atw - International Journal for Nuclear Power | 11/12.2019
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Ever since its first issue in 1956, the atw – International Journal for Nuclear Power has been a publisher of specialist articles, background reports, interviews and news about developments and trends from all important sectors of nuclear energy, nuclear technology and the energy industry. Internationally current and competent, the professional journal atw is a valuable source of information.
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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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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...?
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
Feature
Environment and Safety
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
Environment and Safety
521 Root Causes of the Three Mile Island Accident
Zoltan R. Rosztoczy
Research and Innovation
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
Decommissioning and Waste Management
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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,
Inside Nuclear with NucNet
Key to Unlocking Bulgaria’s Belene Project is Finding Right Financing Package
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
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atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
FEATURE | ENVIRONMENT AND SAFETY 516
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
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
5) Transfer and package the cut pieces to a storage container
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
4) Transfer and package the cut pieces to a storage container
5) Repeat the procedure for the cutting the remaining Core Barrel
FEATURE | ENVIRONMENT AND SAFETY 517
| 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
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
FEATURE | ENVIRONMENT AND SAFETY 518
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
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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
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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
FEATURE | ENVIRONMENT AND SAFETY 519
| Tab. 4.
PSF Classification System in the Reactor Decommissioning Activity.
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FEATURE | ENVIRONMENT AND SAFETY 520
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.
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[6] Center for Chemical Process Safety : Guidelines for preventing human error in process safety
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[7] Byung-Sik Lee: Optimization of reactor pressure vessel internals segmentation in KoreaATW
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[8] Boucau, J., et.al.: Best practices for preparing vessel internals segmentation projects. No.
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[9] Boring, R.L.: Modelling human reliability analysis using MIDAS. In: International Workshop on
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[10] Human performance reference manual, Institute of Nuclear Power Operations, INPO-06-003,
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[11] Chang, Y. H. J., Mosleh, A.: Cognitive modeling and dynamic probabilistic simulation of operating
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[12] Blackman, H. S., Gertman, D. I., Boring, R. L.: Human error quantification using performance
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[15] Mindock, J.: Development and Application of Spaceflight PSFs for HRA. PhD Thesis , University of
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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
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
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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
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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
Environment and Safety
Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy
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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
ENVIRONMENT AND SAFETY 523
Environment and Safety
Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy
atw Vol. 64 (2019) | Issue 11/12 ı November/December
ENVIRONMENT AND SAFETY 524
| 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
Environment and Safety
Root Causes of the Three Mile Island Accident ı Zoltan R. Rosztoczy
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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 thermal- hydraulic 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
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 ı Andrea Bersano and Fulvio Mascari
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
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 ı Andrea Bersano and Fulvio Mascari
atw Vol. 64 (2019) | Issue 11/12 ı November/December
Parameter Average value Range of variation PDF type
RESEARCH AND INNOVATION 528
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
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 ı Andrea Bersano and Fulvio Mascari
atw Vol. 64 (2019) | Issue 11/12 ı November/December
| 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).
RESEARCH AND INNOVATION 529
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).
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RESEARCH AND INNOVATION 530
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.
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| 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.
RESEARCH AND INNOVATION 531
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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
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RESEARCH AND INNOVATION 532
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
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 ı Andrea Bersano and Fulvio Mascari
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
RESEARCH AND INNOVATION 533
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atw Vol. 64 (2019) | Issue 11/12 ı November/December
RESEARCH AND INNOVATION 534
| 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.
Research and Innovation
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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.
RESEARCH AND INNOVATION 535
| 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.
Research and Innovation
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
RESEARCH AND INNOVATION 536
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,
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[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
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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
Beijing 100084, China
Key Laboratory of Advanced Reactor
Engineering and Safety of Ministry of
Education
Tsinghua University
Beijing 100084, China
Research and Innovation
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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 and Waste Management
Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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.
Decommissioning and Waste Management
Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
DECOMMISSIONING AND WASTE MANAGEMENT 539
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Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz
atw Vol. 64 (2019) | Issue 11/12 ı November/December
DECOMMISSIONING AND WASTE MANAGEMENT 540
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
Decommissioning and Waste Management
Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz
atw Vol. 64 (2019) | Issue 11/12 ı November/December
(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
DECOMMISSIONING AND WASTE MANAGEMENT 541
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Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz
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DECOMMISSIONING AND WASTE MANAGEMENT 542
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
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Ulf Kutscher
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Andreas Loeb
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ISSN 1431-5254
Decommissioning and Waste Management
Decommissioning & Dismantling of the Rossendorf Research Reactor RFR ı Reinhard Knappik, Klaus Geyer, Sven Jansen and Cornelia Graetz
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
DECOMMISSIONING AND WASTE MANAGEMENT 543
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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
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DECOMMISSIONING AND WASTE MANAGEMENT 544
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.
Decommissioning and Waste Management
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atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
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DECOMMISSIONING AND WASTE MANAGEMENT 546
| 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.
Decommissioning and Waste Management
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
| 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 %
DECOMMISSIONING AND WASTE MANAGEMENT 547
Decommissioning and Waste Management
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
DECOMMISSIONING AND WASTE MANAGEMENT 548
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
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[14] Nilaya, J. P.; Raote, P.; Kumar, A.; Biswas, D. J.: Laser-assisted
decontamination – A wavelength dependent study. In:
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[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.
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[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
Decommissioning and Waste Management
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
Special Topic | A Journey Through 50 Years AMNT
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.
Special Topic | A Journey Through 50 Years AMNT
Protection of Man and Environment – Nuclear Usage Outside of Energy Sector ı P.P. De Regge, Wien
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
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Special Topic | A Journey Through 50 Years AMNT
Protection of Man and Environment – Nuclear Usage Outside of Energy Sector ı P.P. De Regge, Wien
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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SPECIAL TOPIC | A JOURNEY THROUGH 50 YEARS AMNT
| 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
Special Topic | A Journey Through 50 Years AMNT
Protection of Man and Environment – Nuclear Usage Outside of Energy Sector ı P.P. De Regge, Wien
atw Vol. 64 (2019) | Issue 11/12 ı November/December
| 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
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SPECIAL TOPIC | A JOURNEY THROUGH 50 YEARS AMNT
Special Topic | A Journey Through 50 Years AMNT
Protection of Man and Environment – Nuclear Usage Outside of Energy Sector ı P.P. De Regge, Wien
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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SPECIAL TOPIC | A JOURNEY THROUGH 50 YEARS AMNT
| 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
Special Topic | A Journey Through 50 Years AMNT
Protection of Man and Environment – Nuclear Usage Outside of Energy Sector ı P.P. De Regge, Wien
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
Erwähnung Ihres
Geburtstages in der
atw wünschen, teilen
Sie dies bitte der KTG-
Geschäftsstelle mit.
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atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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]
Energy generated, gross
[MWh]
Time availability
[%]
Energy availability
[%] *) Energy utilisation
[%] *)
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
Operating Results August 2019
Plant name Country Nominal
capacity
Type
gross
[MW]
net
[MW]
Operating
time
generator
[h]
Energy generated, gross
[MWh]
Month Year Since
commissioning
Time availability
[%]
Energy availability
[%] *) Energy utilisation
[%] *)
Month Year Month Year Month Year
OL1 Olkiluoto 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]
Energy generated, gross
[MWh]
Time availability
[%]
Energy availability
[%] *) Energy utilisation
[%] *)
Month Year Since Month Year Month Year Month Year
commissioning
KBR Brokdorf DWR 1480 1410 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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
* Actual nominal USD prices, not real prices referring to a base year. Year
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
p Conversion: 6.00–14.50
p Separative work: 34.00–42.00
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
* Actual nominal USD prices, not real prices referring to a base year. Year
| 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
p Conversion: 13.50–14.50
p Separative work: 41.00–44.00
February 2019
p Uranium: 27.50–29.25
p Conversion: 13.50–14.50
p Separative work: 42.00–45.00
March 2019
p Uranium: 24.85–28.25
p Conversion: 13.50–14.50
p Separative work: 43.00–46.00
April 2019
p Uranium: 25.50–25.88
p Conversion: 15.00–17.00
p Separative work: 44.00–46.00
May 2019
p Uranium: 23.90–25.25
p Conversion: 17.00–18.00
p Separative work: 46.00–48.00
June 2019
p Uranium: 24.30–25.00
p Conversion: 17.00–18.00
p Separative work: 47.00–49.00
July 2019
p Uranium: 24.50–25.60
p Conversion: 18.00–19.00
p Separative work: 47.00–49.00
August 2019
p Uranium: 24.90–25.60
p Conversion: 19.00–20.00
p Separative work: 47.00–49.00
September 2019
p Uranium: 24.80–26.00
p Conversion: 20.00–21.00
p Separative work: 47.00–50.00
| Source: Energy Intelligence
www.energyintel.com
Jan. 2013
Jan. 2014
Jan. 2015
Jan. 2016
Jan. 2017
Jan. 2018
Jan. 2019
Jan. 2020
News
atw Vol. 64 (2019) | Issue 11/12 ı November/December
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
Kommunikation und
Training für Kerntechnik
Strahlenschutz – Aktuell
In Kooperation mit
TÜV SÜD Energietechnik GmbH
Baden-Württemberg
Seminar:
Das neue Strahlenschutzgesetz –
Folgen für Recht und Praxis
Seminarinhalte
1. Teil | Das neue Strahlenschutzgesetz (StrlSchG)
ı Das neue StrlSchG: Historie
ı Inkrafttreten des StrlSchG
ı Überblick über grundlegende Änderungen
ı Die Entwürfe der neuen Strahlenschutzverordnung(en) (soweit zum Seminarzeitpunkt vorliegend)
2. Teil | Auswirkungen auf die betriebliche Praxis
ı Genehmigungen, Zuständigkeiten
ı Begriff der Expositionssituation (geplant, bestehend, Notfall), NORM
ı Aufsichtsprogramm, § 180 StrlSchG, Rechtfertigung
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sowie an Juristen.
Referenten
Dr. Maria Poetsch
Dr. Christian Raetzke
ı Strahlenschutzexpertin bei der TÜV SÜD Energietechnik GmbH
Baden-Württemberg
ı Rechtsanwalt, Leipzig
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