacing concrete

was sampled.

In Figure 12 the sampled surfaces

are shown before PCB sampling (A),

after PCB sampling (B) and after

decontamination and the consecutive

PCB sampling (C). The corresponding

PCB concentration is shown in

Table 3.

The laser-based decontamination

process resulted in an average reduction

of PCB of 98.7 % within a practical

application. This value is even

higher than the value of the laboratory

experiments performed at the TU

Dresden [16]. The remaining PCB

concentration after the decontamination

is mainly caused by the PCB

inside the concrete matrix. After sealing

the walls with PCB containing

paint the PCB diffuses from the decontamination

paint (primary source)

into the concrete. Alternative laserbased

technologies are available to

remove and to vitrify concrete surfaces

in a single process step [20; 21].

| Fig. 11.

Laser head for concrete decontamination;

A- laser head (round version), B- laser head (angled), C- laser head internal corner.

After decontamination

PCB in mg/kg

Difference

mg/kg

| Fig. 12.

Examples of a demonstration field:

(A)- before sampling, (B)- after decontamination and (C)- after decontamination and sampling.

6 Conclusion and future

prospects

The TU Dresden successful demonstrated

the application of laser-based

decontamination on radiologic and

chemical-toxic contaminated metal

and concrete surfaces at the MZFR in

Karlsruhe. All necessary documents to

obtain acceptance of the KTE were

prepared by TU Dresden and approved

by KTE. An independent work

permit was available to TU Dresden.

The selective decontamination on

metal surfaces is characterized by a

process stop after removing any coating

(decontamination paint, oxide

and/or contamination). The practical

removal rate on metal is 0.42 m²/h

using a mean laser power of 150 W

and an average paint thickness of

165 µm. The application of laser-based

decontamination to PCB-contaminated

concrete surfaces resulted in

a mean reduction of 98.7 % of the

PCB-concentration. During the onsite-demonstration

a removal rate of

6 m²/h of paint from concrete walls

with 3.7 kW laser power were accomplished.

The handheld laser tools

Reduction

%

1 Floor 9.3 0.23 9.07 97.5

2 Floor 30.1 0.04 30.06 99.9

3 Wall 3,810 67.3 3,742.7 98.2

4 Wall 2,360 19 2,341 99.2

5 Wall 1,976 18.9 1,957.1 99.0

6 Wall 2,510 14.2 2,495.8 99.4

7 Wall 2,272 49.7 2,222.3 97.8

| Tab. 3.

PCB concentration of the sampling before and after the decontamination.

Ø 98.7 %






enable high flexibility and reduce

restraining forces for the working

staff. The dry laser cleaning process

runs completely without abrasive or

chemical aids thus reducing the

amount of secondary waste drastically.

The determined removal rates on

concrete and metal are highly affected

by the thickness and optical pro perties

of the decontamination paint. The

great advantage of lasers is the scalability

of their power. The use of

commercial available lasers with

higher power would result in higher

ablation rates without any changes to

the applied laser tools.

The on-site demonstration at

MZFR Karlsruhe verified the possibilities

of that technology for decontamination

of contaminated surfaces.

Robust laser tools can offer an

alternative for surface decontamination

in the near future. Following the

positive results of the research project

LaPLUS more detailed insight into the

behavior of mobilized aerosols as well

as further tests within control areas

will be provided from follow up projects.

Industrial application of the

laser-based decontamination process

should become feasible in the next

years

Acknowledgement

The on-site demonstration at the

MZFR in Karlsruhe was carried out

inside the Research and Development

Project LaPLUS financed by The

German Federal Ministry of Education

and Research (BMBF) under the

contract number 15S9215A.

We would like to thank all colleagues

of the MZFR that supported

the on-site-demonstration with enthusiasm

and useful evidences as well

as practical drive. Our special thanks

go to the members of our technical

staff, who were a great support during

the exhausting time of demonstration

and the whole project as well.

References

[1] Nuclear Energy Agency: R&D and Innovation Needs for

Decommissioning of Nuclear Facilities. 2014 (7191)

[2] BLIEDTNER, M.: Lasermaterialbearbeitung. München:

CARL HANSER Verlag GMBH, 2013

[3] Leitz, K.-H.; Redlingshöfer, B.; Reg, Y.; Otto, A.; Schmidt, M.:

Metal Ablation with Short and Ultrashort Laser Pulses. In:

Physics Procedia 12 (2011), S. 230–238

[4] Büchter, E.: Entwicklung eines Hochleistungs-Laserstrahl-

Reinigungsgerätes zur Ressourcen schonenden

Entschichtung von Oberflächen. Herzogenrath, 2004

[5] Carvalho, L.; Pacquentin, W.; Tabarant, M.; Maskrot, H.;

Semerok, A.: Growth of micrometric oxide layers to explore

laser decontamination of metallic surfaces. In: EPJ Nuclear

Sciences & Technologies 3 (2017), S. 30

[6] Delaporte, Ph.; Gastaud, M.; Marine, W.; Sentis, M.;

Uteza, O.; Thouvenot, P.; Alcaraz, J. L.; Le Samedy, J. M.;

Blin, D.: Radioactive oxide removal by XeCl laser.

In: Applied Surface Science 197-198 (2002), S. 826–830

[7] Kim, D.; Lim, H.: Laser Decontamination of Carbon Steel

Surfaces. In: ISIJ International (2003), Nr. 43, S. 1289–1291

[8] Leontyev, A.; Semerok, A.; Farcage, D.; Thro, P.-Y.; Grisolia, C.;

Widdowson, A.; Coad, P.; Rubel, M.: Theoretical and

experimental studies on molybdenum and stainless steel

mirrors cleaning by high repetition rate laser beam.

In: Fusion Engineering and Design 86 (2011),

9-11, S. 1728–1731

[9] Edelson, M. C.; Pang, H.: A laser-based solution to industrial

decontamination problems. 1995 (ICALEO 1995 768)

[10] Potiens, A. J.; Dellamano, J. C.; Vicente, R.; Raele, M. P.;

Wetter, N. U.; Landulfo, E.: Laser decontamination of the

radioactive lightning rods. In: Radiation Physics and

Chemistry 95 (2014), S. 188–190

[11] Sadanori, S.; Seiji, A.; Inoue, T.: Applying laser technology to

decommissioning for nuclear power plant (Advanced

High-Power Lasers and Applications). Osaka, Japan, 1999

[12] Takakuni, H.; Yutaka, K.; Masato, M.: Application of a laser to

decontamination and decommissioning of nuclear facilities

at JAERI (Advanced High-Power Lasers and Applications).

Osaka, Japan, 1999

[13] Vatry, A.; Grisolia, C.; Delaporte, Ph.; Sentis, M.: Removal

of in vessel Tokamak dust by laser techniques. In: Fusion

Engineering and Design 86 (2011), 9-11, S. 2717–2721

[14] Nilaya, J. P.; Raote, P.; Kumar, A.; Biswas, D. J.: Laser-assisted

decontamination – A wavelength dependent study. In:

Applied Surface Science 254 (2008), Nr. 22, S. 7377–7380

[15] Delaporte, Ph.; Gastaud, M.; Marine, W.; Sentis, M.; Uteza, O.;

Thouvenot, P.; Alcaraz, J. L.; Le Samedy, J. M.; Blin, D.: Dry

excimer laser cleaning applied to nuclear decontamination.

In: Applied Surface Science 208-209 (2003), S. 298–305

[16] Anthofer, A.; Kögler, P.; Friedrich, C.; Lippmann, W.;

Hurtado, A.: Laser decontamination and decomposition

of PCB-containing paint. In: Optics & Laser Technology 87

(2017), S. 31–42

[17] Anthofer, A.: Oberflächenentschichtung mittels

Laserstrahlung. Untersuchungen zur Dekontamination

radioaktiv und chemisch-toxisch belasteter Betonoberflächen

mittels Lasertechnologie. Dissertation, Dresden, 2014

[18] Bundesministerium der Justiz und für Verbraucherschutz:

Gesetz über die Durchführung von Maßnahmen des

Arbeitsschutzes zur Verbesserung der Sicherheit und des

Gesundheitsschutzes der Beschäftigten bei der Arbeit

(idF v. 31. 8. 2015) (1996-08-07)

[19] Bundesministerium der Justiz und für Verbraucherschutz:

Verordnung über Arbeitsstätten

(idF v. 18. 10. 2017) (2004-08-12)

[20] Lippmann, W.; Herrmann, M.; Pietsch, C.; Reinecke, A.;

Hille, C.; Wolf, R.; Zeuner, A.: LASABA II : Dekontamination

silikatischer Oberflächen in kerntechnischen Anlagen mittels

Laserabtrag bei gleichzeitiger Abproduktkonditionierung.

Abschlussbericht, Dresden, 2008

[21] Hurtado, A.; Littwin, R.; Lippmann, W.: MANOLA: Manipulator

gesteuerter Oberflächenabtrag durch Lasertechnologie.

Abschlussbericht, Dresden, 2011

Authors

Torsten Kahl,

Georg Greifzu,

Marion Herrmann,

Wolfgang Lippmann,

Antonio Hurtado

Technische Universität Dresden

Chair of Hydrogen and

Nuclear Energy

Institute of Power Engineering

George-Bähr-Str. 3b

01062 Dresden

Germany

Carsten Friedrich,

Christian Held

Kerntechnische

Entsorgung Karlsruhe

Hermann-von-Helmholtz-Platz 1

76344 Eggenstein-Leopoldshafen

Germany




Scientists and Professionals from all around the World in Karlsruhe:

The 25 th Edition of the Frédéric Joliot/Otto Hahn Summer School

on Nuclear Reactors “Physics, Fuels and Systems“

Victor Hugo Sanchez-Espinoza

The Institute of Neutron Physics and Reactor Technology (INR) of the Karlsruhe Institute of Technology (KIT) together

with the Commissariat à l’Énergie Atomique (CEA) hosted this year the “Frédéric Joliot/Otto Hahn (FJOH) Summer

School“ at the Akademiehotel in Karlsruhe from August 21 st to August 30 th 2019. The topic of this year’s school was

“Innovative Reactors: Matching the Design to Future Deployment and Energy Needs”.

549

REPORT

It was organized in six technical blocks

and one seminar devoted to “Public

Acceptance of Energy Technologies”

given by S. Hirschberg from PSI

Switzerland. The first introduc tory

block was devoted to “Various Innovative

Reactor Concepts for Various

Missions” and consisted in two

lectures. The first one entitle “The

Deciding Factors in Opting for a

Particular Reactor Technology” was

given by J. Guidez (CEA, France) and

the second one entitled “Promising

Reactor Concepts for Multiple Applications”

given by S. Monti (IAEA).

The second block was focused on the

topic “Near-term-deployment Powerto-grid

Integrated LWR Technology”

and it consisted of three lectures as

follows. The first lecture deals with

“From Specific Design Criteria to an

Advanced Modular LWR Concept: Approach,

Methods, Validation” given by

F. Morin (CEA, France). The second

lecture was entitled “The Challenges

of Designing and Licensing a First-ofa-kind

Reactor Prototype, even a Small

One” given by D. Delmastro (CNEA,

Argentina) and finally the third lecture

was entitled “The Reliability and

Safety Case of a Reactor Equipped

with Passive Systems” presented by

A. Schaffrath (GRS, Germany). The

third block was dedicated to the topic

“Multi-purpose Molten Salt Reactors”

and it was organized as three

lectures to the following issues. The

first lecture was entitled “MSR Design

Principles, Concepts, Modelling Approaches,

and Methods” given by T.

Abram (Univ. of Man chester, UK). The

second lecture was devoted to “From

the MSR Physics Principles to a Plant

Layout” given by E. Merle (Grenoble

INP & CNRS) and the third lecture

entitled “Fuel Salts Chemistry and

Materials Compati bility” given by

V. Ignatev (KI, Russia). The fourth

block was devoted to “Nuclear Technology

for Space Propulsion and

Manned Space Exploration” and it

consisted of two lectures. The first one

entitled “ Nuclear Rocket Propulsion:

Background, Physics and Methods,

Design and Tests” given by W. Emrich

Jr. ( NASA, USA) and the second one

entitled “The Challenge of Fueling a

Nuclear Engine for Space Exploration”

given by J. Witter (BWXT, USA).

The fifth block was devoted to

“Minimum-intervention Long-life

Breed-and-burn Fast Reactors” and

it consisted of two lectures. The first

one entitled “Physics of Breed-andburn

Reactors, Optimized Core and

Fuel Design, Licensing Case” given by

K. Weaver (INL, USA) and the second

one entitled “Cladding and Structural

Materials for Very Long In-core

Residence Times“ given by Y. Decarlan

(CEA, France). The sixth block was

devoted to “Reactor Concepts for

Process Heat and Power-to-gas

Applications” and it consisted of two

lectures. The first one entitled

“ Designing a Small Reactor to Bring

Power to Remote Areas or to Produce

Process Heat” given by J. Kloosterman

(TU Delft, Netherlands) and the

second one entitled „Techno-economic

Assessment of Hydrogen Production

from Nuclear Energy“ given by

J. Witter (BWXT, USA). In the frame of

the technical visit, a guided tour was

organized for the TrasnetBW GmbH

in Wendlingen, which is one of the

largest system Control Centre for

Baden Württemberg with headquarter

in Stuttgart. It operates the electricity

transmission grid in the German state

of Baden-Württemberg, control and

monitor the energy flows through the

grid, and perform the necessary

maintenance and network planning

and development activities. This year,

44 participants of 18 countries (EU,

Asia, Latin America, East Europe,

Middle East, Africa, and USA) attended

the FJOH Summer School.

Recognized experts of 10 different

countries from Asia, EU, South

America, USA and East Europe from

Academia, industry, research and

TSOs gave high-level lectures on topics

of their expertise. During the ten days,

the participants had the opportunity

to exhaustive discussions with the

lecturers and other participants. Apart

| The 25 th Edition of the Frédéric Joliot/Otto Hahn Summer School on Nuclear

Reactors “Physics, Fuels and Systems“, participants and lecturers in Karlsruhe.

from the technical issues, another goal

of the FJOH Summer School is to

intensify the networking among the

participants of different continents

and nationalities with the common objective

of enhanced safety worldwide.

A well-recognized tradition during

the FJOH Summer School is an extensive

and diverse program with social

events to get familiar with the German

culture and way of life as well as to

foster the exchange among the participants.

This year, the Sunday trip

consisted in the visit of the Technical

Museum Speyer and afterwards, a

Canoe Tour at the Old Rhine. After the

Museum visit, the participants have

free time to get familiar with the

downtown of Speyer and to visit the

Romanesque Cathedral that houses

the grave f most important kings and

emperors e.g. it is the burial place for

emperor Konrad II and his wife.

The next Summer School will be

hosted by CEA in Aix-en-Provence,

from August 26 th to September 4 th ,

2020 and it will be devoted to “High-

fidelity Modelling and Simulation

of Nuclear Reactors: Turning a

Promise into Reality”.

Author

Dr.-Ing. Victor Hugo Sanchez-Espinoza

Head of Group

“Reactor Physics and Dynamics”

Project Leader

“LWR Safety Methods and Codes ”

Karlsruhe Institute of Technology (KIT)

Institute for Neutron Physics and

Reactor Technology (INR)

Hermann-von-Helmholtz-Platz 1

76344 Eggenstein-Leopoldshafen

Report

The 25th Edition of the Frédéric Joliot/Otto Hahn Summer School on Nuclear Reactors “Physics, Fuels and Systems“ ı Victor Hugo Sanchez-Espinoza



550

SPECIAL TOPIC | A JOURNEY THROUGH 50 YEARS AMNT

Am 7. und 8. Mai

2019 begingen wir

das 50. Jubiläum

unserer Jahrestagung

Kerntechnik. Zu

diesem Anlass öffnen

wir unser atw-Archiv

für Sie und präsentieren

Ihnen in jeder

Ausgabe einen

historischen Beitrag.

Überarbeitete

Fassung eines

Vortrags gehalten

am 14. Mai 2002

auf der Jahrestagung

Kerntechnik 2002,

Stuttgart,

14.-16.5.2002

Schutz von Mensch und Umwelt –

Nukleare Anwendungen außerhalb des

Energiesektors

P.P. De Regge, Wien

1 Einleitung Nukleare Technologien liefern immer mehr bedeutende Beiträge zum Schutz des Menschen und

seiner Umwelt sowie zur Verbesserung der Lebensstandards. Die Internationale Atomenergie Organisation (IAEO) ist

durch die Vereinten Nationen (UN) per Statut beauftragt, solche nuklearen Technologien zu entwickeln, zu fördern

und deren Einsatz zu unterstützen. Sie erreicht dies durch Veröffentlichung von Informationen, Ausbildung, Zulieferung

von Ausrüstung, Dienstleistungen und Erstellung von Sicherheitsnormen.

In diesem Beitrag wird diese Rolle der IAEO erläutert

und es wird weiter der heutige Stand der nuklearen

Anwendungen in folgenden Bereichen dargestellt:

p Nuklearmedizin und Gesundheitspflege

p Veterinärmedizin und Viehzucht

p Bodenkultivierung und Düngemittel

p Wasserversorgung und Umweltschutz

p Anbau von Kulturpflanzen

p Schädlingsbekämpfung

p sonstige Anwendungen, wie Landminenräumung und

Schutz des kulturellen Erbes

2 Rolle der Internationalen Atomenergie

Organisation

Die Ziele der IAEO hinsichtlich der Anwendung von

nuklearen Technologien sind ein Beitrag zur nachhaltigen

Entwicklung und zum Umweltschutz in den Bereichen

medizinische Versorgung, Ernährung, Landwirtschaft und

Industrie sowie Versorgung mit Wasser. Die Aktivitäten

der IAEO in diesen Bereichen haben Grundbedürfnisse der

Menschheit im Fokus. Nukleare Technologien sollen dabei

wesentlichere und konkurrenzfähigere Vorteile bieten als

vergleichbare andere Technologien.

Entsprechende nukleare Technologien und Anwendungen

werden im Rahmen von koordinierten, angewandten

Forschungsprojekten entwickelt. Dazu werden

über mehrere Jahre Beiträge von potenziellen Anwendern

aus Industriestaaten und Entwicklungsländern koordiniert.

Bereits einsetzbare Technologien werden verfügbar

gemacht und interessierten Ländern bereit gestellt; über

technischer Kooperationsprojekte wird die technische

Ausrüstung beschafft und es wird für eine entsprechende

notwendige Ausbildung zum Betrieb von Analgen und

Einsatz von Technologien gesorgt.

Im derzeitigen Gesamtbild der weltweiten Entwicklung

erkennt man folgende wichtige Problemfelder:

p Die Weltbevölkerung hat fast 6 Milliarden Menschen

erreicht und wächst pro Jahr um rund 80 Millionen

an – dies entspricht etwa der deutschen Gesamtbevölkerung.

p Fast eine Milliarde Menschen sind chronisch unterernährt

und es fehlt ihnen zudem an einer zuver lässigen

Wasserversorgung.

p Sechs Millionen Kinder in der Dritten Welt sterben

jedes Jahr aufgrund Mangelernährung; ebenso viele

Erwachsene sterben in den Industriestaaten jährlich an

Krebs. Weitere Millionen sind erkrankt oder sterben an

bakteriellen oder von Viren verursachten Infektionen

und Krankheiten. Zum Beispiel streben allein an den

Folgen der Malaria weltweit mehr als 2 Millionen

Menschen pro Jahr.

Viele heutige Aktivitäten des Menschen verschmutzen und

verändern die Umwelt und sind auf längere Sicht nicht

tragbar.

Die von der IAEO entwickelten und unterstützten

nuklearen Anwendungen außerhalb des Energiesektors

haben daher auch

p die Verbesserung der Ernährungssituation und Wasserversorgung,

p die Verbesserung des Gesundheitsschutzes sowie

p den Schutz der Umwelt durch Analyse, Vorbeugung

und Sanierung belasteter Bereiche

zum Ziel. Trends der IAEO-Unterstützung von Projekten

im Rahmen der technischen Kooperation sind in Tabelle 1

dargestellt.

3 Gesundheitsvorsorge

Die meist verbreiteten diagnostischen Anwendungen von

ionisierender Strahlung und nuklearer Technologie sind

die Röntgendiagnose sowie die Diagnostik mit radioaktiven

biologischen Indikatoren in der Nuklearmedizin.

Die therapeutische Anwendung ist die Strahlungsonkologie,

bei der mit einer wirksamen Strahlungsdosis an

bestimmten krebsgefährdeten Stellen des Körpers mittels

einer externen Strahlungsquelle (Teletherapie) oder

mittels einer implantierten Strahlungsquelle (Brachytherapie)

gearbeitet wird. Auch offene Strahlungsquellen,

wie radioaktive Substanzen zur Behandlung von Schilddrüsenkrebs,

werden manchmal verwendet. Neue Anwendungen

von radioaktiven Substanzen werden erforscht

und entwickelt.


Protection of Man and Environment – Nuclear Usage Outside of Energy Sector ı P.P. De Regge, Wien


Projekte

Eine revolutionäre Entwicklung der diagnostischen

Radiologie war die Computertomografie, die genaue

Schnittbilder von bestimmten Körperregionen in allen

anatomischen und funktionellen Einzelheiten darstellt.

Die Nuklearmedizin verwendet Radioisotope in über

Hundert standardisierten Methoden für diagnostische,

therapeutische und Forschungszwecke in den Bereichen

Onkologie, Endokrinologie, Kardiologie, Neurologie und

Nephrologie. Industriestaaten haben derzeit etwa

20 Gammakameras pro einer Million Einwohner zur

Verfügung, während in Entwicklungsländern nur eine

pro eine Million Einwohnern verfügbar ist, und dies

nur in den größeren Metropolen.

Molekulare Nuklearmedizin ist ein neuer Anwendungsbereich.

Erkrankungen werden auf zellulären oder

gene tischem Niveau untersucht und identifiziert. Polymerase

Kettenreaktionen bilden extrem empfindliche

Diagnosetechniken für epidemische Infektionen, Krebs

oder weiterer Symptome und sind z. B. von der US Food

and Drug Administration als in vitro Tests genehmigt

worden.

Radioimmunoassay ist eine der wichtigsten Komponenten

der in vitro Diagnosetechniken für die Quan tifizierung

von Proteinänderungen oder anderer erkrankungsbedingter

Metabolismusprodukte. Ein Zuwachs des

Weltumsatzes für diese in vitro Diagnosetechniken von

20 Milliarden Dollar im Jahr 1999 auf bis zu 26 Milliarden

Dollar im Jahr 2004 wird prognostiziert und spiegelt die

Bedeutung dieser Diagnosemöglichkeit wieder.

Die Anzahl von Krebserkrankungen, eine der häufigsten

Todesursachen in den Industriestaaten, nimmt auch aufgrund

der steigenden Lebenserwartung in den Ländern

der Dritten Welt ständig zu. Von etwa 10 Millionen neuen

Krebsfällen im Jahr 2000 entfällt jeweils die Hälfte auf

Industriestaaten und auf Entwicklungsländer. Es wird

erwartet, dass 10 Millionen von 15 Millionen neuer

Jahr

1980 1985 1990 1995 2000

Kernenergie 26 10 10 5 5

Gesundheitsvorsorge 11 11 11 16 23

Physik und Chemie 25 25 20 15 10

Hydrologie 2 3 5 6 11

Landwirtschaft und Ernährung 24 25 22 24 18

Industrie 5 12 12 8 8

Strahlenschutz 6 12 17 24 20

Abfallverarbeitung 1 2 3 2 5

| Tab. 1.

IAEO Unterstützung von Projekten im Rahmen der Technischen Kooperation (Prozent der Gesamtfinanzierung)

| Teletherapie.

Krebsfälle im Jahr 2015 in Ländern der Dritten Welt auftreten

werden. Fünfzig Prozent der Krebspatienten werden

mittels Radiotherapie behandelt, mit einer Überlebenschance

von ungefähr 45%.

Am meisten wird noch Teletherapie verwendet, jetzt

mit Präzisionsbestrahlung von komplexen und unregelmäßigen

Tumorformen. Für Brachytherapie mittels

implantierter Strahlungsquellen werden derzeit Caesium-

137 und Iridium-192 anstelle von Radium-226 verwendet.

Die IAEO, zusammen mit der Weltgesundheitsorganisation

(World Health Organisation, WHO), verwaltet

eine Datenbank von Spitälern und Kliniken, in denen

Radio therapie zur Anwendung kommt. Länder der

Dritten Welt mit 85 % der Bevölkerung verfügen über

ein Drittel der Radiotherapieinstallationen, etwa 2200

Teletherapie geräte mit Kobalt-60 Quellen und 850

Brachytherapie geräten allerdings nur einem Fünftel der

Elektronen beschleuniger. Mit jedem Gerät können

600 Patienten pro Jahr behandelt werden oder insgesamt

nur 1,9 Millionen von 2,5 Millionen Patienten in diesen

Ländern. Mehr als 5000 zusätzliche Radiotherapiegeräte

werden daher bis zum Jahr 2015 gebraucht werden. Am

Rande sei erwähnt, dass die IAEO und die WHO jährlich

die ordnungsgemäße Anwendung sowie die eingesetzte

Strahlungsdosis von etwa 600 dieser Geräte in der Dritten

Welt überprüfen.

Nukleare Diagnose- und Forschungtechniken auf Basis

stabiler oder radioaktiver Isotope werden auch zu Maßnahmen

auf dem Ernährungssektor verwendet, insbesondere

zum Nachweisen von Vitamin-, Spuren element- und

Mikronährstoffmangel. Diese sind mit Ursache für Anämie,

Beeinträchtigungen des Sehver mögens, Wachstum und

geistige Entwicklung. Differen tielle Röntgenabsorp tiometrie

wird derzeit auch als nicht intrusive Methode zur

Knochendichtemessungen verwendet.

4 Landwirtschaft

Nukleare Technologien werden in der Landwirtschaft zur

Verbesserung und zum Schutz von Nahrungsmitteln

verwendet sowie zur Optimierung der Viehzucht und

Vorbeugung gegen Seuchen und zur Bekämpfung von

Schädlingen. Biodiversität ist entstanden durch zufällige

Mutationen, induziert durch kosmische und natürliche

Strahlung und auch durch Transkriptionsfehler im Erbgut.

Durch Anwendung von Strahlung kann deshalb die

Mutationsfrequenz erhöht und beschleunigt werden

und man kann in kurzer Zeit eine große Anzahl von

neuen Pflanzenvarianten bilden. Die meisten erzeugten

Mutationen sind nicht weiter lebensfähig oder sogar

schlechter überlebensfähig als die ursprüngliche Pflanze,

aber einige sind unter Umständen besser angepasst und

werden daher weiter gezüchtet. Einige Beispiele werden

551





552


| Züchtungen von Pflanzen:

Höhere Erträge durch den Einsatz von Bestrahlungen.

hier angegeben, da sie für die betreffenden Länder und

Regionen eine erhebliche Bedeutung in der zukünftigen

Versorgung besitzen. Reissorten, geeignet zum Wachsen

im Wechsel von Salzwasser und Süßwasser der asiatischen

Flussdeltas, dienen als Grundnahrungsmittel von

Millionen Menschen in Pakistan, China und Bangladesh.

Gerstevarianten ohne Hülse, dürretolerantes Sorghum

und Reis werden durch strahlungsinduzierte Mutationen

hergestellt und gezüchtet, weil sie den klimatologischen

und geologischen Bedingungen in Peru oder auch in Mali

besser angepasst sind. Am besten eignen sich strahlungsinduzierte

Mutationen für die Optimierung von Pflanzen,

die sich durch sterile Klonung vermehren, wie Bananen.

Eine früh blühende, herrlich schmeckende und seuchenresistente

Variante Novaria wurde auf diese Art in Malaysia

entwickelt. Seuchenresistente Sorten bringen nicht nur

bessere Ausbeute sondern benötigen auch weniger oder

gar keinen Pestizideinsatz und sind deshalb sowohl billiger

als auch umweltschonender. Bananen und Reis sind die

wichtigsten Grundnahrungsmittel und Landwirtschaftsprodukte

in der Dritten Welt, aber auch in der Industriestaaten

werde fast zweitausend durch strahlungsinduzierte

Mutationen verbesserte Produkte konsumiert.

In Österreich wurde aus Golden Delicious die Variante

„Golden Haidegg” gezüchtet, mit einer schöneren Farbe

und längerer Haltbarkeit ohne Rostflecken. Japanische

Birnen mit strahlungsinduzierter Resistenz gegen

Schwarzfleckenseuche werden derzeit gezüchtet. Dies

reduziert den früher notwendigen Pestizideinsatz auf ein

Viertel.

Nukleare Techniken werden nicht nur zur Verbesserung

der Nahrungsmittel sondern auch zu deren Schutz

und Konservierung verwendet. In 30 Ländern wird

Bestrahlung von Nahrungsmitteln zur Gesundheits- und

| Bestrahlung von Nahrungsmitteln.

Qualitätssicherung genutzt sowie zur Einhaltung der

Quaran täneverordnungen für bestimmte Nahrungsmittel,

wie Fleisch, Früchte, Kräuter und getrocknete Gemüse.

Seit Mitte 2000 wird die Bestrahlung von Hackfleisch in

den Vereinigten Staaten aus mikrobiologischen Sicherheitsgründen

eingesetzt. In mehr als 2000 Supermärkten

werden diese Produkte ohne signifikante Vorbehalte der

Verbraucher angeboten. Viele Länder, insbesondere in

Asien, Lateinamerika, dem Mittler Osten und Afrika

würden enorme Nutzen durch der Anwendung dieser

Technologie zum Schutz der Nahrungsmittel erzielen.

Die weltweit internationale Autorität für die Sicherheit

der Nahrungsmittel, die Kodex Alimentarius Kommission,

hat auch im Bereich der Strahlenbehandlung von Nahrungsmitteln

Protokolle, Richtlinien und Empfehlungen

veröffent licht.

Im Landwirtschaftsbereich findet man auch die Anwendung

von Radioisotopen durch den Radioimmunoassay

bei Tieren für Hormonanalysen von Milch, Serum oder

Plasma. Zweck ist die Optimierung der Planung und

Diagnose von der Trächtigkeit, die frühzeitige Erkennung

von Gesundheits- und Reproduktionsmängeln und die

zeitgerechte Identifikation und Vorbeugung vor Seuchen,

wie Rinderpest, Trypanosomosis, Brucellosis und Maulund-Klauen-Seuche.

Nukleare Technologien und Isotopmarkierungen

mit Stickstoff, Kohlenstoff und Phosphor

finden auch verbreitet Anwendungen zur Untersuchung

von biologischer Stickstofffixierung durch Bakterien, zur

Optimierung der Anwendung von Düngemitteln nach

Menge, Art und Jahreszeit in den verschiedensten Klimazonen,

Bodentypen und Ernährungskulturen sowie zur

Minimierung der ungenutzten Phosphat- und Stickstoffmengen

in der Umwelt.

| Insektensterilisierung: Schädlingsbekämpfung.

Beim Ausrotten von Schädlingen werden Nukleare

Technologien unter der Bezeichnung „Sterile Insekten-

Technik“ verwendet. Diese Technik beinhaltet die

großräumige Produktion und systematische Freisetzung

von strahlungssterilisierten männlichen Fruchtfliegen,

Tsetse-Fliegen, Holzwürmer usw. Freigelassen, verhalten

sie sich auf ganz normale Weise, aber produzieren keinerlei

Nachwuchs. Die schon geschwächte nächste Generation

wird wieder mit neuen sterilen Männchen versorgt,

bis sie lokal gezielt ausgerottet ist, ohne Pestizide, ohne

Umwelt zerstörung und unter Erhalt des biologischen

Gleich gewichtes. Insbesondere die Nachfrage und

Produktionskapazität für sterilisierte Mittelmeerfruchtfliegen,

notorische Schädlinge mit erheblichen Auswirkungen,

sind im Lauf der letzten Dekade von einer auf

drei Milliarden Fliegen pro Woche gestiegen. Am meisten

Erfolg hatten die Ausrottungskampagnen in Mittel- und




| Insektensterilisierung: Schädlingsbekämpfung.

Lateinamerika, den Südlichen Vereinigten Staaten,

Australien und Japan. Subtile genetisch eingebaute Unterschiede

erlauben die Sterilisierung und Aussortierung der

männlichen Insekten schon im Puppenstadium. Dementsprechend

werden die Zuchtkosten gesenkt und die

Effektivität wird erhöht. Auf diese Weise wird diese Technologie

gleich einem selektiven Insektizid zur kompletten

Ausrottung verwendet. Im Rahmen eines Projektes des

IAEO Programms zur „Technischen Kooperation“ wurde in

Tansania eine Produktionsanlage für Millionen sterilisierte

Tsetse Fliegen errichtet, womit verschiedene afrikanische

Regionen von diesen Schädlingen befreit werden konnten.

Derzeitige Forschungsprojekte in diesem Bereich entwickeln

weitere sterile Insekten-Techniken für Motten und

Würmer. Die IAEO hat ein ehrgeiziges Projekt für die

Ausrottung von Malaria Moskitos gestartet.

Die wirtschaftlichen Erfolge dieser Technologie sind

insgesamt herausragend. Die Bekämpfung von Fruchtfliegen

in Mexiko mit einem Kapitaleinsatz von

10 Millionen Dollar pro Jahr bewirkt einen Ertrag von

einer Milliarde Dollar Wert an Zitrusfrüchten und Gemüse.

Ähnliche Ergebnisse wurden in Chile erzielt, insbesondere

da die Zitrusfrüchte dort während des nördlichen Winters

produziert werden und somit einen maßgeblichen Einfluss

auf den Export haben.

5 Hydrologie

Nukleare und Isotopentechniken werden auch in der

Hydrologie und Klimatologie genutzt, wo sich aufgrund

der natürlich entstandenen Fingerabdrücke der Isotopenzusammensetzung

die Herkunft und das Alter der Wasservorräte

nachweisen lässt. Isotopentechniken, basierend

auf Messungen von stabilen und radioaktiven Wasserkomponenten,

werden zur Modellierung von Wassersystemen

verwendet, um die in vielen Regionen

| Isotopentechniken in der Hydrologie: Wasservorräte erweitern.

beschränkten und kostbaren Wasserressourcen optimal zu

nutzen und ein lokales Gleichgewicht zwischen verbrauchten

und erneuerbaren Wassermengen zu erzielen.

Diese Methoden werden auch zur Identifizierung und zum

Nachweis von unerwünschten Vermischungen der Wasserversorgung

mit Abwässern verwendet, wie es in dichtbesiedelten

Großmetropolen öfter vorkommt. Die

Temperaturabhängigkeit der Wasserisotopenzusammensetzung

beim Verdampfen und Kondensieren wird in

weltweiten Forschungen von Klima- und Treibhauseffekt

genutzt. So ist die Klimageschichte von zehntausenden

Jahren aufgrund der Isotopenanalysen von Gletschereisschichten

oder antarktischen Ablagerungen rekonstruiert

worden.

Beispiele zur Angabe der wirtschaftlichen und humanitären

Bedeutung dieser nuklearen Techniken in der

Hydrologie sind:

p In Venezuela wurde das Trinkwasserdefizit um 30 %

verringert durch Inbetriebnahme von 50 mit Isotopentechniken

georteten Frischwasserquellen.

p Durch Grundwassermessungen im Niltal und in den

ägyptischen und äthiopischen Wüsten wurden erneuerbare

Wasserreserven in Nubischen Sandsteinschichten

geortet womit die Wasserversorgung um 20 % erhöht

werden konnte.

p Ebenfalls in Venezuela wurde aufgrund von Isotopenmessungen

die Zuverlässigkeit eines Stauseedammes

nachgewiesen. Eingeplante Reparaturkosten in

Höhe von 6 Mio. US-Dollar konnten somit eingespart

werden.

p Mit dem Ausbau von geothermischen Energiequellen

werden in Mittelamerika und Südostasien Millionen

Dollar an Ölimporten gespart. Isotopentechniken

dienen dabei zur Suche von geeigneten Standorten und

zur Optimierung der Anlageneinrichtungen.

6 Umweltschutz

Nukleare Technologien und Messtechniken werden weit

verbreitet zur Überwachung und Erforschung der Umwelt

eingesetzt. Überhaupt werden künstliche und natürliche

Radioisotope in der Atmosphäre, in der Hydrosphäre und

im Boden für radiologische Zwecke überwacht. Sie sind

zudem ausgezeichnete Indikatoren für atmosphärische

und ozeanografische Transportprozesse, für das Verhalten

und den Transport von nicht radioaktiven Umweltgiften,

für die geochronologische Altersbestimmung und für

ökologische und biologische Forschungszwecke. Einfache

tragbare Röntgenfluoreszenzgeräte mit Radioisotopenquellen,

Forschungsreaktoren für die Neutronenaktivierungsanalyse

und Teilchenbeschleuniger zur

Charakterisierung von mikroskopischen Teilchen formen

zusammen ein Arsenal von nuklearen Techniken zum

Identifizieren von schädliche Substanzen in der Umwelt

und zur genauer Identifizierung ihrer Zusammensetzung

und Herkunft.

Nicht nur zur Diagnose von schädlichen Substanzen,

sondern auch zur Vorbeugung oder Eliminierung ihrer

Effekte werden nukleare Technologien verwendet.

Elektronenbeschleuniger werden in Kohlekraftwerken

verschiedener Länder zur Abgasbehandlung eingesetzt.

Schwefel- und Stickstoffoxiden werden im Rauchgas

umgewandelt und mit zugefügtem Ammoniak zu Düngemittel

umgesetzt. Abwässer, organisch verseuchtes Wasser

und Klärschlamm werden mit ionisierender Strahlung

behandelt und desinfiziert.

Zur Umweltsanierung gehört auch die Ortung und

Entfernung von rund 60 Millionen Landminen, die in

553





554


| Lokalisierung von Landminen.

62 Ländern Jahr für Jahr Ursache für Tausende Tote und

Verletzte sind. Nukleare Technologien werden auch für

diese Anwendungen entwickelt. Derzeit verfügbare Geräte

benutzen Neutronengeneratoren und Gammadetektoren

zur Auswertung von Elementverhältnissen, insbesondere

Wasserstoff, Stickstoff, Kohlenstoff und Sauerstoff, in

minenverdächtigen Gegenständen im Boden.

7 Industrie und Archäologie

Unzählige Anwendungen von nuklearen Technologien,

ionisierender Strahlung und Isotopen werden auch in

allgemeinen Bereichen der Industrie genutzt, so in

Transport- und Bauunternehmen für radiografische

Materialkontrolle, in der Polymerverarbeitung (z. B.

Autoreifen), für die Sterilisierung sowie das Vernichten

von Viren, Bakterien und Pilzen im medizinischen

Bereich.

Minuten quadratmillimeter kleine Einzelheiten von

Gemälden, Skulpturen, Polychromen, Münzen oder

Keramiken zu charakterisieren und auszuwerten. Ein

Prototyp kam im Kunsthistorischen Museum in Wien für

die Identifizierung von Gemälden des 16ten Jahrhunderts

und zur Katalogisierung von Etruskischen Bronzen und

Münzen zum Einsatz.

8 Zusammenfassung

Nukleare Technologien, Isotopentechniken und Strahlungsanwendungen

sind weit verbreitete Anwendungen

und liefern unersetzbare Beiträge zur Verbesserung und

Erhaltung des heutigen Lebensstandards sowie zum

Schutz des Menschen und seiner Umwelt. Das Ausmaß

dieser Anwendungen für diagnostische und thera peutische

Zwecke im Gesundheitsbereich wurde dargestellt und

Erfolge und Erwartungen der Entwicklungen von

Computertomografie und molekularer Nuklearmedizin

wurden erwähnt. Im therapeutischen Bereich ist ohne

Zweifel eine Erweiterung der Anwendungen zu erwarten,

inbesondere in der Dritten Welt. Anwendungen in der

Landwirtschaft, der Nahrungsversorgung und dem Schutz

der Nahrung, wie Pflanzenoptimierung, Ausrottung von

Schädlingen, Vorbeugung vor Seuchen und Konservierung

von Nahrungsmittel sind vollständig ausgereift und

werden akzeptiert. Bei komplexeren Problemen kommen

ebenfalls erfolgreiche Maßnahmen zum Einsatz, wie

im Zusammenhang mit der Bekämpfung von Malaria,

Trypanosomosis und Myasis. Anwendungen in der Hydrologie

und im Umweltschutz sind seit Jahren etabliert und

anerkannt. Im Bereich der Klimatologie und in der meeresoder

landesbezogenen Ökologie werden sie erweitert

genutzt. Nukleare Techniken und Technologien in den

hier beschriebenen Anwendungsbereichen werden nicht

aus reinem Interesse an der Sache oder aufgrund

popukärer Überlegungen eingesetzt, oder weil besondere

Förderungen zu erwarten wären, sondern weil sie nach

vergleichender Begutachtung und Kosten-Nutzen-Analyse

optimalere Ergebnisse versprechen als vergleichbare

verfügbare Technologien.

Literatur

1. Annual Report 2000. International Atomic Energy Agency, Vienna, 2001

2. Nuclear Technology Review 2002. Document GOV/2002/7, International Atomic Energy Agency,

Vienna, February 2002

3. The International Atomic Energy Agency‘s Laboratories, Meeting the Challenges of Research and

International Cooperation in the Application of Nuclear Techniques. Document IAEA/PI/A67E

99-02059, August 1999

| Kulturelles Erbe: Analysen zur Bestätigung der Herkunft.

Interessant und vielleicht weniger bekannt ist die

Anwendung von zerstörungsfreien nuklearen Technologien

und Messmethoden im Bereich der Archäologie und

der Konservierung des Kulturellen Erbes. Vor kurzer Zeit

hat die IAEO ein Projekt in Lateinamerika zur Förderung

von interdisziplinärer Forschung und Unterstützung in

diesem Bereich koordiniert. Jedes teilnehmende Land

hat sich mit bestimmten ungelösten Fragen bezüglich

Keramiken befasst und mittels Spurenanalyse durch

Röntgenfluoreszenz- und Neutronenaktivierung das

erwünschte Ergebnis endgültig erhalten.

In Zusammenarbeit mit polnischen, kubanischen und

österreichischen Wissenschaftlern hat die IAEO ein tragbares

Röntgenfluoreszenzgerät entwickelt, das ideal für

die zerstörungsfreie Analyse von Kunstgegenstände

eingesetzt werden kann. Es ist mit Computer- und Lasersteuerung

ausgestattet und in der Lage, innerhalb von

Verfasser

Peter P. De Regge

Head, PCI Laboratorien,

Internationale Atomenergie Organisation (IAEO),

Wagramer Straße 5, 1400 Wien, Austria




Inside

Women in Nuclear bei Framatome/ANF in Karlstein

Die deutsche Sektion des international agierenden Frauennetzwerks

WiN (Women in Nuclear) traf sich am 10. und

11. Oktober 2019 bei Framatome in Karlstein. Nach der

Mitgliederversammlung von WiN Germany wurde das

Treffen mit Teilnehmerinnen aus Schweden, Finnland und

der Schweiz international.

Unter dem Dach des gemeinnützigen Vereins Women in

Nuclear (WiN) Germany – gegründet 2008 – möchten

Frauen, die auf den Gebieten Kernenergie, Strahlenschutz,

Nuklearmedizin und nukleare Wissenschaften arbeiten,

dazu beitragen, einen transparenten Dialog über die Notwendigkeit

der nuklearen Kompetenzen für Deutschland

zu führen.

Chantal Greul, Präsidentin von WiN Germany, erklärt:

„Auch wenn sich unser Land mehrheitlich für den Ausstieg

aus der Stromerzeugung durch Kerntechnik entschieden

hat, bleiben das Know-how und das hohe Sicherheitsverständnis

Deutschlands nach dem Ausstieg 2022

wichtig: für den Rückbau, die Endlagerung – und nicht

zuletzt als Beitrag zum Erhalt des höchstmöglichen Sicherheitsniveaus

weltweit. Wir ermutigen junge Frauen, eine

berufliche Laufbahn in der Nukleartechnik zu wählen.

Denn nur durch gut ausgebildete Nachwuchskräfte bleiben

die notwendigen Ressourcen in Deutschland langfristig

gesichert.“

| Im Foyer der Framatome GmbH in Karlstein mit den Standortleitern

Stefan Rosenberger der Framatome GmbH (links) und Matthias Gutjahr

der Advanced Nuclear Fuels GmbH (Mitte).

| v.l. Chantal Greul, WiN Germany Präsidentin,

Preisträgerin Jenny Jessat, Martina Ezmuß,

Vorstand WiN Germany.

Einen wichtigen Beitrag zu diesem Vereinsziel leistet

die jährliche Verleihung des WiN-Germany Preises. Für

diesen mit 500 Euro dotierten Preis gab es in diesem Jahr

gleich fünf Bewerberinnen. Beiträge aus Wissenschaft und

Forschung sowie aus Unternehmen ließen eine hohe

fachliche Kompetenz erkennen, die die Entscheidung für

eine Bewerberin schwer fallen ließ. Jenny Jessat überzeugte

den Kreis mit dem Thema „Studies on the interaction

of plant cells with uranium (VI) and europium (III)

an on stress-induced metabolite release” („Studien zur

Wechselwirkung von Pflanzenzellen mit Uran (VI) und

Europium (III) und zur stressinduzierten Metabolitfreisetzung“).

Es handelte sich um die Präsentation

ihrer Masterarbeit in Chemie an der TU Dresden. Jenny

Jessat ist derzeit wissenschaftliche Mitarbeiterin am

Helmholtz-Zentrum Dresden-Rossendorf, Institut für

Ressourcenökologie und bereitet dort ihre Promotion vor.

Neben weiteren Fachvorträgen über kerntechnische

Themen fanden die Werksführungen sehr viel Anklang bei

den Teilnehmerinnen. „Wir möchten uns herzlich bei den

beiden Standortleitern der Framatome GmbH und ANF

bedanken. Insbesondere die Standortbesuche waren ein

wirkliches Highlight und die Teilnehmerinnen waren

damit sehr zufrieden“, lobte Chantal Greul den Einsatz der

Framatome für die Veranstaltung.“

| Aus der Komponentenfertigung der ANF

Karlstein stellte Petra Denner die

Entwicklungs historie des neuen GAIA

Abstandhalters vor und zeigte an diesem

Beispiel auf, wie die hohen Qualitätsanforderungen

der Kunden erfüllt werden.

Karin Reiche

WiN

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

555

KTG INSIDE

Herzlichen Glückwunsch!

www.ktg.org

Die KTG gratuliert ihren Mitgliedern sehr herzlich zum Geburtstag

und wünscht ihnen weiterhin alles Gute!

Dezember 2019

50 Jahre | 1969

13. Bernd Gulich, Tiefenbach-Ast

55 Jahre | 1964

30. Thomas Schmidt, Lörrach

60 Jahre | 1959

15. Axel Lenzen, Titz

20. Martin Schlieck-Weber, Hausen

65 Jahre | 1954

24. Reinhold Paul, Hanau

70 Jahre | 1949

2. Dipl.-Ing. Berndt Standfuß, Dresden

28. Fritz Grimm, Alzenau

76 Jahre | 1943

8. Dr. Dieter Herrmann, Brandis

77 Jahre | 1942

8. Karl Georg Weber, Neckarwestheim

14. Günter Breiling, Weinheim

78 Jahre | 1941

13. Dipl.-Ing. Klaus-Dieter Hnilica,

Rodenbach/Hanau

Wenn Sie künftig eine

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Geburtstages in der

atw wünschen, teilen

Sie dies bitte der KTG-

Geschäftsstelle mit.

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556

79 Jahre | 1940

8. Dipl.-Ing. Wolfgang Heess, Laudenbach

21. Dr. Jürgen Wehmeier, Springe

94 Jahre | 1925

10. Dr. Arthur Pilgenröther, Kleinostheim

83 Jahre | 1937

9. Dipl.-Ing. Werner Rossbach,

Bergisch Gladbach

NEWS

79 Jahre | 1940

16. Dipl.-Ing. Wolfgang Breyer, Buckenhof

19 Prof. Dr. Wernt Brewitz, Wolfenbüttel

80 Jahre |1939

1. Dipl.-Ing. Georg Dumsky, Gräfelfing

6. Dipl.-Ing. Hans-Henn. Kuchenbuch,

Laboe-Brodersdorf

27 Dr. Horst Bauer, Sigless/AT

81 Jahre | 1938

1. Dr. Gert Spannagel,

Linkenheim-Hochstetten

82 Jahre | 1937

30. Dipl.-Ing. Wilhelm Weiss, Weinheim

83 Jahre | 1936

7. Dipl.-Ing. Aurel Badics, Bad Kreuznach

17. Prof. Dr.-Ing. Rolf Theenhaus, Linnich

86 Jahre | 1933

10. Prof. Dr. Jürgen Vollradt,

Unna-Königsborn

Januar 2020

55 Jahre | 1965

31. Eckhard Stengert, Worms

60 Jahre | 1960

26. Dr. Friedhelm Funke, Dormitz

70 Jahre | 1950

15. Dipl.-Ing. Andreas Hüttmann, Oering

78 Jahre | 1942

31. Dipl.-Phys. Werner Scholtyssek,

Stutensee

79 Jahre | 1941

15. Dipl.-Ing. Ulf Rösser,

Heiligkreuzsteinach

81 Jahre | 1939

16. Dr. Wolfgang Kersting, Blieskastel

82 Jahre | 1938

7. Dipl.-Ing. Manfred Schirra, Stutensee

12. Dipl.-Ing. Hans Dieter Adami, Rösrath

22. Dr. Franz Müller, Erlangen

84 Jahre | 1936

5. Obering. Peter Vetterlein, Oberursel

23. Prof. Dr. Hartmut Schmoock,

Norderstedt

30. Dipl.-Phys. Wolfgang Borkowetz,

Rüsselsheim

30. Dipl.-Ing. Friedrich Morgenstern, Essen

85 Jahre |1935

10. Dipl.-Ing. Walter Diefenbacher,

Karlsruhe

17. Dipl.-Ing. Helge Dyroff, Alzenau

24. Theodor Himmel, Bad Honnef

87 Jahre |1933

9. Prof. Dr. Hellmut Wagner, Karlsruhe

88 Jahre | 1932

3. Dipl.-Ing. Fritz Kohlhaas, Kahl/Main

91 Jahre | 1929

20. Dr. Devana Lavrencic-Cannata, Rom/I

93 Jahre | 1927

1. Prof. Dr. Werner Oldekop,

Braunschweig

Top

World Energy Outlook 2019

highlights deep disparities in

the global energy system

(iea) Deep disparities define today’s

energy world. The dissonance

between well-supplied oil markets

and growing geopolitical tensions and

uncertainties. The gap between the

ever-higher amounts of greenhouse

gas emissions being produced and the

insufficiency of stated policies to

curb those emissions in line with

international climate targets. The gap

between the promise of energy for

all and the lack of electricity access

for 850 million people around the

world.

The World Energy Outlook 2019,

the International Energy Agency’s

flagship publication, explores these

widening fractures in detail. It explains

the impact of today’s decisions

on tomorrow’s energy systems, and

describes a pathway that enables the

world to meet climate, energy access

and air quality goals while maintaining

a strong focus on the reliability

and affordability of energy for a

growing global population.

As ever, decisions made by

governments remain critical for the

future of the energy system. This is

evident in the divergences between

WEO scenarios that map out different

routes the world could follow over the

coming decades, depending on the

policies, investments, technologies

and other choices that decision

makers pursue today. Together, these

scenarios seek to address a fun damental

issue – how to get from where

we are now to where we want to go.

The path the world is on right now

is shown by the Current Policies

Scenario, which provides a baseline

picture of how global energy systems

would evolve if governments make no

changes to their existing policies. In

this scenario, energy demand rises by

1.3 % a year to 2040, resulting in

strains across all aspects of energy

markets and a continued strong

upward march in energy-related

emissions.

The Stated Policies Scenario,

formerly known as the New Policies

Scenario, incorporates today’s policy

intentions and targets in addition to

existing measures. The aim is to hold

up a mirror to today’s plans and

illustrate their consequences. The

future outlined in this scenario is still

well off track from the aim of a secure

and sustainable energy future. It

describes a world in 2040 where

hundreds of millions of people still go

without access to electricity, where

pollution-related premature deaths

remain around today’s elevated levels,

and where CO 2 emissions would

lock in severe impacts from climate

change.

The Sustainable Development

Scenario indicates what needs to be

done differently to fully achieve

climate and other energy goals

that policy makers around the world

have set themselves. Achieving this

scenario – a path fully aligned with

the Paris Agreement aim of holding

the rise in global temperatures to well

below 2° C and pursuing efforts to

limit it to 1.5° C – requires rapid and

widespread changes across all parts of

the energy system. Sharp emission

cuts are achieved thanks to multiple

fuels and technologies providing

efficient and cost-effective energy

services for all.

“What comes through with crystal

clarity in this year’s World Energy

News


Outlook is there is no single or simple

solution to transforming global energy

systems,” said Dr Fatih Birol, the IEA’s

Executive Director. “Many technologies

and fuels have a part to play

across all sectors of the economy. For

this to happen, we need strong leadership

from policy makers, as governments

hold the clearest responsibility

to act and have the greatest scope to

shape the future.”

In the Stated Policies Scenario,

energy demand increases by 1% per

year to 2040. Low-carbon sources, led

by solar PV, supply more than half of

this growth, and natural gas accounts

for another third. Oil demand flattens

out in the 2030s, and coal use edges

lower. Some parts of the energy sector,

led by electricity, undergo rapid

transformations. Some countries,

notably those with “net zero”

aspirations, go far in reshaping all

aspects of their supply and consumption.

However, the momentum behind

clean energy is insufficient to offset

the effects of an expanding global

economy and growing population.

The rise in emissions slows but does

not peak before 2040.

Shale output from the United

States is set to stay higher for longer

than previously projected, reshaping

global markets, trade flows and

security. In the Stated Policies

Scenario, annual US production

growth slows from the breakneck

pace seen in recent years, but the

United States still accounts for 85 % of

the increase in global oil production to

2030, and for 30 % of the increase in

gas. By 2025, total US shale output

(oil and gas) overtakes total oil and

gas production from Russia.

“The shale revolution highlights

that rapid change in the energy

system is possible when an initial push

to develop new technologies is

complemented by strong market

incentives and large-scale investment,”

said Dr Birol. “The effects

have been striking, with US shale now

acting as a strong counterweight to

efforts to manage oil markets.”

The higher US output pushes down

the share of OPEC members and

Russia in total oil production, which

drops to 47 % in 2030, from 55 % in

the mid-2000s. But whichever pathway

the energy system follows, the

world is set to rely heavily on oil supply

from the Middle East for years to

come.

Alongside the immense task of

putting emissions on a sustainable

trajectory, energy security remains

paramount for governments around

the globe. Traditional risks have not

gone away, and new hazards such as

cybersecurity and extreme weather

require constant vigilance. Meanwhile,

the continued transformation

of the electricity sector requires policy

makers to move fast to keep pace with

technological change and the rising

need for the flexible operation of

power systems.

“The world urgently needs to put a

laser-like focus on bringing down

global emissions. This calls for a grand

coalition encompassing governments,

investors, companies and everyone

else who is committed to tackling

climate change,” said Dr Birol. “Our

Sustainable Development Scenario

is tailor-made to help guide the

members of such a coalition in their

efforts to address the massive climate

challenge that faces us all.”

A sharp pick-up in energy efficiency

improvements is the element that

does the most to bring the world

towards the Sustainable Development

Scenario. Right now, efficiency

improvements are slowing: the 1.2 %

rate in 2018 is around half the average

seen since 2010 and remains far below

the 3 % rate that would be needed.

Electricity is one of the few energy

sources that sees rising consumption

over the next two decades in the

Sustainable Development Scenario.

Electricity’s share of final consumption

overtakes that of oil, today’s

leader, by 2040. Wind and solar PV

provide almost all the increase in

electricity generation.

Putting electricity systems on a

sustainable path will require more

than just adding more renewables.

The world also needs to focus on

the emissions that are “locked in”

to existing systems. Over the past

20 years, Asia has accounted for

90 % of all coal-fired capacity built

worldwide, and these plants potentially

have long operational lifetimes

ahead of them. This year’s WEO considers

three options to bring down

emissions from the existing global

coal fleet: to retrofit plants with carbon

capture, utilisation and storage or

biomass co-firing equipment; to

repurpose them to focus on providing

system adequacy and flexibility; or to

retire them earlier.

| www.iea.org

New report and webinar

on the supply of medical

radioisotopes

(nea) The NEA hosted a webinar on

18 November 2019 to present findings

from a new report on the supply of

medical radioisotopes, jointly produced

with the Organisation for

Economic Co-operation and Development

(OECD) Health Committee.

Technetium-99m (Tc-99m) is the

most commonly used medical

radioisotope, essential for 85 % of the

nuclear medicine diagnostic scans

performed worldwide. There are no

comparable substitutes available for

diagnoses of various cancers and for a

range of diagnostics in children.

Unfortunately, the global supply of

Tc-99m is not technically and economically

robust, and the existing

supply-chain continues to experience

chronic shortages. This new study

analyses the current market structure

and identifies barriers for the implementation

of full cost recovery.

Report and the webinar recording

are available at:

oe.cd/nea-med-rad-webinar-2019.

| www.oecd-nea.org

World Nuclear Performance

Report 2019 Asia Edition

launched

(wna) The World Nuclear Performance

Report 2019 Asia Edition shows

that nuclear energy in Asia is meeting

the growing demand for electricity,

whilst helping to tackle

air pollution and climate change. The

report, published by World Nuclear Association,

was launched today at Singapore

International Energy

Week.Nuclear generation in Asia

continued its rapid growth in 2018, increasing

by 12 %. By replacing

coal-fired generation nuclear energy

avoided the emission of over

500 million tonnes of carbon dioxide

last year.Agneta Rising, Director

General of World Nuclear Association

said, “Nuclear is fast, scalable and a

long-lasting way to add clean

557

NEWS

News


Operating Results July 2019

558

NEWS

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]

Energy generated, gross

[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability

[%] *) Energy utilisation

[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto 1) BWR FI 910 880 744 681 827 4 427 294 266 082 502 100.00 95.66 100.00 94.49 99.61 94.60

OL2 Olkiluoto BWR FI 910 880 744 678 000 4 090 942 255 987 484 100.00 88.26 99.94 87.58 99.05 87.41

KCB Borssele 3) PWR NL 512 484 662 326 759 4 445 853 166 167 541 88.61 80.62 88.66 80.51 85.71 76.95

KKB 1 Beznau 7) PWR CH 380 365 744 265 244 1 583 283 128 917 393 100.00 82.35 100.00 82.10 93.63 81.77

KKB 2 Beznau 7) PWR CH 380 365 744 266 954 1 925 011 136 275 418 100.00 100.00 100.00 100.00 94.22 99.56

KKG Gösgen 3,7) PWR CH 1060 1010 611 634 890 4 627 036 318 502 564 82.06 86.72 82.03 86.04 80.50 85.81

KKM Mühleberg BWR CH 390 373 744 274 120 1 942 150 129 346 465 100.00 100.00 99.89 99.73 94.47 97.89

CNT-I Trillo PWR ES 1066 1003 744 784 619 4 605 674 251 897 342 100.00 86.01 100.00 85.50 98.03 84.41

Dukovany B1 PWR CZ 500 473 744 363 425 2 485 594 114 715 087 100.00 99.86 99.31 99.61 97.70 97.72

Dukovany B2 1,2) PWR CZ 500 473 0 0 1 610 472 109 844 643 0 65.17 0 64.65 0 63.32

Dukovany B3 PWR CZ 500 473 744 357 397 1 966 445 108 464 485 100.00 79.26 100.00 78.81 96.07 77.31

Dukovany B4 PWR CZ 500 473 744 365 370 2 533 257 108 976 525 100.00 100.00 100.00 99.92 98.22 99.60

Temelin B1 PWR CZ 1080 1030 744 798 674 3 926 675 118 287 717 100.00 72.28 99.96 71.97 99.21 71.34

Temelin B2 1,2) PWR CZ 1080 1030 0 0 4 477 198 113 749 715 0 80.87 0 80.81 0 81.34

Doel 1 PWR BE 454 433 744 340 008 1 582 284 137 026 746 100.00 67.09 99.98 66.66 97.84 67.15

Doel 2 PWR BE 454 433 744 334 234 1 915 142 135 717 081 100.00 83.89 99.29 82.19 98.53 82.55

Doel 3 2) PWR BE 1056 1006 181 144 566 4 063 502 259 195 987 24.37 75.48 20.70 74.72 18.01 75.13

Doel 4 PWR BE 1084 1033 744 778 346 5 427 730 265 801 140 100.00 100.00 99.94 97.96 94.76 96.96

Tihange 1 PWR BE 1009 962 744 725 978 5 120 495 303 951 352 100.00 100.00 99.99 99.99 96.68 99.91

Tihange 2 2) PWR BE 1055 1008 693 659 639 659 639 255 311 569 93.13 13.62 87.24 12.76 84.43 12.35

Tihange 3 PWR BE 1089 1038 744 780 007 5 404 413 276 631 686 100.00 99.96 100.00 99.02 96.67 98.05

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]


[MWh]

Time availability

[%]

Energy availability


[%] *)

Month Year Since Month Year Month Year Month Year

commissioning

KBR Brokdorf 1,2) DWR 1480 1410 544 722 179 5 437 455 356 005 265 73.08 81.11 67.73 76.05 65.25 71.95

KKE Emsland DWR 1406 1335 744 1 033 507 5 701 515 352 520 484 100.00 81.41 100.00 81.27 98.81 79.71

KWG Grohnde DWR 1430 1360 744 1 000 783 5 717 420 383 291 634 100.00 82.88 99.55 82.50 93.39 78.10

KRB C Gundremmingen SWR 1344 1288 744 983 969 5 481 876 336 423 631 100.00 81.32 100.00 80.60 97.93 79.74

KKI-2 Isar 1,2) DWR 1485 1410 389 518 222 6 788 871 360 514 684 52.28 93.02 50.22 92.71 46.55 89.53

GKN-II Neckarwestheim 4) DWR 1400 1310 744 910 310 6 858 810 336 685 644 100.00 100.00 100.00 99.88 86.99 96.47

KKP-2 Philippsburg 1,2) DWR 1468 1402 448 580 690 6 339 374 372 500 529 60.22 88.81 59.74 88.52 52.23 83.60

*)

Net-based values

(Czech and Swiss

nuclear power

plants gross-based)

1)

Refueling

2)

Inspection

3)

Repair

4)

Stretch-out-operation

5)

Stretch-in-operation

6)

Hereof traction supply

7)

Incl. steam supply

8)

New nominal

capacity since

January 2016

9)

Data for the Leibstadt

(CH) NPP will

be published in a

further issue of atw

BWR: Boiling

Water Reactor

PWR: Pressurised

Water Reactor

Source: VGB

elec tricity generation.“For the 90

reactors that have started operating

from 2000 to today, the typical construction

time is 5 to 7 years. Of those

90 reactors, 27 % were built in less

than five years – and they will provide

clean and reliable electricity for more

than 60 years or more.”Many reactors

in operation today are planned to

operate for 60-80 years. Reactors are

already demonstrating high performance

irrespective of how long

they have been in operation, with

capacity factors of around 80 %

maintained egardless of age.The

report profiles Tarapur 1, a reactor

located in Palghar, India, which

marked 50 years of operation in

April 2019. Four other reactors will

match this achievement in 2019,

the first year in which reactors have

passed this milestone. Worldwide

nuclear generation in 2018 increased

for the sixth successive year, reaching

2563 TWh. This is more than

10 % of global electricity demand.

Overall, capacity additions for the

period 2016-2020 are expected to

reach the targets of the nuclear

industry’s Harmony programme.

But build rates will have to increase

significantly to achieve the overall goal

of supplying 25 % of global electricity

demand before 2050. Agneta Rising

said, “Nuclear energy is key to Asia

meeting the twin challenges of a growing

demand for electricity, and an

urgent need to switch to less polluting,

low-carbon generation sources. More

and more organizations are recognizing

that nuclear energy is vital to the

goal of a sustainable future for people

and the planet.”

| www.world-nuclear.org

World Science Day:

International Nuclear

Information System

highlights open access

(iaea) Public awareness of developments

in scientific research is crucial

for building a more informed global

society, and the IAEA is providing

access to a trove of research on the

peaceful uses of nuclear science and

technology.

The IAEA’s International Nuclear

Information System (INIS) is one of the

world’s largest repositories of published

research on the peaceful uses of nuclear

science and technology, with more

than 4.2 million bibliographic records

and access to more than 1.6 million

full-text documents. Each year, more

than 1 million unique users perform

around 2 million searches and download

3 million pages.

News


Operating Results August 2019

Plant name Country Nominal

capacity

Type

gross

[MW]

net

[MW]

Operating

time

generator

[h]


[MWh]

Month Year Since

commissioning

Time availability

[%]

Energy availability


[%] *)

Month Year Month Year Month Year

OL1 Olkiluoto 1) BWR FI 910 880 744 680 095 5 107 389 266 762 596 100.00 96.22 99.94 95.18 99.36 95.21

OL2 Olkiluoto BWR FI 910 880 744 678 578 4 769 520 256 666 062 100.00 89.76 100.00 89.16 99.14 88.91

KCB Borssele PWR NL 512 484 693 336 333 4 782 186 166 503 874 91.08 81.95 91.06 81.86 88.20 78.39

KKB 1 Beznau 7) PWR CH 380 365 733 272 559 1 855 842 129 189 952 100.00 84.60 100.00 84.38 96.27 83.62

KKB 2 Beznau 1,2,7) PWR CH 380 365 239 86 668 2 011 679 136 362 086 32.12 91.34 31.92 91.31 30.14 90.70

KKG Gösgen 7) PWR CH 1060 1010 495 511 876 5 138 912 319 014 440 66.47 84.14 66.04 83.49 64.91 83.14

KKM Mühleberg BWR CH 390 373 744 280 420 2 222 570 129 626 885 100.00 100.00 99.93 99.75 96.64 97.73

CNT-I Trillo PWR ES 1066 1003 744 785 304 5 390 978 252 682 646 100.00 87.79 99.87 87.33 98.16 86.16

Dukovany B1 3) PWR CZ 500 473 379 176 941 2 662 535 114 892 028 50.94 93.62 50.34 93.32 47.57 91.32

Dukovany B2 2) PWR CZ 500 473 0 0 1 610 472 109 844 643 0 56.85 0.10 56.39 0 55.24

Dukovany B3 PWR CZ 500 473 744 355 841 2 322 286 108 820 326 100.00 81.91 99.95 81.50 95.66 79.65

Dukovany B4 PWR CZ 500 473 733 355 048 2 888 305 109 331 573 98.52 99.81 97.61 99.62 95.44 99.07

Temelin B1 PWR CZ 1080 1030 744 793 753 4 720 428 119 081 470 100.00 75.82 99.86 75.52 98.60 74.82

Temelin B2 1) PWR CZ 1080 1030 511 537 389 5 014 587 114 287 104 68.68 79.32 66.85 79.03 66.76 79.48

Doel 1 PWR BE 454 433 744 341 195 1 923 479 137 367 941 100.00 71.29 99.97 70.94 98.34 71.16

Doel 2 PWR BE 454 433 744 330 999 2 246 141 136 048 080 100.00 85.95 99.28 84.37 97.52 84.46

Doel 3 PWR BE 1056 1006 744 775 861 4 839 363 259 971 848 100.00 78.61 99.74 77.91 98.19 78.07

Doel 4 PWR BE 1084 1033 744 696 829 6 124 559 266 497 969 100.00 100.00 87.42 96.62 84.39 95.36

Tihange 1 PWR BE 1009 962 744 725 185 5 845 679 304 676 537 100.00 100.00 100.00 99.99 96.57 99.48

Tihange 2 PWR BE 1055 1008 744 756 385 1 416 023 256 067 953 100.00 24.64 99.92 23.88 97.11 23.16

Tihange 3 PWR BE 1089 1038 744 778 719 6 183 131 277 410 404 100.00 99.96 100.00 99.14 96.50 97.86

559

NEWS

Plant name

Type

Nominal

capacity

gross

[MW]

net

[MW]

Operating

time

generator

[h]


[MWh]

Time availability

[%]

Energy availability


[%] *)

Month Year Since Month Year Month Year Month Year

commissioning

KBR Brokdorf DWR 1480 1410 744 994 792 6 432 247 357 000 057 100.00 83.52 94.05 78.35 89.88 74.24

KKE Emsland DWR 1406 1335 744 1 026 542 6 728 057 353 547 026 100.00 83.78 100.00 83.66 98.11 82.06

KWG Grohnde DWR 1430 1360 744 990 908 6 708 328 384 282 542 100.00 85.06 100.00 84.73 92.41 79.93

KRB C Gundremmingen SWR 1344 1288 744 978 185 6 460 061 337 401 816 100.00 83.70 99.40 83.00 97.38 81.99

KKI-2 Isar 4) DWR 1485 1410 744 1 031 644 7 820 515 361 546 328 100.00 93.91 98.93 93.51 92.84 89.95

GKN-II Neckarwestheim 1,2,4) DWR 1400 1310 207 223 200 7 082 010 336 908 844 100.00 100.00 21.19 89.84 21.19 86.86

KKP-2 Philippsburg 1,2,5) DWR 1468 1402 409, 536 941 6 876 315 373 037 470 55.01 84.50 54.82 84.22 48.33 79.10

In line with the theme of this year’s

World Science Day for Peace and

Development – “Open Science, leaving

no one behind” – on 10 November,

INIS plays a vital role in providing

open access to information and

research.

“The IAEA has been an advocate of

open access to knowledge since the

inception of INIS in 1970,” said

Dobrica Savic, Head of the Nuclear

Information Section. All the documents,

presentations and records are

free of charge and accessible to anyone

via the Internet.

Open science is defined by

UNESCO as making scientific research

and data accessible to all. It includes

publishing open scientific research,

promoting open access and making

open source software available.

Libraries and repositories are key for

open science as many play an active

role in the preservation, curation,

publication and dissemination of

digital scientific materials, according

to the Organisation for Economic

Co-operation and Development.

The INIS repository includes

research on a wide range of nuclearrelated

topics, from environmental science

and energy storage to radio logy

and many more. It’s made possible

through an ongoing colla borative

effort between the IAEA and experts

from 132 countries and 24 international

organizations. Member States

submit peer-reviewed research from

their country/organization to INIS,

and approximately 120,000 bibliographic

records and 13,000 full-text

PDF documents are added each year.

To be sure, most publications are

not for the lay reader. A quick glance

turns up titles such as “Method and

apparatus for the nondestructive

assay of bulk nuclear fuel using 1 keV

to Mev range neutrons” or “Regional

comparison of nuclear and fossil electric

power generation costs”. Many

were once only available in hard copy

but are now online after a massive

effort to convert millions of microfiche

pages to fully searchable electronic

files.

Improving INIS inputs

To continually improve the quantity

and quality of submissions to the INIS

repository, the IAEA periodically

offers trainings and e-learning courses

to specialists, librarians and scientific

officers involved in knowledge

management. Last month, experts

gathered in Vienna to discuss different

aspects of INIS operations, such as

News


560

NEWS

selection criteria and descriptive

cataloguing, and improve their skills

for the preparation of high-quality

input and use of the repository. The

participants, more than half of them

women, came from Africa, Asia, Europe

and the Americas.

Full-text documents available in

the INIS repository represent almost

entirely nuclear-related non-conventional

literature. Non-conventional

literature includes any literature

which is not normally available

through commercial distribution

channels and which is generally

difficult to locate. The depth and

breadth of the INIS repository and

the large number of daily users

demonstrate how a well-planned and

implemented international cooperation

project can make a significant

contribution to open science.

| www.iaea.org

Company News

Framatome successfully

implements innovative

maintenance technique

on reactor vessel

component underwater

(framatome) Framatome applied a

cutting-edge maintenance technique

on reactor vessel primary nozzles at

Dominion Energy’s Millstone Power

Station during the plant’s spring 2019

outage. This was the first application

of Framatome’s ultra-high pressure

(UHP) cavitation peening process on

reactor pressure vessel nozzles to

primary pipe welds. Because it is

deployed directly to the inner surface,

it is uniquely suited to remediate the

component regardless of external

space restrictions or dose constraints.

“Framatome’s innovative solutions

are ensuring the efficient and reliable

operation of today’s reactor fleet,” said

Catherine Cornand, Framatome’s

senior executive vice president in

charge of the Installed Base Business

Unit. “This new underwater application

of UHP cavitation peening on a

primary nozzle is another example of

our team’s expertise and dedication to

innovation and continuous improvement

in servicing our customers

worldwide.”

To prepare for the work, Framatome

demonstrated the qualified reactor

vessel primary nozzle cavitation

peening technology on a full-scale

mock-up at the company’s world-class

Technical Training Center in

Lynchburg, Virginia, in early 2019.

UHP cavitation peening is designed

to prevent primary water stress

corrosion cracking. The process uses

ultra-high-pressure water jets to

generate vapor bubbles that collapse

with enough force to create beneficial

compression of the components’

surfaces. This surface compression

improves components’ material

properties and enhances resistance

to corrosion and other types of

degradation, which reduces the

effects of aging.

UHP cavitation peening can extend

the life of nuclear reactor primary

components, including the hot leg

primary nozzles, for up to 40 additional

years. Additionally, the process

reduces outage time and saves money

by eliminating the need to replace

components or address indications

with traditional repair methods. UHP

cavitation peening can be used for

several different applications in most

reactor designs.

“Cavitation peening is an industry

game-changer that was recognized in

2017 as one of the Top Innovative

Practices for work completed on the

Byron and Braidwood reactor vessel

closure heads,” said Craig Ranson,

senior vice president of Framatome’s

North America Installed Base Business

Unit. “We are proud to work with

Dominion to expand our proven

capabilities and engineer a solution

for this unique primary nozzle repair.”

Located in Waterford, Connecticut,

the Millstone Power Station’s two

pressurized water reactors produce

enough electricity to power 2.1 million

homes.

| www.framatome.com

UK Government and industry

champion new compact

nuclear power station

(rolls-royce) UK Research and Innovation

(UKRI) has confirmed it has

provided initial match funding to the

consortium of companies designing a

new type of nuclear power station in

the UK.

| Artist's view of the SMR design of a new UK nuclear power station project.

The initial joint investment of

£18million from UKRI will be matched

by nuclear, civil engineering construction

and manufacturing industry

firms, who have been working on the

preliminary design for four years.

The power station is a compact

design, the components for which are

manufactured in sections in regional

UK factories, before being transported

to existing nuclear sites for rapid

assembly inside a weatherproof

canopy. This cuts costs by avoiding

weather disruptions and secures

gradual efficiency savings by using

streamlined and standardised manufacturing

processes for its components.

By 2050 a full UK programme of up

to 16 of these power stations could

create:

p Up to 40,000 jobs

p £ 52 bn of value to the UK

economy

p £ 250 bn of exports

Paul Stein, Chief Technology Officer

for Rolls-Royce, which leads the

consortium, said: “Tackling climate

change requires collaboration across

industries and governments to find

effective, affordable and sustainable

ways of achieving net zero by 2050.

“The consortium’s work with the

Government shows that action is

being taken to decarbonise our

economy and meet our society’s vital

and growing power needs. This is a

very positive step forward to this next

phase of the programme.”

The partners in the consortium

are Assystem, BAM Nuttall, Laing

O’Rourke, National Nuclear Laboratory

(NNL), Rolls-Royce, Atkins,

Wood, The Welding Institute (TWI)

and Nuclear AMRC.

The target cost for each station is

£1.8 billion by the time five have been

built, with further savings possible.

Each power station will be able to

operate for 60 years and provide

440 MW of electricity, which is

enough to power a city the size of

Leeds.

News


The shared initial investment will

be used to progress the significant

opportunities presented by the programme;

prepare it for the UK’s

regulatory Generic Design Assessment

process; and make final decisions on

which innovations to pursue and realise.

It will also generate the valuable

confidence that the supply chain needs

to begin to prepare for a programme

that could create around £52 billion of

value for the UK economy.

When licensed and supported by

the required enabling legislation and

siting processes, the power station

could provide reliable low carbon

energy from the early 2030s.

The Government’s intent to

support the programme was announced

in July 2019.

| www.rolls-royce.com

Uranium

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

140.00

120.00

100.00

80.00

60.00

40.00

20.00

0.00

1980

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

1985

1990

1995

2000

2005

) 1

2010

2015

2019

Year

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

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

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

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

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

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

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

180.00

22.00

) 1

160.00

140.00

120.00

100.00

80.00

60.00

40.00

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

140.00

120.00

100.00

80.00

60.00

40.00

20.00

0.00

20.00

18.00

16.00

14.00

12.00

10.00

Jan. 2008

8.00

6.00

4.00

Jan. 2009

Jan. 2010

) 1

Jan. 2011

Jan. 2012

Jan. 2013

Jan. 2014

Jan. 2015

Jan. 2016

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020

561

NEWS

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

Westinghouse signs

agreement to acquire Rolls-

Royce Civil Nuclear Systems

and Services Business

(west) Westinghouse Electric Company

signed a definitive agreement to

acquire Rolls-Royce’s Civil Nuclear

Systems and Services business in North

America, expanding Westinghouse’s

global capabilities in digital, engineering

services, plant automation and

monitoring systems, field services and

manufacturing.

“Creating customer value and

supporting our customers’ operations

is a key driver for Westinghouse.

Acquiring Rolls-Royce will strengthen

our ability to serve the nuclear

operating fleet through an expanded

presence in our core business while

adding new digital offerings,” said

Patrick Fragman, Westinghouse president

and chief executive officer. “This

acquisition is an important step in our

growth strategy. We look forward to

welcoming the employees of Rolls-

Royce to Westinghouse.”

The acquisition of Rolls-Royce will:

p Expand Westinghouse’s operating

plant services capabilities

p Enhance the company’s digital

innovation efforts to optimize

customer planning and maintenance,

and provide engineering

solutions to maximize cost

effectiveness and obsolescence risk

p Support both Westinghouse’s and

Rolls-Royce’s global customer base

through an expanded presence

and synergies between both

companies, enhancing customer

offerings and experience in field

services and plant automation

p Further enable Westinghouse’s

growth while supporting

20.00

0.00

Jan. 2008

Jan. 2009

Jan. 2010

Jan. 2011

Jan. 2012


customers in the North American

and European nuclear markets

Rolls-Royce operates 11 sites in

Canada, France, the United Kingdom

and the United States.

| www.westinghousenuclear.com

Market data

(All information is supplied without

guarantee.)

Nuclear Fuel Supply

Market Data

Information in current (nominal)

U.S.-$. No inflation adjustment of

prices on a base year. Separative work

data for the formerly “secondary

market”. Uranium prices [US-$/lb

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

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

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

(Separative work unit)].

Jan. 2013

Jan. 2014

Jan. 2015

Jan. 2016

2017

p Uranium: 19.25–26.50

p Conversion: 4.50–6.75

p Separative work: 39.00–50.00

2018

p Uranium: 21.75–29.20



2019

January 2019

p Uranium: 28.70–29.10

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020

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

2.00

0.00


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

)1

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

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

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

Jan. 2008

Jan. 2009

Jan. 2010

Jan. 2011

Jan. 2012



February 2019

p Uranium: 27.50–29.25



March 2019

p Uranium: 24.85–28.25



April 2019

p Uranium: 25.50–25.88



May 2019

p Uranium: 23.90–25.25



June 2019

p Uranium: 24.30–25.00



July 2019

p Uranium: 24.50–25.60



August 2019

p Uranium: 24.90–25.60



September 2019

p Uranium: 24.80–26.00



| Source: Energy Intelligence

www.energyintel.com

Jan. 2013

Jan. 2014

Jan. 2015

Jan. 2016

Jan. 2017

Jan. 2018

Jan. 2019

Jan. 2020

News


562

NUCLEAR TODAY

John Shepherd is a

freelance journalist

and communications

consultant who has

covered the nuclear

industry for the past

20 years and is

currently editor-in-chief

of UK-based Energy

Storage Publishing.

Sources

World Nuclear

Association

https://cutt.ly/weEfVJC

IEA Sweden report

https://cutt.ly/TeEfZIs

Cryospheric Sciences

Division blog

https://cutt.ly/1eEfLyD

Taking a Leaf out of Greta’s

Climate Change Book

The media loves a ‘warrior underdog’ and so journalists around the world have largely fallen in love with Swedish eco

champion Greta Thunberg. The diminutive teenager has captured the imagination of a new generation of environmental

activists since she burst onto the international scene around a year ago, when she protested on school days outside the

Swedish parliament. She said her school strike was to draw attention to global warming.

Her actions caught on and soon led to other students

making similar protests. The rest, to coin a phrase, is

already modern history. Thunberg has been feted worldwide

and she even addressed the 2018 United Nations

Climate Change Conference.

Whatever one thinks of Thunberg’s approach, it’s hard

to deny that she has reignited interest in the climate change

debate and raised the political temperature on all sides.

So could nuclear’s proponents take a leaf out of the

young environmentalist’s book? Earlier this year, a group

of reactor physicists, operators and politicians were among

those who followed Thunberg’s lead and gathered outside

the Swedish parliament to raise awareness of the benefits

of nuclear energy.

One of the organisers of the ‘Stand up for Nuclear’ event

was Swedish Nuclear Society president Marcus Eriksson.

He told me that, on the same day of the event, the group

published an opinion article in one of Sweden’s major

newspapers, Svenska Dagbladet, which explained the

country’s energy situation in Sweden, making a parallel to

Germany’s energy sector.

The Stand up for Nuclear event is an important development

in Sweden and one that could and should have

wider implications for Europe. Sweden has a chequered

past where nuclear is concerned.

According to the World Nuclear Association, Sweden

generated a combined 156 terawatt hours (TWh) of

electricity in 2016, of which 63 TWh (40%) was from its

eight nuclear power reactors and 62 TWh (40%) from

hydro. Wind provided 15.5 TWh (10%), various fossil fuels

2 TWh, and biofuels and waste 13 TWh (8%). Total

installed generating capacity at the end of 2016, as

recorded by the International Energy Agency‘s (IEA)

‘ Electricity Information 2018’, was 40 gigawatts electric.

In 1980, the Swedish government decided to phase out

nuclear power but this policy was repealed by lawmakers

in June 2010. The country‘s 1997 energy policy had

allowed 10 reactor units to operate longer than envisaged

by the 1980 phase-out policy. However, the policy also led

to the premature closure of a two-unit (1,200 megawattelectric)

plant, although some 1,600 MWe was subsequently

added in uprates to the remaining 10 units.

In 2015, decisions were made to close four older

reactors by 2020. A year later, a tax discriminating against

nuclear power was abolished.

Nuclear’s impact on Sweden’s environmental credentials

is impressive. The country has almost fully decarbonised its

electricity generation, which the IEA has described as “a

feat which is quite unique among the IEA member countries”.

In a review of Swedish energy policy published last

April, the IEA said the country should assess the contribution

of nuclear power over the next 20 years and the impact

of further potential early closures on energy security.

I have no scientific basis for saying this, only a gut

feeling (from reporting on nuclear and the wider energy

industry over the past 20 years), that there is often a silent

majority who would speak up in favour of nuclear –

provided they are given the facts.

It appears such a majority could exist in Sweden, if a

survey conducted last summer by Novus Opinion is

anything to go by. That survey indicated a majority of

Swedes now believe that nuclear power could be a means

to tackle the climate crisis. Which brings me back to the

Stand up for Nuclear event organised in Stockholm.

Marcus Eriksson told journalists covering the event:

“Two years ago none of the political parties wanted to talk

about nuclear power, now everyone is talking about it. It

reflects a stronger opinion that the technology has an

important role to play to combat climate change.”

The Swedish Nuclear Society should be praised for

raising awareness of the benefits of nuclear in this way.

Every day, when we are out and about, there is a manifestation

of some sort or another with campaigners competing

for our attention on issues. So why not something

that speaks up for nuclear? In these gloomy winter nights,

perhaps there is an inbuilt advantage for the nuclear cause

in pointing out just how dark and cold many of our communities

would be without the benefit of atomic power.

As we head into a new year, it’s a good time to start

thinking about the resolutions we will make for the months

ahead. In addition to resolution favourites, such as going

on a diet, or giving up smoking, those involved in the

nuclear sector should pledge to make a special effort to

stand up for nuclear in 2020.

The initial unassuming campaign launched by Greta

Thunberg just a year ago is, I suggest, a template for what

could be a ‘2020 Vision for Nuclear’ campaign for the year

ahead. Perhaps with the advent of social media we’ve

forgotten about the impact of ‘grassroots’ campaigning.

Social media is of course tremendously valuable in

reaching millions. But let’s not forget the impact that can

be had by being out on the streets and chatting to passersby

about the technology that keeps the lights on and can

help clean up our local and global environments.

There’s no time like the present to get out there and

follow Greta’s footprint. In fact, the beloved Santa Clause,

a figure who returns to prominence at this time of the year,

could be recruited to underline the importance of the

nuclear cause.

As the Cryospheric Sciences Division of the European

Geosciences Union has pointed out, Santa might well have

to think about moving from his fabled home of Lapland

because of global warming and polar amplification. In the

absence of snow, Santa does have the possibility of

converting his sleigh to an all-electric model. But even if

there were enough charging points around the globe, they

would need reliable supplies of clean electricity, which

brings us back to nuclear’s importance for our planet’s

energy mix.

In wrapping up this festive article, it only remains to

wish everyone the compliments of the season and, if not a

white Christmas, at least an increasingly green one.

Nuclear Today

Taking a Leaf out of Greta’s Climate Change Book ı John Shepherd

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