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Nuclear Technology Competence Pool

Issues of Nuclear Safety and Repository Research in Germany

2002 - 2006

Reactor Safety Research

9 July 2003/1

2

Contents Page

1 Introduction 3

2 Evalution commission 3

3 Nuclear Technology Competence Pool 6

3.1 Tasks 6

3.2 Proceedings 6

3.3 Universities 8

3.4 Conclusions 10

4 Technical fields of reactor safety research 2002 – 2006 12

4.1 Component safety and integrity assessment 12

4.2 Thermal hydraulics during transients and loss-of-coolant accidents, 22

reactor physics and control rod behaviour, processes during core

degradation in the reactor pressure vessel

4.3 Core melt in the containment, steam explosion, hydrogen 42

distribution/combustion and countermeasures, fission product

behaviour in the containment

4.4 Development of methods for probabilistic safety analyses, for 55

instrumentation and control, as well as for the assessment

of the human factor

4.5 Know-how transfer regarding safety assessments of Eastern reactors 66

4.6 Innovative concepts 72

5 International co-operation 75

Annex: Abbreviations 81

Literature 88

3

1 Introduction

The report of the working group (Evaluation Commission) convened by the Federal

Ministry of Economics and Technology (BMWi)1“Nuclear Reactor Safety and

Repository Research” of 21 January 2000 has to be detailed in the description of

personnel and task planning until 2006 in order to offer qualitative orientation on the

future development and the maintenance of safety competence at the German

government-funded research-institutes.

2 Evaluation Commission

The intensive funding of research in the field of nuclear reactor safety by the German

Federal Government in the last decades was a decisive contribution to keeping

German reactors among the safest in the world. This result was made possible by

close co-operation between research centres and institutions, expert organisations,

universities and the industry in Germany as well as by close technical co-operation with

institutions abroad.

In view of the political goals laid down in the Atomic Energy Act /AtG 02/ to put an end

to the utilisation of nuclear energy in an orderly manner and substantially reducing

funding of the nuclear reactor safety and repository research in the next years, the

Federal Minister of Economics and Technology (BMWi) considered it to be necessary

to subject the entire field to a review by an Evaluation Commission.

The Evaluation Commission was convened by the BMWi with letter dated

24 September 1999. The tasks were defined to include:

− Establishment of priorities in the field of nuclear safety and repository research in

Germany with special regard to the tight funding;

1 Now the Federal Ministry of Economics and Labour (BMWA)

4

− establishing the medium-term staffing and technical co-operation between the

institutions engaged in nuclear safety and repository research, in particular the

Federal Institute for Geosciences and Natural Resources (BGR), the

Forschungszentrum Jülich GmbH (FZJ), the Forschungszentrum Karlsruhe GmbH

(FZK), the Rossendorf Research Centre (FZR) and the Gesellschaft für Anlagen-

und Reaktorsicherheit (GRS) mbH;

− consideration of the medium-term financial planning;

− special efforts in the framework of research policy so that scientific skills pertaining

to nuclear reactor safety and waste disposal be maintained.

Another task defined for the Evaluation Commission was to recommend ways of closer

co-operation between the research institutes by establishing a so-called Competence

Pool.

Under the chairmanship of the BMWi, the assigned members of the Evaluation

Commission included the leading managers of the research institutes (BGR, FZJ, FZK,

FZR, GRS), the heads of the Project Management Organisations (PT) Reactor Safety

Research (R) and Water Technology and Waste Management (WT+E), and, finally,

representatives from the Federal Ministry of Finance (BMF) (occasionally), the Federal

Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) and the

Federal Ministry for Education and Research (BMBF).

In three meetings (27 October 1999, 26 November 1999 and 21 January 2000), the

Evaluation Commission adopted unanimously the following recommendations. With its

report, the Evaluation Commission has established four basic points:

− A comprehensive survey of the facilities receiving fund from BMWi and BMBF in

the field of nuclear reactor safety and repository research in Germany;

− a suitable basis for the Competence Pool to be founded in the future;

− the identification of high priority tasks and perspectives, as well as

− the identification of additional, important tasks in the fields of nuclear reactor safety

and repository research to be carried out in national and international co-operation.

Against this background, the Evaluation Commission made the following

recommendations:

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1. A co-operation in the field of nuclear reactor safety and repository research, in

terms of both personnel and contents, should be pursued vigorously in Germany

aiming at improved efficiency. The Competence Pool should contribute

substantially by task co-ordination regarding technical matters and contents.

2. The immediate tasks specified in the evaluation report /EVK 00 (Part A, Sec. 5.1)/

shall be handled with high priority.

3. The additional tasks specified in the evaluation report /EVK 00 (Part A, Sec. 5.2)/

shall be carried out within the bounds of the developing safety-related and

temporal necessities as well as of the financial possibilities.

4. Besides the research conducted at BGR, FZJ, FZK, FZR and GRS, importance is

also attached to the research at other facilities in Germany. In this respect,

research activities on nuclear reactor safety and waste disposal at the universities

should be funded sustainably, not at least from the viewpoint of maintaining

scientific competence (promotion of young talents).

5. It should be ensured that Germany continues to be involved efficiently in important

international activities and projects in connection with the maintenance and

continuation of nuclear reactor safety and repository research. This applies to the

co-operation with partners both in the Western world as well as the Central and

Eastern European countries.

6. In view of the limited public funding available, which has already been reduced in

the last years and continues to be so significantly, it should be ensured that no

further reduction occurs with respect to project funding and institutional funding by

the government in order to prevent a further drain in manpower as well as a

decline in competence in this area. The financial means for nuclear reactor safety

and repository research must be sufficient for the Federal Government to fulfil its

legal obligations.

6

3 Nuclear Technology Competence Pool

On 16 March 2000, the Competence Pool had its first meeting. At present, its members

are FZJ, FZK, FZR and GRS. Regular participants are PT R and PT WT+E. Regular

guests are BMBF and BMWA.

3.1 Tasks

In its final report the Evaluation Commission of the BMWi recommends a technical task

co-ordination of German nuclear reactors safety research within the framework of the

Nuclear Technology Competence Pool.

The Competence Pool asked the Project Management Organisation of Reactor Safety

Research to undertake this task in co-operation with the leading research institutions.

The task is to specify the topics according to the technical requirements with respective

quantification of the scientific and technical personnel for the period until 2006. The

technical update of the report of the Evaluation Commission shall serve as guideline for

German research institutions in the next years. The results are presented with this

report.

3.2 Proceedings

For each of the six technical fields of nuclear reactor safety research, i. e.

• component safety and integrity assessment,

• effects of transients and loss-of-coolant accidents on thermal hydraulics, reactor

physics and fuel rod behaviour, processes inside the reactor pressure vessel during

core degradation,

• core melt in the containment, steam explosion, hydrogen distribution and

combustion and countermeasures, fission product behaviour in the containment,

7

• development of methods for probabilistic safety analyses, for instrumentation and

control, diagnostics, and for the assessment of human factors,

• know how transfer regarding safety assessments of Eastern reactors,

• innovative concepts

a working group was established. The members of the working group were chosen so

as to represent the most important institutions with competence in the respective

technical field. Task of the working groups was the co-ordination of future work

regarding technical matters and contents for the years 2002 to 2006, as sell as the

prognosis for the employment of scientific and technical personnel.

The technical structure defined by the Evaluation Commission showed to be too rough

for a substantiated estimation of future research needs. Therefore, more technical

differentiation was required that should be as transparent as possible in its

presentation. For this purpose, the technical fields of reactor safety research were

grouped in topic areas and again in specific topics. In this way, it was possible to

assign the specific topics to the research institutions working in the respective fields.

Table I shows this grouping by the example of component safety and integrity

assessment. In the case of project-funded institutions (in particular universities), these

were assigned to the collective term Project Management Organisation (PT R). Under

“remarks”, it was stated, e. g., where more intensive co-operation and co-ordination

(pool) will be required in future for the purpose of efficiency increase if several

institutions are dealing with the same specific topic. Likewise, already existing

competence gaps or those to be expected und specific topics whose treatment has

been finished were stated there.

8

The institutions dealing with the respective specific topics developed:

• Prognoses on the employment of personnel for the years from 2002 to 2006.

Here, distinction was to be made between technicians (for estimation of

experimental work) and scientists, as well as between prognosticated funding of

person years from institutional funding and project funding.

• Technical substantiation of the respective prognoses according to the state of the

art.

The research institutions involved agreed that the topics of research substantiated here

must not be misunderstood as acquisition efforts of the respective institutions. They

only serve to establish the bases for a future efficient co-operation with optimal use of

the available resources.

The following subject chapters represent the findings obtained on the basis of the

extensive data material. All subject chapters are structured in the same way:

• Statement of the Evaluation Commission on the respective technical field

• Topic areas, specific topics, research institutions (Table No.)

• Personnel prognosis for the respective topic area (Fig. No.)

• Personnel prognosis for the respective technical field (Fig No.)

• Technical description and rationale of work on specific topics.

3.3 Universities

Regarding the maintenance of competence in the field of nuclear safety in Germany,

the future courses offered at the universities are of essential importance. On this issue,

the Nuclear Technology Competence Pool presented the results of a survey, from

which quotations are given in the following /KVK 01/.

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In this context, corresponding questionnaires were sent to 22 universities (universities,

technical universities) and 13 universities of applied sciences (FH) in spring 2000. The

returned questionnaires showed that the current offer of courses in the field of nuclear

technology consists of 17 courses at universities and technical universities and 11 at

universities of applied sciences. The courses offered at the universities vary a lot and

range from reactor physics to general nuclear engineering (which occasionally is dealt

with within the frame of general energy systems), nuclear and radiochemistry and to

radiation protection. The universities were very cautious in their answers to the

medium-term perspective of their courses offered in the nuclear sector due to a number

of imponderabilities. Fig. Ia shows that the number of courses offered at the

universities in the nuclear sector decreases almost linear over the years, i. e. at the

universities from 17 (2000) to 10 (2010) and at the universities of applied sciences from

11 to 3. If this trend continues until the planned shut down of the last nuclear power

plant in Germany in 2021, then the teaching at the universities shrinks to a niche

function, no longer meeting the actual needs /KVK 01/.

Fig. Ib shows the personnel prognosis (reactor safety research) for the universities

until 2006. Fig. Ic illustrates the expected personnel employment according to specific

topics and years.

Accordingly, until 2006 the demand for young scientists at the universities remains

constant for nearly all subject matters of reactor safety research.

When correlating the trend regarding the offer of courses with the number of graduates

in a first approximation, and comparing it with the actual demand, a considerably

increasing deficit would have to be expected.

Topic Areas Specific Topics Research Institutions RemarksDetermination of components loads

GRS / FZR / MPA Competence Pool

• RPV internals FZR• Piping MPA / FZR Competence Pool

Model development for beyond-design-basis accidents

• RPV FZK / (GRS) / (FZR)• Containment (Uni. Karlsr.) Loss of competence

• Pressure boundary GRS / MPA Competence Pool• Containment (leakage behaviour) GRS / MPA Competence Pool

• cyclic loads MPA• irradiation FZK / FZR / MPA Competence Pool• corrosion MPA

…….…….…….Non-destructive material description and material testing

MPA / IZFP Competence Pool

IZFP

IZFP

Further development and verification of the methods for non-destructive testing on austenitic structuresApplication of non-destructive testing for the description of material behaviour

Thermal fluid dynamics/fluid-structure interaction

Quantitative non-destructive testing: Quantification of the interrelation between failure condition, material properties and load conditions

Assignment of Technical Topics to Research Institutions (Example)

Further development of methods for the overall modelling of circuits and components at all safety levels

Material behaviour

Modelling of the overall systems (pressure boundary and main steam-feedwater system)

Structure reliability (probabilistic calculation methods)

Ageing (long-term behaviour) due to

Table I: Assignment of Technical Topics to Research Institutions (Example)

Tabelle-I-e.xls

Personalprognose Hochschulen (Version 1).xls

Fig. Ia: Courses Offered at Universities

Courses Offered at Universities

22

17

13

10

13

11

7

3

0

5

10

15

20

25

1995 2000 2005 2010

Num

ber

Uni/TUFH

Personalprognose Hochschulen (Version 1).xls

Fig. Ib: Personnel Prognosis for Universities (EVK: Evaluation Commission, Mean Value 1996-1998)

Personnel Prognosis for Universities

0

10

20

30

40

50

60

EVK 2002 2003 2004 2005 2006

Man

Yea

rs

Bild Ic.xls

Fig. Ic: Research Activities at Universities

Research Activities at Universities

0

2

4

6

8

10

12

14

Analysys of international concepts

Primary source terms

Core melt in the containment

Level 1 and Level 2

Human behaviour

Reliability of computer codes

Reactor physics for LWRs

Fission product behaviour

Technical systems to support human performance

Thermal hydraulics in the containment

Advanced safety concepts

Instrumentation diagnosis methods

Avoidance of plutonium

Processes during core degradation

Verification of ATHLET

Verification of ATHLET-CD

Man

Yea

rs

20022003200420052006

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3.4 Conclusions

The report of the Evaluation Commission was the essential basis for the future

research activities prognosticated in this report. In accordance with the state of the art

at that time, the Evaluation Commission did not consider some topic areas since they

have been dealt with sufficiently. These issues will not be considered in this report

either so that corresponding statements on the situation with regard to competence

cannot be made. A more detailed technical structure contributes to transparency and

allows orienting quantitative statements on the future work and its efficient co-

ordination, as well as the competence-related focal points regarding the specific topics.

It has to be taken into account that the data material is subject to uncertainties both

under the boundary conditions of national sponsorship and under the aspect of

declining funds for research on reactor safety within the EU’s 6th Framework

Programme. However, this circumstance should not call into question the overall

statement of the findings obtained. In this connection, emphasis is to be laid on the co-

operation of all persons involved in view of the common goal. This is to be judged as

clear indication of a trustful co-operation also in the future.

On the basis of the available data material, trends can be derived for future personnel

employment. On the basis of the prognosticated personnel employment, the following

trends were obtained for the years 2002 to 2006 in comparison to the figures of the

Evaluation Commission of 1998 (Fig. 1):

Decrease: Core-melt phenomena in the containment

(at present decreasing research intensity, e. g. in the fields of

hydrogen behaviour, core-melt spreading)

Increase: Probabilistic safety analyses, instrumentation and control and

diagnosis, human factor

(PSA incl. influence of human behaviour increasingly basis for risk-

based assessments)

The expectations of the research institutions regarding the types of funding

(institutional and project funding) is remarkable. So, there is a considerable reduction of

the available funds with regard to institutional funding and an increase with regard to

project funding (Fig. 3a and Fig. 3b). In view of the medium-term financial planning for

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project-funded reactor safety research, the expectations do not seem to be realistic and

are to be discussed within the Competence Pool.

If comparing the prognosticated number of scientists and technicians for 2006

(about 370 person years) with the actual situation presented in the report of the

Evaluation Commission of 1998 (Figures 2a, 2b) with about 415 person years, a

decline can be noted. Therefore, funding under consideration of the medium-term

budgeting of the Federal Government only seems to be possible, provided the activities

in the centres on reactor safety research within the frame of institutional funding will not

be reduced.

For many specific topics, the personnel employment has already been reduced in the

past years to such an extent that redundancy desirable from a scientific point of view

does no longer exist in Germany, but only within the frame of international co-

operation. This applies, in particular, to experimental work that is still indispensable for

the identification of significant phenomena and for the validation of their analytical

description.

In summary, it can be stated that this report represents a guideline for future co-

operation of the German research institutions in the field of reactor safety research

within the frame of the Nuclear Technology Competence Pool. The success of its

implementation will depend on the readiness of the research institutions for co-

operation and their commitment which is agreed upon by the research institutions.

Auswertung Fachbereiche.xls

Fig. 1: Personnel Prognosis Development (EVK: Evaluation Commission; Mean Value 1996-1998)

Technical Fields of Nuclear Safety Research

0

20

40

60

80

100

120

140EVK 2002 2003 2004 2005 2006

Man

Yea

rs

Thermal hydraulics during transients and loss-of-coolant accidents, reactor physics and control rod behaviour, processes during core degradation in the RPVCore melt in the containment, steam explosion, hydrogen distribution/combustion and countermeasures, fission product behaviour in the containment Component safety and integrity assessmentInnovative concepts Development of methods for probabilistic safety analyses, for instrumentation and control, and for human factor assessmentKnow-how transfer regarding safety assessments of Eastern reactors


Fig 2a: Issue-related Personnel Prognosis (in Person Years) for EVK (Evaluation Commission; Mean Value 1996-1998) and 2006

Evaluation Commission

124,8

120,4

80

61,1

18,2 10,1

Thermal hydraulics during transients and loss-of-coolant accidents, reactor physics and control rod behaviour, processes during core degradation in the RPVCore melt in the containment, steam explosion, hydrogen distribution/combustion and countermeasures, fission product behaviour in the containment Component safety and integrity assessmentInnovative concepts

Development of methods for probabilistic safety analyses, for instrumentation and control, and for human factor assessmentKnow-how transfer regarding safety assessments of Eastern reactors

121,6

68,260,85

73,5

34,39,9

2006



Fig. 2b: Issue-related Personnel Requirements (in Percent) for EVK (Evaluation Commission; Mean Value 1996-1998) and 2006

Evaluation Commisision

31%

29%

19%

15%

4% 2%

Thermal hydraulics during transients and loss-of-coolant accidents, reactor physics and control rod behaviour, processes during core degradation in the RPVCore melt in the containment, steam explosion, hydrogen distribution/combustion and countermeasures, fission product behaviour in the containment Component safety and integrity assessmentInnovative concepts

Development of methods for probabilistic safety analyses, for instrumentation and control, and for human factor assessmentKnow-how transfer regarding safety assessments of Eastern reactors

32%

19%17%

20%

9%3%

2006


Finanzierungsart.xls

Fig. 3a: Personnel Prognosis According to Project Funding and Institutional Funding (EVK: Evaluation Commission; Mean Value 1996-1998)

Type of Funding

0

50

100

150

200

250

300

350

400

450

EVK 2002 2003 2004 2005 2006

Man

Yea

rs

Project fundingInstitutional funding

Finanzierungsart.xls

Fig 3b: Personnel Prognosis According to Technical Fields and Type of Funding (EVK: Evaluation Commission; Mean Value 1996-1998)

Personnel Prognosis According to Technical Fields and Type of Funding

0

20

40

60

80

100

120

140

2002 2003 2004 2005 2006

Man

Yea

rsThermal hydraulics, core degradation in RPV (project funding)Thermal hydraulics, core degradation in RPV (institutional funding)Probabilistic safety analyses, instrumentation and control, human factor (project funding)Probabilistic safety analyses, instrumentation and control, human factor (institutional funding)Component safety and integrity assessment (project funding)Component safety and integrity assessment (institutional funding)Safety assessment of Eastern reactors (project funding)Safety assessment of Eastern reactors (institutional funding)Core melt in the containment (project funding)Core melt in the containment (institutional funding)Innovative concepts (project funding)Innovative concepts (institutional funding)

EVK Mean Value 96-98

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4 Technical fields of reactor safety research 2002 – 2006

4.1 Component safety and integrity assessment

Statement of the Evaluation Commission:

“Questions concerning ageing of components and materials and the consequential

reduction of safety margins for components and functions are increasingly gaining

importance with the increasing operating time of the facilities“.

Table 1 : Topic areas, specific topics, research institutions

Fig. 4 : Personnel prognosis for topic areas

Fig. 5 : Personnel prognosis for the technical field

Chapter Topic Areas Specific Topics Research Institutions Remarks

4.1.1 Determination of components loadsGRS / FZR / MPA Competence pool

• RPV internals FZR• Piping MPA / FZR Competence pool


• RPV FZK / (GRS) / (FZR)• Containment (Karlsruhe University) Loss of competence

• Pressure boundary GRS / MPA Competence pool• Containment (leakage behaviour) GRS / MPA Competence pool

4.1.2 Material behaviour• cyclic loads MPA• irradiation FZK / FZR / MPA Competence pool• corrosion MPA

Material laws for• high temperatures MPA• high speed loading MPA

MPA

4.1.3 Fracture mechanics FZR / MPA Competence pool

MPA

FZR / MPA Competence pool

MPA

Technical Field - Component Safety and Integrity Assessment

Further development of methods for the overall modelling of circuits and components at all safety levels


Structure reliability (probabilistic calculation methods)

Thermal fluid dynamics/fluid-structure interaction

Further development of methods for the description of instable crack propagation

Application of the crack arrest concept in nil ductility transition

Technological influences on the material behaviour (internal stresses)

Ageing (long-term behaviour) due to

Reliability of the crack initiation value as material property

Verification of the applicability of crack toughness curves, determined by laboratory tests, to components

Table 1: Main Research Topics of the Research Institutions in the Field of Components Safety and Integrity Assessment ( ):expiring

4.1 Themen-KS- 02.12-e.xls

1 v 2


Technical Field - Component Safety and Integrity Assessment

4.1.4 Simulation of processes at the nano-, micro- and mesostructure level for the description of material properties

FZR / MPA Competence pool

4.1.5 Non-destructive material description and material testing MPA / IZFP Competence pool

IZFP

IZFP

Further development and verification of the methods for non-destructive testing on austenitic structuresApplication of non-destructive testing for the description of material behaviour

Quantitative non-destructive testing: Quantification of the interrelation between failure condition, material properties and load conditions

Table 1: Main Research Topics of the Research Institutions in the Field of Components Safety and Integrity Assessment ( ):expiring

4.1 Themen-KS- 02.12-e.xls

2 v 2

Auswertung Einzelthemen Drilldown.xls

Fig. 4: Personnel Prognosis for Component Safety for Topic Areas in the Field of Component Safety and Integrity Assessment

Component Safety and Integrity Assessment

0

5

10

15

20

25

30

35

40

45

2002 2003 2004 2005 2006

Man

Yea

rs

Fracture mechanicsDetermination of component loadsSimulation of processes at the nano-, micro- and mesostructure level for the description of material propertiesFurther development of methods for the overall modelling of circuits and components at all safety levels Material behaviourNon-destructive material description and testing

Auswertung Cluster.xls

Fig. 5: Personnel Prognosis for the Technical Field "Component Safety and Integrity Assessment" (EVK: Evaluation Commission; Mean Value 1996-1998)

Focal Points of Competence of the Technical Field "Component Safety and Integrity Assessment"

0

10

20

30

40

50

60

70

80

90

EVK 2002 2003 2004 2005 2006

Man

Yea

rs

FZK, Karlsruhe University FZR, TU Dresden, HTWS Zittau GRS, TU München FZJ, TH Aachen mainly MPA und IZFP

13

Most of the work planned in the technical field of component behaviour and integrity

assessment is performed in co-operation.

4.1.1 Determination of component loads, further development of methods for the overall modelling of circuits and components at all safety levels


In the past years, the methods for the simulation of the mechanical system behaviour

(i. e. consideration of the interaction between the components) has been further

developed and both generic quasi-statical and dynamic analyses have been conducted

on various issues.

The aim of current work on sensitivity of the overall system in case of earthquake loads

is to develop a model for the quantification of the loss of safety functions by

earthquake-induced failure of components affecting more than one system and

redundancy, under consideration of load-related dependencies. The work shall be

continued particularly regarding the implementation of the models in analyses tools.

Thermal fluid dynamics/fluid-structure interaction: RPV internals and piping

The operational monitoring systems installed in the past years showed different new

types of thermal load conditions. These are, above all, thermal stratification processes

that not only lead to high local but also to global loads. Conservative simplified

calculation methods showed that it is not possible to develop general calculation

methods for this load condition. By modelling the entire cooling circuit under

consideration of the structure-fluid interaction, it is not only possible to identify local but

also global load conditions correctly. Due to the complexity of such calculation models

there is still a great demand for research.

14

The consideration of fluid-structure interactions is also of importance to FEM modelling

(Finite Element Method) pressure surges in pipes (water hammer). On this issue,

investigations are currently being performed that are funded by the EU.

For further qualification of the FEM models on load analysis for RPV internals,

mechanical and thermal hydraulic computer codes should directly be coupled, the

radiation exposure in the near-core area identified and considered, the modelling of

non-linear material behaviour enhanced and the applicability of the FEM programs

extended to cases with large deformations.


The past showed that the simulation of incidents or the recalculation of accidents

occurred provides essential findings on actual safety margins. Therefore, the simulation

of material and component behaviour under beyond-design-basis-accident conditions

provides a useful data base for the assessment of accidents that might occur and

which serves to assess the hazard potential in a more reliable manner.

The aim of the current work is to further develop the analysis methods on the

simulation of the structure-mechanical behaviour of components that are exposed to

both high thermal and mechanical loads for a longer period of time, thus creeping is the

dominant effect. By means of analyses on the tests conducted at Sandia National

Laboratories (USA), the precision of the results obtained by the available analysis

methods with regard to a realistic simulation of the creep behaviour of components at

high temperatures is shown. Moreover, a criterion for the determination of component

failure due to creep and plastification at high temperatures is derived that can be

applied to safety-related issues.

Structure reliability (probabilistic calculation methods): pressure boundary, containment

At present, the application of probabilistic methods within the framework of risk-based

maintenance is an international trend that will become widely accepted in the next

years. In order to avoid decoupling of the German safety research in this sector,

15

studies on the assessment of the capability of probabilistic calculation methods are

indispensable under safety aspects.

The aim of current work is to develop an analysis tool for the quantitative assessment

of the structure reliability of defective welds in pipes for German nuclear power plants.

The work is based on the already existing PRAISE (Piping Reliability Analysis

Including Seismic Events) computer code, that is oriented towards the US-American

rules and regulations, and material data and operating experience reflecting the safety

level existing at German nuclear power plants. After finalisation of this project, the

results obtained shall be used for the assessment of structure reliability of further

components, e. g. the reactor pressure vessel.

With regard to fracture-mechanical analysis methods, dissimilar weld joints, i. e. joints

between ferritic and austenitic materials, are very special. Such welds exist, e. g. in the

safety-relevant emergency and residual-heat-removal systems both in pressurised and

in boiling water reactors. The aim of future work is the qualification and further

development of the analysis methods for the assessment of the integrity of dissimilar

weld joints in the coolant circuit, because the experts’ opinions on the preciseness of

the results obtained by fracture-mechanical analysis methods vary. The verification of

the fracture-mechanical methods requires, among others, a participation in the

international reference analyses on a large-scale test of Electricité de France (EdF).

For the treatment of uncertainties, probabilistic methods shall be applied.

In the past years, numerous research activities were conducted on RPV integrity as

well as large-scale tests on the behaviour of cracks in thick-walled structure for

different operating conditions. The aim of current work in the EU-funded QUAMET

(quantification methodology) project is the development of computer-aided user

interfaces for the use of the analyses methods to assess the integrity of large

components and for the qualification of users.

For the determination of leak rates of concrete containments, the analysis methods on

the assessment of the barrier effectiveness have to be enhanced. For this purpose,

simulations of large-scale tests shall be conducted. Further, methods for the

determination of the crack opening behaviour required for leak rate determination shall

be developed. Finally, probabilistic methods shall be applied for the treatment of

uncertainties.

16

4.1.2 Material behaviour

Ageing due to cyclic loads, irradiation and corrosion

It is necessary to verify the fatigue curves defined in the rules and regulations for

corrosive ambient conditions for various load frequencies, as well as for exclusively

thermal loads. This requires further scientific corroboration of the curves applied in the

rules and regulations.

This shall serve as a basis for a realistic determination of the operating lifetime.

With increasing age of plants in operation, potential material damages have to be

identified at an early stage in order to avoid an increase of failures. In particular, stress-

corrosion cracking still contains a high degree of uncertainty regarding the

determination of essential influencing factors. However, they have not to be known

necessarily for the definition of suitable corrective measures for damage prevention.

With advanced damage-mechanical material models, based on material physics and

verified by selected tests, it is possible to predict the influence of stress-corrosion

cracking on the operating lifetime more precisely, in particular under realistic load

conditions.

Likewise, research is needed regarding the aspect of material fatigue. Here, it is of

particular disadvantage that the highly frequent load cycles at the inner side of

components cannot be detected at the outer side and the consequential loads cannot

be described as adequately verified yet. On this issue, there is additional research

demand.

Mechanical tests on ageing due to irradiation require the use of hot cells. Apart from

the industry, there are only two research institutions left that use these cells.

The focal points of the studies on RPV embrittlement are covered by activities of the

EU (Technical Assistance to the Commonwealth of Independent States “TACIS“ and

Ageing Materials Evaluation and Studies “AMES“), the IAEA and the WTZ projects

(scientific-technical co-operation). In this field, German institutions are also involved in

the IAEA working group and the AMES network.

17

Further, studies are conducted to determine the influence of hydrogen on the behaviour

of irradiated RPV steel. With regard to aged RPVs, this additional aspect is considered

in the corresponding integrity assessment. Accompanying simulation calculations on

hydrogen diffusion, their coupling with micromechanical models and consequential

mechanical properties will contribute to understand the correlated phenomena.

Problems still to be solved within the frame of the focal point of ageing/irradiation result

from the behaviour of the austenitic materials of the reactor internals and the RPV

cladding under the influence of gamma radiation. In this respect, the aim is not only to

consider the phenomenological characterisation of ageing processes but also the

clarification of the basic mechanisms. As of 2003, it will be possible to use material

from the decommissioned reactors of the Greifswald Nuclear Power Plant for post-

irradiation examinations that are unique by now. The results of corresponding research

would provide comprehensive information on material behaviour over a longer period of

operation.

For the determination of the internal stresses in the components, e. g. caused by

welding, knowledge on material models and solidification laws is necessary. In this

respect, the material models used for the calculations must also be able to correctly

describe the influence of a cyclic thermal and mechanical load on the deformations and

stresses.

Material laws for high temperatures and high speed loading

For the protection against beyond-design-basis loads and the determination of resulting

impacts, the special material behaviour at temperature has to be considered also in the

time-dependent deformation range (creep).

There will be a demand for research in the field of creep-fatigue interaction with regard

to the experimental determination of the material properties and their application in

material laws and integral material models. These shall be used to model deformation

behaviour and component failure realistically. In addition, special load situations,

overlapping with those of specified normal operation, have to be considered and

analysed. If a load situation exceeds the limit values and is characterised by locally

high temperatures or fast load transients, the existing material models have to be

adapted to the respective deformation and damage mechanisms. It is known from

experiences in the field of non-nuclear technology that time-dependent deformation

18

and failure processes cannot be dealt with analytically, and that they always have to be

coupled with the material condition. Under this aspect, the impact of the operation-

induced change of the material condition has to be included in the analysis of the high-

temperature behaviour. On this issue, there is no knowledge available.

Technological influences on the material behaviour

Technological influences on the material behaviour result from production and

processing of the materials. These have an influence, among others, on the corrosion

and irradiation behaviour and the specific mechanical-physical properties that are

important to the integrity behaviour of the components. The determination of the

technological influences on the material behaviour and material properties are the

object of current research activities. The findings obtained allow an early assessment

of the damage in the material, such as crack formation in endangered zones. The

extensively existing competence is to be maintained.

4.1.3 Fracture mechanics

Reliability of the crack initiation value as material property

The introduction of the J-integral principally allows the derivation of physical crack

initiation values for the entire toughness range and its determination by experiments.

These values can be determined by small specimens that are also suitable for

irradiation, and they also establish the basis for load analyses under PTS conditions

under which a ductile crack initiation cannot be excluded. The validation of the

applicability of the fracture-mechanical properties, determined by means of small

specimens using experimental data of large specimens to real components still has to

be performed.

Verification of the applicability of crack toughness curves, determined by laboratory tests, to components

Results of experiments on large specimens showed that, in dependence of the

multiaxiality of the stress state, the failure behaviour of thick-walled components, as

19

they exist in nuclear installations, may be brittle. The quantitative influence of the

multiaxiality, in particular on the fracture mode, has not been verified adequately yet.

This requires numerical analyses of component-type specimens in order to determine

the development of the degree of multiaxiality with increasing load and to identify the

beginning of the failure.

Further development of methods for the description of instable crack propagation

New fracture-mechanical assessment methods, such as the master-curve concept,

have been developed in the last years and partly have been adopted in the rules and

regulations. However, a general application and their applicability to real components,

in particular regarding the proof of RPV integrity, is still obstructed by numerous open

questions. Not least, the proof on the adequacy of the concept for the description of the

toughness behaviour of irradiated RPV steel at German plants or under dynamic loads

and the linkage of the concept with the procedures laid down in rules and regulations

still have to be dealt with. Finally, the integration of the concept into the embrittlement

monitoring programs has to be prepared. In this respect, numerous international

activities (e. g. within the framework of the IAEA) are in progress or in preparation and

should be accompanied in a competent manner also from a German point of view.

Application of the crack arrest concept in nil ductility transition

With regard to the safety analysis, it still has to be demonstrated that an initiated crack

is arrested at the latest at 75 % of the wall thickness. Since, in general, this area of the

RPV wall is in the upper shelf of toughness even in case of an incident or in end-of-life

(EOL) state, corresponding procedures are to be developed on the experimental

determination of crack-arrest properties that consider the theoretical, material-

mechanical or damage-mechanical processes.

In addition, the determined and verified crack-arrest/toughness limit curve has to be

coupled to the master curve to enable a consistent proceeding on a more realistic basis

regarding material mechanics.

20

This requires the definition of criteria for the minimum size of specimens to obtain

transferable material properties.

4.1.4 Simulation of processes at the nano-, micro- and mesostructure level for the description of material properties

The occurrence of faults during production processes or repair but also the material

properties can be described by the determination of processes at the nano- and

microstructure level. Good progress was made on the assessment of the impacts of

irradiation on the RPV material. The extension of the theories, e. g. for the description

of diffusion processes during welding of a dissimilar weld joint not only increases the

physical understanding of the technological processes, but also enables the

assessment of safety and reliability of the components.

The material simulation for the description of ageing phenomena is of great

importance. With increasing operating time, operation-induced loads may lead to

microstructural changes. The extent of these changes and their impact on the material

properties have to be investigated in advance with regard to future changes. The aim of

future work is to also make use of the nano simulation for understanding the

characteristics of complex material systems. In the medium term, such work shall serve

to establish the basis for an objective and qualitative assessment of potential changes

of safety margins due to operational loads.

It will be absolutely necessary to link these model developments with experimental

results. Above all with regard to the consideration of radiation effects, there are almost

no results available. So, e. g. various damage models have been developed, but hardly

adapted to the description of the damage behaviour of irradiated materials or verified

for these.

21

4.1.5 Non-destructive material description and testing

Non-destructive tests play an important role in the quality assurance for a component

during production and operation. Various damage events showed that there are still

open questions not only regarding the searching technique itself but also regarding the

determination of the defect size.

Quantitative non-destructive testing

The more the real material condition approaches its designed operating lifetime due to

loads, the more important is the non-destructive testing, corroborated in its quantitative

statements. Therefore, further work should focus on improvements regarding the proof

sensitivity and the possibility of classification and sizing of verifiable non-conformances.

Further development and verification of the methods for non-destructive tests on austenitic structures

The reliability of the non-destructive testing still has deficiencies regarding the test on

austenitic cladding, austenitic welds and dissimilar weld joints. For this reason, the

clarification of these deficiencies is a particular objective of the non-destructive test.

This shall be realised by simulation of the sound propagation in these anisotropic

materials. An enhanced understanding of the sound propagation can lead to an

optimisation of the methods on interpretation of the measurement results.

Application of non-destructive testing for the description of the material behaviour

Non-destructive testing provably led to preliminary success with the prediction of

mechanical-technological properties by measurement of test quantities that are

sensitive to microstructure parameters and their changes and that have an influence on

load and internal stresses. The activities on the characterisation of material behaviour

with regard to thermal ageing (problems associated with Cu-precipitation steels),

neutron embrittlement and fatigue should be continued.

22

4.2 Thermal hydraulics during transients and loss-of-coolant accidents, reactor physics and control rod behaviour, processes during core degradation in the reactor pressure vessel


“The realistic description of processes in the reactor core and in the cooling circuits

during incidents and accidents is of essential importance to the safety assessment and

the further improvement of precautionary measures. New demands result from the

progressing development of systems engineering and the operating procedures, as

well as from the increasing scope of simulations.”




Technical Fields - Thermal Hydraulics in Case of Transients and LOCAs, Reactor Physics and Fuel Rod Behaviour, Processes in RPV During Core Degradation

Chapter Topic Areas Specific Topics Research Institutions Remarks Experimental and analytical work

on thermal hydraulics in case of transients and LOCAs

4.2.1 Dynamic flow regime and condensation processes

• Experiments and bases of development PT R (FZR, IKE)

• Model development and implementation in system codes GRS

Multi-dimensional description of two-phase flows

• Experiments and bases of development PT R (TUM, HTWS, FZR) Competence Pool

• Model development and numerics GRS, PT R

• Experiments and bases of development PT R (FZR)

• Model and code development GRS

External validation of ATHLET PT R (HTWS)

Heat transfer at high temperatures and under the influence of inert gas GRS, PT R (HTWS)

GRS, PT R (HTWS)

4.2.2

4.2.3 FZK

FZR

GRS

Further development and experimental verification of ATHLET

One- and multi-dimensional description of single-phase flows (boron dilution and mixture processes)

Effectiveness of accident management measures (i.a. evaluation of PKL / UPTF-TRAM)

Development of methods for safety assessments

Themal hydraulics of advance safety conceptsContributions to international verification of system codes (RELAP, CATHARE)

Thermal- and fluid dynamics during supercritical pressure

Effectiveness of passive systems

Table 2: Main Research Topics of the Research Institutions in the Field of Thermal Hydraulics 4.2 Themen-TH- 02.12-e.xls 1 v 4



4.2.4 Reactor physics for LWR

• generation of cross-sections e.g. for plutonium avoidance/destruction FZK

• generation of cross-sections for BWR and PWR PT R (IKE)

GRS, FZK, FZJ, PT R ( FZR, TUM ) Competence Pool

• increased burn-up GRS

• increased enrichment GRS

• use of MOX fuel GRS, PT R (FZR)

• use of advanced nuclear fuel PT R (TUM), FZJ, FZK Competence Pool

PT R (RUB)

PT R (FZR)

4.2.5 Interaction between neutron kinetics and thermal-fluid dynamics

• ATHLET / QUABOX-CUBBOX GRS

• ATHLET / DYN3D PT R (FZR)

• RELAP / PANBOX FZK

• multi-dimensional fluid dynamics codes/physics PT R (FZR, TUM)

• maintenance and updating

• early phase GRS Work almost completed

• late phase FZK

Development of methods for the generation of nuclear cross-sections

Development of time-dependent neutron transport calculation codes for transient analyses of reactivity-initiated accidents

Further development of Monte Carlo Methods (i. a. reactor dosimetry)

Physics of the reactor core in case of

Development of methods for on-line core monitoring

Further development and verification of the coupling of neutron kinetics and thermal-hydraulic computer codes

Study on recriticality in case of core degradation




4.2.6 Fuel rod (-element) behaviour

• stationary operation (Transuranus) ITU Inst. for Transur. Elements, Karlsruhe

• operationally realistic transients GRS, FZR Competence Pool

• LOCA - analytics GRS - experiments -

• Reactivity-initiated accident GRS

CABRI experiments on LWR fuel rods under accident conditions

Analytical accompaniment and evaluation GRS

4.2.7 Processes during core degradation

• experiments on the early phase of core degradation during reflooding (QUENCH experiments) FZK

• core-melt/coolant interaction (steam explosion) FZK

• debris cooling and crust formation PT R (IKE), FZK Competence Pool

4.2.8 Further development and verification of ATHLET-CD • development of individual models PT R (IKE)

• development and verification of the overall code GRS

• external validation PT R

4.2.9 Contributions to international verification of system codes (e.g. MELCOR, SCDAP / RELAP, ICARE / CATHARE)

FZK Loss of competence

4.2.10 Integrity of the RPV bottom

GRS

Coolability (internal and external cooling)

• model development and evaluation for FOREVER II (EU) FZR

• analytical accompaniment and evaluation of Lower Head Failure (OLHF) GRS

• development of an ATHLET-CD module GRS

Further development of analysis tools for the assessment of fuel rod behaviour

Coupled problems of thermal hydraulics, neutronics and core material behaviour

Interaction of melt or debris bed with RPV bottom

Analytical accomp. and evaluation of RASPLAV/MASCA (OECD/NEA)




4.2.11 Thermal hydraulics

• ATHLET GRS

• RELAP FZK Loss of competence

Reactor physics GRS, FZK Competence Pool

Interaction between neutron kinetics and thermal-fluid dynamics GRS

Fuel rod behaviour GRS

Development of methods for the assessment of the reliability of computer codes also with regard to the applicability to real plants



Fig. 6: Personnel Prognosis for Topic Areas in the Field of Thermal Hydraulics

Thermal Hydraulics During Transients and Loss-of-coolant Accidents, Reactor Physics and Fuel Rod Behaviour, Processes During Core Degradation in the Reactor Pressure Vessel

0

5

10

15

20

25

30

35

40

45

2002 2003 2004 2005 2006

Man

Yea

rs

Contributions to international verification of system codes (e.g. MELCOR, SCDAP/RELAP, ICARE/CATHARE)Fuel rod behaviourIntegrity of the RPV bottomDevelopment of methods for the assessment of the reliability of computer codes also with regard to the applicability to real plantsReactor physics for LWRThermal hydraulics of advanced safety conceptsProcesses during core degradationInteraction between neutron kinetics and thermal-fluid dynamics Further development and experimental verification of ATHLETFurther development and experimental verification of ATHLET-CD


Fig. 7: Personnel Prognosis for the Technical Field of Thermal Hydraulics (EVK: Evaluation Commission; Mean Value 1996-1998)

Thermal Hydraulics During Transients and Loss-of-coolant Accidents, Reactor Physics and Fuel Rod Behaviour, Processes During Core Degradation in the Reactor Pressure Vessel

0

10

20

30

40

50

60

70

80

90

EVK 2002 2003 2004 2005 2006

Man

Yea

rs

FZK, Karlsruhe University FZR, TU Dresden, HTWS Zittau GRS, TU MünchenFZJ, TH Aachen e.g. Industry, TÜV and other Universities

23

4.2.1 Further development and experimental verification of ATHLET

At present and for the foreseeable future, the thermal-hydraulic code ATHLET

represents an essential basis for safety assessments of reactors operated in Germany.

This code is also increasingly used coupled with codes for core melt and fission

product behaviour (ATHLET-CD), neutron kinetics (QUABOX/CUBBOX, DYN3D) and

containment behaviour (COCOSYS). It is also the most important process model in the

ATLAS analysis simulator. In addition, this code serves as basis for essential activities

within the framework of scientific-technical co-operation with eastern partners.

For this reason it is necessary, besides further development of thermal-hydraulic

models and supplementation by a 3D module, to put new findings consequently into

practice, especially those from validation and feedback from user experience, and to

ensure permanent quality assurance.

Dynamic flow regime and condensation processes

At present, the dynamic transition between the respective flow regimes during an

accident is not considered adequately in the computer codes. Analytical and

experimental studies have to be continued. This includes experiments on an air/water

test loop to investigate the flow regime along the vertical pipe axis. These experiments

are precursors to TOPFLOW studies with greater pipe diameters and with steam/water

mixtures planned for 2003 onwards. The measurements particularly concentrate on the

void distribution measured with wire mesh sensors providing high temporal and spatial

resolution. These experiments have to be accompanied by theoretical developments

on the description of the gas fraction distribution across the pipe cross section in

dependence of the size of the observed single bubbles. For the calculation of the

development of the bubble size distribution along the pipe axis, coalescences and

fractioning rates are derived from the experiments. With regard to condensation

phenomena, the focal point so far have been condensation processes in horizontal

pipes.

The modelling of the dynamic development of flow configurations is still in its initial

phase. This applies both to computer codes such as ATHLET or CATHARE and

especially to Computational Fluid Dynamics (CFD) codes, which, so far, only include

very limited two-phase-flow models. In ATHLET, a first approach has already been

24

implemented that is based on a dynamic model for the interfacial area. The new

approaches are expected to enable scalable modelling.

Extensions for dispersing flows, for the area of water entrainment and for horizontal

flows with transient stratification processes still have to be realised. In this respect, it is

to be expected that the well-instrumented experiments conducted within the framework

of the German CFD research pool will deliver important information.

Multi-dimensional description of two-phase flows

Safety assessments require the application of computer codes with increased

prediction accuracy, particularly when modelling multi-dimensional flow processes. A

more detailed analytical simulation is necessary in particular for the quantification of

safety margins of nuclear power plants in operation, e. g. in case of capacity increases

and new core refuelling strategies. One of the main causes for the uncertainties that

still exist is the insufficient capability of the one-dimensional computer codes for the

modelling of multi-dimensional two-phase flow processes. Today, high-resolution CFD

codes are already successfully applied for the simulation, e. g., of three-dimensional

complex mixture processes. However, an advanced state of development has only

been reached with regard to single-phase multi-dimensional flows. Thus, one of the

main objectives is the development of two-phase multi-dimensional models for the CFD

code CFX-5 and the later coupling of these models with the one-dimensional ATHLET

code. This task can only be solved in co-operation within the framework of concerted

experimental and analytical projects. The work to be performed, scheduled over

several years by several research institutions has been started. On the one hand, it is

planned to conduct the experiments at the TOPFLOW test facility on transient two-

phase flows in vertical pipes with and without cross-sectional obstructions. On the other

hand, it is planned to conduct experiments on stratified steam/water flows (sudden

change from subcritical flow to supercritical flow and countercurrent flow limitations), as

they may occur in the hot leg of a PWR in case of loss-of-coolant accidents.

Two-phase flow models for CFD codes have to be developed within the framework of

long-term projects. The scope of application of a CFD code will first be limited to a few

flow configurations and gradually extended to the entire two-phase flow area. However,

even before such a CFD module can be coupled with a computer code, multi-

25

dimensional simulations in the RPV are necessary with ATHLET that comprise the

entire two-phase flow area, although with coarser spatial resolution compared to

application of CFD codes. Here, the FLUBOX module, already coupled with ATHLET,

has to be further developed and validated on the basis of the split-coefficient-matrix

method. The development of this method is sponsored by EU within the framework of

the ASTAR project.

One- and multi-dimensional description of single-phase flows

The experimental work concentrates on the analysis of coolant mixture in PWR in

connection with cold water and boron dilution transients. The experiments are

conducted at the ROCOM facility (1:5 model of a Konvoi reactor). The experiments are

recalculated with CFD codes and simplified mixture models. They are also relevant to

PTS scenarios. For this reason, it is planned to extend the experiments into this

direction at a later time.

System effects related to incidents with boron dilution are still investigated at the PKL

test facility. The results of the experiments are available for ATHLET validation and, if

required, model improvement. The tests on boron dilution conducted at the ROCOM

facility are to be used for validation and improvement of CFX modelling.

External validation of ATHLET

Due to the advanced validation state of ATHLET, supportive external validation is only

necessary in individual cases.

Heat transfer at high temperatures and under the influence of inert gas

The modelling of the heat transfer remains to be an optimisation task in the

development of computer codes. With regard to CFD codes, the development of boiling

models is still in its initial phase. The influence of non-condensable gases on the heat

transfer in the steam generator and the influence of these gases on condensation has

26

been modelled basically, but should be extended to higher pressures and

temperatures.

Effectiveness of accident management measures

The UPTF-TRAM and PKL-III D tests established the basis for the investigation and

modelling of thermal-hydraulic phenomena in connection with accident management

measures. The corresponding findings have to be implemented in the development of

the 3D code and the final validation.

4.2.2 Thermal hydraulics in enhanced safety concepts

In future, the assessment of thermal-hydraulic processes in enhanced safety concepts

requires the further development and application of the CFX code for the detailed

description of the two-phase flows. In parallel, large-eddy simulations for two-phase

flows will be performed.

Thermal- and fluid dynamics during supercritical pressure

The availability of experimental data on the development of analytical heat transfer

correlations and on the description of the transition from subcritical into supercritical

condition will be required in the future. The data are required with regard to stationary

and transient conditions during normal operation and in accident situations and serve

as basis for a more detailed core design, further material developments and statements

on reactor safety.

27

Effectiveness of passive systems

Experiments on the behaviour of passive systems for residual heat removal from the

RPV and the containment as well as on the dynamic behaviour of passive impulse

generators could be conducted at the TOPFLOW test facility.

Development of methods for safety assessments

With regard to the maintenance of an independent assessment basis, the development

of methods for the assessment of advanced safety concepts also has to be continued

according to the state of the art for international developments.

4.2.3 Contributions to international verification of system codes (e. g. RELAP, CATHARE)

Within the framework of the OECD International Standard Problem No. 45 (ISP-45), the

QUENCH-06 experiment is used for blind calculations with eight different computer

codes (e. g. RELAP, CATHARE) by 21 participants from 15 nations under German

leadership. The evaluation of the calculation results is completed for the most part.

Most of the results more or less agree with the experimental data. The reasons for a

number of clear deviations will have to be investigated in detail.

28

4.2.4 Reactor physics for light-water reactors

Development of methods for availability of nuclear cross sections

The handling of safety-relevant issues of reactor physics requires both computer codes

and the data of the nuclear cross-sections of such a high quality that these are

adequate for realistic calculations of very complex configurations and the analysis of

accidents. The latest and best validated nuclear cross sections available world-wide

are compiled and processed by complex mathematical methods in such a way that

they can be used as basic data for computer codes applied for calculations on

criticality, burn-up, etc.

The state of knowledge on nuclear cross sections reached so far is summarised in the

European library Joint European File (JEF) 2.2 and the US-American correspondent

ENDF/B V and VI. The previous validation concerns fuels for LWR up to an enrichment

of 5 %. Regarding the realisation of new fuel concepts, further validation of the nuclear

data is necessary.

The data on nuclear cross sections are subject to permanent improvement by means of

recent measurements and continual evaluation. The processing of the respective latest

cross-section data for state-of-the-art programs remain to be an important topic for the

future.

Development of time-dependent neutron transport calculation codes for transient analyses of reactivity-initiated accidents

The aim is to assess the inaccuracies due to the application of the diffusion theory

approach for the analysis of reactivity transients and to remove them, if possible, by

neutron transport calculation codes. In the end, 3D transport methods shall be

integrated into diffusion codes.

The codes shall be extended in such a way that they can calculate heterogeneity

effects resulting from new refuelling strategies, modes of operation or accident

scenarios. It is planned to extend them by the explicit use of the time-dependent

neutron transport equation in such a way that they are suitable for the analysis of fast

29

transients, such as reactivity-initiated accidents. The work shall be performed within the

framework of the Nuclear Technology Competence Pool.

Physics of the reactor core in case of increased burn-up, use of mixed-oxide fuel (MOX) and advanced nuclear fuel

There are discussions on new nuclear fuels for the LWR for the reduction of the

existing amounts of plutonium and minor actinides. In connection with this, the design

of suitable fuel elements, their burn-up behaviour and the reactivity coefficients of the

reactor cores have to be analysed.

The established codes of reactor physics were developed for UO2 fuel. In case of

further increase of the share of MOX, of the burn-up and also of the enrichment due to

the resulting inevitably increasing heterogeneities in the reactor core, these codes have

to be further developed. So, e. g., OECD benchmark activities in 1995 revealed a

considerable research demand regarding the further development of reactor physics

calculation methods for reactor cores with increased deployment of plutonium.

The aim of future research work is the determination and assessment of the safety

aspects with regard to reactor physics that are given by the increased use of MOX and

increased burn-up in existing and advanced thermal reactors.

Use of advanced nuclear fuels

To reduce the amounts of Pu and minor actinides (MA), new designs of fuel elements

have to be analysed. In connection with this, the use of IMF (inert matrix fuel) is

discussed. IMF is free of the fertile material U-238 so that no fissile plutonium is

produced in this nuclear fuel during operation. In addition to the analysis of amount of

Pu and MA produced in a IMF-loaded reactor (at least partially), the changed dynamic

properties also have to be analysed. This applies both to the feedback coefficients and

to the heat transfer from the fuel into the coolant. The work shall be performed within

the framework of the Nuclear Technology Competence Pool.

30

Development of methods for on-line core monitoring

For strengthening the preventive level of the defence-in-depth concept within

instrumentation and control it is important to develop a model based on core

instrumentation that calculates all measured safety-relevant parameters in the reactor

core, even those that cannot be measured directly. By this, the instrumentation and

control can exert direct control according to the situation and the response values of

the core protection system can be kept close to the course of the operational

processes in dependence on the situation. In this respect, the problem comprises two

aspects. On the one hand, the reactor model has to be sufficiently fast and accurate at

the same time and, on the other hand, the necessary software proofs have to be

furnished to allow for an application in safety I&C. On this item, relevant work has been

initiated.

Further development of Monte Carlo Methods (i. a. reactor dosimetry)

In reactor physics, the Monte Carlo Method serves, among other things, to analyse

complex correlations of criticality calculation, burn-up calculation and fluence

calculation and to simulate the basic processes in detail. The basic problem of all

Monte Carlo Methods is that the stochastic process leads to slow convergence of the

interim results towards the final result which leads to long calculation times. In order to

keep calculation times acceptable nevertheless, efforts are to be taken to accelerate

convergence by variance reduction and self-optimising Monte Carlo Methods and to

realise shorter calculation times by clustering.

31

4.2.5 Interaction between neutron kinetics and thermal-fluid dynamics

Further development and verification of the coupling of neutron kinetics and thermal-hydraulic computer codes; ATHLET/QUABOX-CUBBOX and ATHLET/ DYN3D

In the reactor core, the nuclear phenomena and the thermal hydraulic phenomena

closely interact at each location and each point in time. Thus, it was just logical to

couple reactor physics codes with those of plant behaviour with advancing computer

technology in such a way that it is possible to calculate both phenomena

simultaneously. This coupling was realised world wide. Since then, the validation for

PWR und BWR has also been intensified at an international level, e. g. by OECD

bench marks.

Until now, validation has only been realised exemplary by means of selected accident

sequences. A comprehensive in-depth validation of enveloping accident scenarios

remains to be an important topic of research for the future.

Today, the coupling of the neutron kinetics computer code QUABOX/CUBBOX with

ATHLET and of the neutron kinetics computer code DYN3D with ATHLET provides

versions for square and hexagonal fuel element geometries. A further development of

these calculation models is required for the improved description of local and rod-to-rod

effects. This also applies to the development of simplified mixture models for the

determination of the entry conditions at the core from the circuit conditions and the

improvement of the fuel rod calculation by coupling of fluid models including transverse

flow or coupling with multi-dimensional CFD codes for single-phase flows.

In order to keep the coupled computer codes manageable at all in the long run

regarding the efforts involved, an acceleration of the codes by consequent application

of clustering (multiple processors and computer networks) is necessary. Software

standards have been developed for these computer architectures.

32

RELAP/PANBOX

The RELAP5/PANBOX/COBRA code system for sensitivity analyses has to be further

developed for selected design-basis accidents of reactors in operation. In this respect,

special emphasis is laid on the advanced modelling of the interaction between neutron

kinetics and thermal hydraulics. The model is developed in close European co-

operation.

Further development of multi-dimensional fluid dynamics codes/physics

For transients with very large local power differences, mixing processes in the reactor

core become relevant. In the foreseeable future it is not realistic to model the reactor

core in detail in CFD codes. For this reason it is currently being analysed

experimentally and theoretically how far a flow-mechanical treatment of the core as a

porous body is reasonable and feasible. To clarify this issue, a test facility was set up

which realises transverse flow in a heater rod bundle and allows a detailed

measurement of the velocity field with laser-optical methods.

Investigation of recriticality in case of core degradation

According to today’s understanding of safety in nuclear technology, the control of

accidents should also include core melt accidents. Most of these safety considerations

are solely based on a thermal hydraulics point of view with emphasis on the coolability

of the melt and its interaction with the RPV. It is postulated that nuclear chain reactions

do not take place during core degradation. This postulation has to be verified, because,

e. g. the CORA core melt experiments showed that melting and flow-off of the absorber

and control rods may occur during the heat-up phase. Flooding of a partially degraded

core with partially molten down control rods might cause recriticality. Until now,

criticality conditions for reactor cores, different degradation levels, different fragment

sizes, different material compositions and different moderation ratios were analysed in

form of parameters by Monte Carlo programs. The methods were qualified by

recalculation of a large number of critical experiments under different neutron-physical

conditions and in international benchmark problems. Since the qualification could not

33

be finalised with this, systematic analyses on recriticality of core conditions during core

degradation should be continued.

4.2.6 Fuel rod (-element) behaviour

Further development of analysis tools for the assessment of fuel rod behaviour

Operationally realistic transients

In German PWR und BWR those fuel elements are increasingly used that are

characterised by longer operational times in the core, higher burn-ups, new fuel

composition (use of MOX and Gadolinium) and also by new cladding alloys. This use

leads to changes in the material behaviour of the fuel rods. So, the accumulation of

hydrogen in the cladding material (corrosion) due to operation caused by a long service

time of the fuel rod leads to significant changes of the cladding material properties. The

assessment of the cladding tube strength during operational transients requires that

these material properties are adequately considered in the tools applied for the safety

analysis (e. g. THAM method). Consequently, a further development of this analysis

tool is necessary.

Loss-of-coolant accident

Within OECD/NEA it is discussed in working groups to which extent the current

practice of setting limit values for loss-of-coolant accident (LOCA) and reactivity-

initiated accidents (RIAs) still is admissible under the changing material properties.

New crack propagation models, e. g., are discussed in connection with corroded fuel

rods which show interrelations others than those known until now. If such corrosion

mechanisms would be involved, break criteria would have to be redefined and would

thus also have an influence on the determination of the extent of core degradation in

case of a loss-of-coolant accident. Therefore, the aim of the work has to be to upgrade

the calculation methods for the determination of the fuel rod behaviour at high burn-up

under LOCA and RIA conditions in view of the new findings and to validate them by

means of current internationally available experimental studies on fuel rod behaviour.

34

Moreover, the safety proof with respect to core damage in case of a loss-of-coolant

accident requires the increased application of methods on uncertainty analyses,

because the number of model formulations rises considerably with increasing

modelling depth. The SUSA method is applied for the determination of uncertainties.

By means of quantification of the state of knowledge on uncertain model formulations

and model parameters this method is able to achieve an analysis result with defined

confidence level (see also Chapter 4.2.11).

The application of these methods to the determination of the damage extent has not

been realised until now. This presents a new challenge insofar as a damage extent

analysis with several analysis steps, applying different programs and different models

(core refuelling analysis, thermal hydraulics analysis, fuel rod analysis), has to be

performed. Since, the safety margins have been increasingly exhausted in the previous

years by capacity and burn-up increases of the cores, a methodical quantification of the

uncertainty in view of the damage extent for future core refuelling is necessary.

Reactivity-initiated accident

Due to the dynamically fast passage of time and the associated potential hazards,

reactivity-initiated accidents have to be considered especially in the research on

prevention and control. Typical reactivity-initiated accidents are, e. g. ejection of a

control rod, boron dilution transients, ATWS (Anticipated Transients Without Scram),

subcooling transients as, e. g., after leakage in the main steam system, and the

neutron-kinetics/thermal-hydraulic instability of the BWR. For the simulation of this

class of accidents, thermal-hydraulic system codes and nuclear kinetics codes were

coupled world wide and validated in first benchmarks.

The restricted calculation capacity available at that time demanded many

simplifications, such as diffusion approximation, limitation to only a few energy groups

and exclusively stationary calculations. Today, the progress in computer technology

allow to leave these approximations and to consider local, heterogeneous effects by

transition to the neutron transport theory and to simulate their sequences realistically

by explicit consideration of the time dependence of fast reactivity transients. The further

development of the codes by abandonment of all previous simplifications and by the

35

explicit consideration of time dependence is necessary so that they can be used for

safety analyses of reactivity-initiated accidents.

CABRI experiments on LWR fuel rods under accident conditions

The general aim of the CABRI experiments is to back up the fuel rod behaviour under

RIA conditions at increased burn-up. In Germany, the strategies for the use of fuel

elements aim at the transition to considerably higher burn-up and the use of MOX fuel

elements. For the planned higher burn-up values, the current experimental basis

available is world wide assessed as insufficient. For this reason, the OECD-IRSN

CABRI Water-Loop project was initiated with a duration until 2008 and with

participation of the Federal Ministry of Economics and Labour (BMWA) and the

German utilities. The programme at the CABRI test facility (French test reactor of CEA)

includes a series of 12 experiments. LWR conditions are realised by modification of the

reactor from the current sodium loop to water loop. These international experiments

with German participation also have to be accompanied nationally by corresponding

analyses.

4.2.7 Processes during core degradation

World wide, calculation codes for the simulation of accident and severe accident

sequences at water-cooled nuclear power plants have been developed to assess their

sequences and preventive or mitigating accident management measures and to

quantify already available safety margins. The system codes still need improvements to

allow for realistic analyses for an extended accident spectrum.

Contrary to the sequence of the early core degradation phase, i. e. until beginning of

core melt, the phenomena occurring during the later core degradation phase still have

to be investigated.

36

Experiments on the early phase of core degradation during reflooding (QUENCH experiments)

After postulated severe reactor accidents with loss of coolant, the overheated and

potentially partly degraded reactor core is reflooded in the emergency cooling phase.

This may lead to the generation of large amounts of hydrogen in case of exothermal

reactions of zircaloy cladding tubes and structure materials with water or steam. The

material behaviour at high temperatures is investigated in different single-effect

experiments and at the large QUENCH facility under quenching conditions.

Until now, seven experiments have been performed successfully at the QUENCH

facility. A major objective of the latest experiment was to obtain data on the

degradation of B4C control rods at BWRs. Here, the formation of gaseous reaction

products during this degradation, in particular H2, CO2 und CH4, play an important role

as well as the impact of control rod degradation on surrounding fuel rods. The control

rod parameters are representative for nuclear power plants as well as for PHEBUS

FPT3 test conditions. Further experiments are planned.

Core-melt/coolant interaction (steam explosion)

If major quantities of core melt are mixed with water during a core melt accident, this

could lead to a steam explosion impairing containment integrity in the extreme case. In

order to be able to exclude this in a reliable and provable manner, realistic upper limits

for loads by steam explosions (pressure-time sequences, explosion energies) shall be

determined. The maximum possible loads in particular depend on the quantities of core

melt and water involved in the actual explosion. Thus, the inherent limitation of these

masses by heavy but not explosive evaporation during the premixing phase, which has

to take place before the actual explosion, is of particular importance. For this reason,

this effect is intensively analysed experimentally and theoretically. This is done, on the

one hand, by experiments in which the core melt is simulated either by a large number

of small hot but solid spheres (QUEOS facility) or by molten AL2O3 (PREMIX facility).

On the other hand, computer codes are developed and verified for the description of

multi-phase multi-component flows (MATTINA, MC3D) that shall particularly allow for

the application of the experimental results to accident situations. These programs shall

serve the mathematical modelling of the premixing and the actual explosion.

37

The participation in the SERENA (Steam Explosion Resolution for Nuclear

Applications) programme of the OECD/NEA will gain in importance. Phase 1 aims at an

assessment of the modelling possibilities regarding the applications for existing nuclear

reactors.

On the basis of experimental findings it was possible to exclude an early containment

failure as a result of steam explosion. However, remaining uncertainties and gaps in

the modelling (in particular with regard to the explosion phase) had to be considered by

postulations that were, in part, relatively rough but strictly conservative. Therefore,

experimental studies on the maximum energy conversion in steam explosions and the

associated pressure-time sequences are conducted at the ECO-1 test facility after

termination of the above-mentioned premixing experiments.

Working versions of the MATTINA and MC3D codes are available. The theoretical

studies on the propagation and expansion phase of the explosion in ECO-1 are

conducted with MC3D. The MATTINA code is still being upgraded for the analysis of

premixing and hydrodynamic processes during the expansion phase. The aim of future

work is to corroborate realistic energy conversion factors physically.

Debris cooling and crust formation

Experimental and analytical work (LIVE project) is performed to investigate the

behaviour of the core melt within the reactor pressure vessel after formation of first melt

areas to the accumulation of melt in the RPV bottom. Here, the main processes during

formation and relocation of melt pools, and the formation and stability of crusts are to

be identified and analysed. Moreover, it has to be investigated in which way cooling

can be restored via flooding of the primary system to control this event sequence.

Future work shall be performed in the Competence Pool and within a European

context.

38

4.2.8 Further development and verification of ATHLET-CD

Since 1987, the ATHLET-CD code system for accidents with core degradation has

been developed and validated on behalf of the competent ministry.

For further development of ATHLET-CD to a comprehensive tool for the simulation and

analysis of severe accidents with core degradation, assessment of preventive or

mitigating accident management measures, for the determination of the source term

from the primary circuit into the containment or directly into the environment in case of

a bypass sequence, and for the quantification of available safety margins, the following

aspects are of primary importance:

• Complete modelling of core degradation and its optimal linkage to the advanced

models of thermal-fluid dynamics also for the late phase of an accident to allow for

the simulation of the entire accident sequence, beginning with specified normal

operation, under consideration of the material interactions, the melt pool formation

in the core area, the flow-off from the core area, the behaviour of the melt or of the

particle bed in the lower plenum, and the interaction with the RPV bottom.

• Modelling of the release of fission products and other aerosols under consideration

of core degradation and the formation of particle beds and melt pools.

• Modelling of the transport and the deposition of fission products and other aerosols

in the primary circuit under consideration of chemical reactions.

• Development of models for the determination of the failure time and failure mode

of the RPV bottom or of a coolant pipe due to high thermal load, and determination

of the break mass flow rate and its components (water, steam, hydrogen, solid or

molten core material, resuspended fission products and aerosols).

For a realistic and largely mechanistic simulation of these events in LWRs and for the

determination of corroborated results (“best estimate“) it is therefore necessary to

complete and extend the ATHLET-CD code with comprehensive and verified models.

Further, it is necessary to couple it with containment programs and to optimise it with

regard to calculation velocity and robustness to enable the application of the code in

simulators. By means of selected single-effect tests and integral experiments, such as

PHEBUS, ARTIST, QUENCH, FOREVER and MASCA or reactor accidents, the code

39

has to be further verified and validated. The verification or validation should be

performed in co-operation with other research institutions.

The aim of the ATHLET-CD development is the realistic simulation of core degradation,

the release of fission products from the fuel, the transport of fission products and

aerosols in the reactor coolant system and the loading of the RPV including the

relevant interactions between these processes. For the description of the core melt

behaviour in the lower plenum, a program module is being developed that models the

physical and chemical processes in the particle bed and in the melt pool.

40

4.2.9 Contributions to international verification of system codes (e. g. MELCOR, SCDAP/RELAP, ICARE/CATHARE)

Coupled problems of thermal hydraulics, neutronics and core material behaviour

The previous work on the international verification of system codes is continued by

participation in selected “International Standard Problems“ of the OECD and

comparative calculations on internationally agreed benchmarks of boiling and

pressurised water reactors.

4.2.10 Integrity of the RPV bottom

Interaction of melt or debris bed with RPV bottom

During the accident in Unit 2 of the Three Miles Island NPP (TMI-2) major amounts of

core material from the core area entered the water-filled lower plenum of the reactor

pressure vessel (RPV). There, the core material formed a melt pool with a solid crust

consisting of solidified melt and an overlying accumulation of fragments (debris bed).

The thermal load of the RPV was far below the loads to be expected according to

simulations. Previous studies showed that considerable experimental and analytical

research work is necessary to understand the processes in TMI-2 and even more

severe accident scenarios and their simulation by means of computer codes. This

includes further evaluations of the RASPLAV (Russian word for melt) tests and the

analytical accompaniment and evaluation of the MASCA tests, also performed within

the framework of an OECD project.

Coolability (internal and external cooling)

The general aim is to point out the possibilities of cooling of the relocated core material

in the lower plenum and of maintenance of RPV integrity with regard to accident

management measures. This requires the analysis of the thermodynamic behaviour of

the melt pool, the debris bed, the efficiency of a gap cooling, if existing, and the

41

interaction of a melt pool with the surrounding crust and of the debris bed with the RPV

wall.

The FOREVER experiments conducted in Sweden offer the possibility to record and

analytically evaluate the descent of melt during RPV creep, the distribution of heat

flows under consideration of crust formation and formation of a gap between melt crust

and RPV wall.

4.2.11 Development of methods for the assessment of the reliability of computer codes also with regard to the applicability to real NPPs

The development of the SUSA (Software System for Uncertainty and Sensitivity

Analysis) method for the quantitative assessment of the reliability of calculation results

has reached an advanced level of maturity. It was applied successfully in several

thermal-hydraulic accident analyses. In a computer code, such as ATHLET,

parameters are processed whose actual value is only known approximately. The state

of knowledge on these parameter values not known precisely can be presented by a

distribution. If these distributions are taken into consideration, it is also possible to

obtain distributions for the calculation results. Only with these it is possible to assess

the distance to the specified maximum values of safety-relevant parameters.

In particular, this also allowed the determination of parameters having the greatest

influence on the final result of the calculation. This also gives indications to the

necessity of future research work on the reduction of identified uncertainties.

In future, the method is also to be applied to processes in the containment, to three-

dimensional thermal-hydraulic analyses and to the fields of fuel rod behaviour and

reactor physics.

42

4.3 Core melt in the containment, steam explosion, hydrogen distribution/combustion and countermeasures, fission product behaviour in the containment


“The integrity of the containment as the last barrier against the release of radioactive

substances into the environment must be assessed for extremely improbable accident

sequences. A realistic assessment requires deepening today’s knowledge on incident

and accident sequences and on the efficiency and reliability of measures to avoid

undue containment loads.“




Technical Fields - Core Melt in the Containment, Steam Explosion, Hydrogen Distribution/Combustion and Countermeasures, Fission Product Behaviour in the Containment

Chapter Topic Areas Research Institutions Remarks4.3.1 Core melt in the containment

• experiments (e.g. DISCO, DCH, KJET) FZK

• analytics: - model development (multi-dimensional) FZK, PT R (RUB), GRS

- model development (fast running) GRS

• experiments (e.g. COMET, OECD/NEA MCCI) FZK, PT R International project MCCI

• analytics (e.g. CORFLOW) FZK, IKE, FZJ, GRS Competence Pool

• experiments (e.g. ECO, OECD/ NEA SERENA) FZK, PT R International study SERENA

• analytics (e.g. MC3D, MATTINA) - development FZK - application, OECD GRS, IKE

• experiments (e.g. OECD/NEA MCCI) PT R International project MCCI

• analytics (e.g. WECHSL) FZK* , GRS *) analytics for ECOSTAR

4.3.2 GRS

GRS

GRS, TUM

FZK, GRS

FZK Loss of competence

- analytics GRS, PT R

- experiments PT R

Thermal hydraulics and fission product behaviour

Molten core/concrete interactions

Short-lived phenomena

Stratification phenomena

Coupling with CFD

Simulation of conventional fires (e.g. GASFLOW)

Thermal hydraulics during H2-combustion (e.g. GASFLOW)


Specific Topics

Core melt release into the containment

Spreading, relocation, cooling and retention of core melt

Core melt/coolant interactions, steam explosion

Table 3: Main Research Topics of the Research Institutions in the Field of Core Melt in the Containment ... 1 v 3


Chapter Topic Areas Research Institutions RemarksSpecific Topics4.3.3 Hydrogen behaviour FZK , GRS

• experiments (e.g. 12 m-pipe, RUT) FZK

• analytics (e.g. DET-3D, INCA) FZK, GRS, PT R

FZK

• experiments FZJ

• analytics (e.g. REKOS, igniters) FZK, GRS

4.3.4 Fission product behaviour FZK , GRS , PT R Analytical accomp. PHEBUS

GRS, PT R (IKE)

GRS , PT R Loss of compentence possible in future

• experiments (e.g. KAREX) FZK

• analytics GRS

• experiments (e.g. ThAI facility) PT R • analytics GRS, PT R

• experiments (e.g. ThAI facility) PT R

• analytics GRS

Loads on containment structures

Chemical and transport behaviour of iodine

Transport and deposition behaviour of aerosols and fission products

Impact of transient events on the fission product and aerosol behaviour (dry and wet resuspension, steam explosion, HPME, DCH, H2-combustion and fires)

Release of fission products from the sump

Hydrogen combustion (incl. transition criteria between combustion regimes)

Countermeasures

Release of fission products and aerosols into the containment (e.g. PHEBEN-2)Fission product and aerosol release from the core melt

Hydrogen distribution



Chapter Topic Areas Research Institutions RemarksSpecific Topics


• analytics GRS


• analytics GRS

4.3.5 Definition of primary source terms GRS, IKE

4.3.6


GRS, FZK, IKE

4.3.7 Accident instrumentation and diagnosis methods

• instrumentation(e.g. utilisation of experiences from the QUENCH experiments) FZK

• diagnosis methods TUM

Interaction between other technical systems (e.g. recombiners) and fission products and aerosols

National and international consultation

Impact of spray systems on the aerosol and fission product behaviour, in particular on iodine



Fig 8: Personnel Prognosis for the Topic Areas in the Field of Core Melt in the Containment ...

Core Melt in the Containment, Steam Explosion, Hydrogen Distribution/Combustion and Countermeasures, Fission Product Behaviour in the Containment

0

5

10

15

20

25

30

35

40

45

2002 2003 2004 2005 2006

Man

Yea

rs

Definition of primary source terms

Core melt in the containment


Fission product behaviour


Accident instrumentation and diagnosis methods

Hydrogen behaviour

Auswertung Cluster Nov.xls

Fig. 9: Personnel Prognosis of the Technical Field "Core Melt in the Containment" (EVK: Evaluation Commission; Mean Value 1996-1998)

Core Melt in the Containment, Steam Explosion, Hydrogen Distribution/Combustion and Countermeasures, Fission Product Behaviour in the Containment

0

10

20

30

40

50

60

70

80

90

EVK 2002 2003 2004 2005 2006

Man

Yea

rs


43

Since 1994, the Containment Code System (COCOSYS) has been developed, as well

as the German-French integral code ASTEC (Accident Source Term Evaluation Code)

in co-operation with a French partner. COCOSYS is mainly based on mechanistic

models for the comprehensive simulation of all relevant processes and plant states

during severe accidents in the containment of light-water reactors, also covering

design-basis accidents. An important aspect is the consideration of interactions

between the individual processes, as e. g. thermal hydraulics, hydrogen combustion,

aerosol and nuclide behaviour, and, where applicable, core melt after RPV failure. The

integral code ASTEC is partly based on simplified models and correlations. This code

shall serve to simulate the entire sequence of severe accidents requiring less time.

In addition, the US-American code MELCOR is used for the complete simulation of

accidents with core degradation. By means of detailed experimental results model

improvements and extensions as well as validation work shall be realised. Further,

work is performed within the framework of an EU project on the development of a fast

real-time code to be used in emergency management (RODOS decision support

system) and as input for calculations of atmospheric transport and dispersion.

4.3.1 Core melt in the containment

Core melt release into the containment

The processes in connection with melt release into the containment have a decisive

influence on the further accident sequence. The type and sequence of the RPV failure

mode and, above all, the pressure under which the core melt in the containment occurs

massively influence the further accident sequence. Depending, above all, on the

pressure the core melt is distributed in the containment and interacts with its

atmosphere. It is necessary to deal extensively with all phenomena that are decisive for

a substantiated assessment of hazard potentials and retention possibilities.

However, it is hardly imaginable that this will result in a mechanistic description. The

plant-type-specific correlations rather should be derived from experiments and

analyses. It is planned to conduct corresponding experiments.

In case of RPV failure under limited internal pressure, the released core melt may

disperse and partly be discharged from the reactor cavity. Which part of the dispersed

44

melt is retained in the adjacent compartments and which part reaches the dome of the

containment together with the generated hydrogen strongly depends on the geometry

of the RPV environment. The different aspects of melt dispersion and direct

containment heating (DCH) in case of moderate failure and different failure types are

investigated by experiments and analyses (by means of multi-dimensional computer

codes). It is planned to establish a sound database especially for the assessment of

the retention capability of the reactor cavity and the impact of core melt on the

containment and its safety components. The applicability to the real reactor case shall

be enabled by appropriate models and tools.

Spreading, relocation, cooling and retention of core melt

In the past years, experiments and analyses have been performed on these issues, as

e. g. COMAS- and KATS spreading tests, development of the LAVA code on melt

spreading and relocation, analyses on porosity formation and coolability during water

injection from the bottom (COMET concept) and from the top, as well as work within

the framework of international projects.

Experiments on the interaction of melt jet with protective and sacrificial materials in the

reactor cavity are being continued.

Experimental studies concentrate on the long-term behaviour of core melt in the

concrete foundation and on the impacts of potential countermeasures to achieve

cooling and fast solidification.

Core melt/coolant interactions, steam explosion

An important aspect within the framework of comprehensive safety analyses for

nuclear power plants is the safety assessment regarding interactions of the molten core

material with the water within the reactor pressure vessel (in-vessel steam explosion)

occurring during core melt accidents and interactions of the core melt discharged from

the RPV with the water in the reactor cavity of the containment (ex-vessel steam

explosion). Such a steam explosion may considerably exacerbate a severe accident.

Further, adequate knowledge and detailed statements on the impacts of melt/water

45

interactions are also necessary for the assessment of potential accident management

measures with water injection for cooling of an already partly molten core in the RPV or

with flooding in the reactor cavity.

A survey prepared within the framework of the OECD/NEA (SERENA) on the studies

on steam explosion performed in the past years showed that there are still safety-

related questions to be solved in the field of steam explosions, particularly with regard

to modelling.

A reassessment of the safety significance of steam explosion on the basis of the

current state of knowledge is aimed at.

Molten core/concrete interactions

In the case of RPV failure during a severe accident, core melt enters the reactor cavity

where it reacts with the concrete structure. The WECHSL code is applied for the

simulation of this molten core/concrete interaction.

In this field, further studies are required to enable more precise quantification and more

realistic assessment of potential accident sequences for medium- and long-term

thermal-mechanical loads on the containment in case of severe reactor accidents with

core melt spreading in the containment. On these issues, experiments and analyses

are performed, e. g. CORESA (Corium on Refractory and Sacrificial Materials)

experiments, development of models for molten core/concrete interaction (e. g. in

COCOSYS and ASTEC) as well as work within the framework of the OECD MCCI

(Molten Core Concrete Interaction) project. In this respect, there is considerable

research demand especially on 2D long-term erosion, at both the national and

international level.

46

4.3.2 Thermal hydraulics in the containment Short-lived phenomena

Models for the simulation of short-lived flows have been implemented in COCOSYS by

way of unifying and supplementing already existing flow models. These supplements

have to be investigated for their impacts on code performance and for potential impacts

on the long-term behaviour.

Stratification phenomena

In the last 20 years, so-called “lumped parameter” codes have been developed for the

simulation of the thermal hydraulics in the containment and successfully validated at

the national and international level. Today, they represent a proven element of many

licensing procedures that was lately applied in Germany in connection with the

backfitting of the existing plants with catalytic H2-recombiners.

These lumped parameter codes do not consider the impulse so that there are

restrictions in the simulation of local velocity fields and so-called jet flows.

The national ThAI “benchmarks” on containment thermal hydraulics, that have just

been finalised, showed that the lumped parameter codes (COCOSYS in this case) are

also able to precalculate typical processes of containment thermal hydraulics correctly.

The International Standard Problem (ISP) 47, currently being dealt with at three

different test facilities (i. a. ThAI) will provide further knowledge on the current capability

of the lumped parameter codes, field and CFD codes as well as indications to further

developments that are necessary.

47

Coupling with CFD

The lumped parameter codes do not consider the impulse equation so that there are

restrictions in the simulation of local velocity fields and so-called jet flows. In order to

overcome the general restrictions of lumped parameter codes and to benefit from the

advantages of CFD codes first couplings between the two types of modelling were

realised.

Simulation of conventional fires

The models for the simulation of oil and cable fires are currently being upgraded.

3d-simulation programs are developed for the optimisation of fire protection at German

nuclear power plants which can be used for the development of improved fire

protection concepts.

Thermal hydraulics and fission product behaviour

The behaviour of the fission products in the containment and thus the extent of

radioactive releases into the environment of a nuclear power plant in case of severe

accidents are considerably dominated by the thermal hydraulic conditions in the

containment. Extensive experimental and analytical work was performed on this issue

at the national and international level. In Germany, this work includes, above all, the

DEMONA, VANAM, KAEVER and ThAI experiments. The analytical work concentrates

on the development of the COCOSYS containment code system and of the integral

code ASTEC which also consider the interaction between the different phenomena, as

e. g. the thermal hydraulics with the aerosol and nuclide behaviour.

48

4.3.3 Hydrogen behaviour

Hydrogen distribution

The distribution and mixture of the hydrogen released during a core melt accident in

the accident atmosphere are largely understood. Phase transition processes of water,

such as boiling of the sump and condensation at the surface (mass and heat

exchange), which interact with the convection in the containment, still have to be

analysed quantitatively. This also has to include the consideration of active and passive

measures of long-term heat removal which are able to counteract inertisation of the

containment atmosphere by steam.

Participation takes place in international programmes on the validation of multi-

dimensional codes, e. g. on the basis of TOSQAN, MISTRA and ThAI experiments.

Hydrogen combustion

With its considerable, fast pressure increase, a combustion of the released hydrogen

may impair the retention capability of the containment, in particular if transition into

detonation occurs. Thus, special attention has to be paid to the implementation of so-

called DDT (deflagration-to-detonation transition) criteria. Lumped parameters and

three-dimensional CFD combustion models therefore have to be further developed and

validated with regard to the influence of the combustion velocity by the spatial

conditions in the containment (branchings, merging of flame fronts, jet ignition etc.).

So, e. g., the scope of application of the CFD code for fast turbulent combustion

(e. g. COM3D) is extended by models for fire extinguishing processes, convective heat

loss, porosity and transition from laminar to turbulent flames. Further, the verification is

provided for lean mixtures and H2-gradients.

The CFD code systems (COM3D und GASFLOW) on turbulent hydrogen combustion

in complex geometries are validated by means of the available experimental data of the

large-scale RUT facility of the Kurchatov Institute.

49

Loads on containment structures

Models are developed and numerical tools validated that allow for the determination of

containment loads on the basis of accident-typical scenarios and plant geometries from

processes during different types of combustion. In case of fixed flames there is a

particular risk of thermal loads, whereas mechanical loads prevail in the case of fast

combustion that may lead to the failure of containment structures. Preliminary studies

are required to determine to which extent future thermal-hydraulic and structure-

mechanical models have to be coupled.

Countermeasures

The Reactor Safety Commission (RSK) recommended to equip PWR containments

with catalysts to counteract the accumulation of hydrogen in the containment

atmosphere. In experimental studies and with the development and validation of

models, calculation tools were made available that can be used to demonstrate the

operating efficiency of catalysts under the different imaginable accident conditions. In

this respect, the possibility of ignition of the mixture due to a superheated catalyst in

case of massive hydrogen release also has to be taken into consideration.

Due to the complexity of the processes, experimental studies and model developments

shall be performed to clarify the complex processes of reactor kinetics, temperature

distribution, mass and heat transport and fluid dynamics that take place inside the

recombiners.

Simulation of catalytic recombiners

In COCOSYS, explicit type-specific models were implemented and validated for

catalytic recombiners. Since this mechanistic modelling requires a lot of calculation

time, correlations were derived from these models and qualified by verification.

The GASFLOW code calculates the integral leaning behaviour of hydrogen by

recombiners in a 3D concentration field of a containment. Potential self-ignition effects

50

and the energy released during recombination and its input into the containment

structures are calculated for measured operating conditions of the recombiners.

4.3.4 Fission product behaviour

The containment represents the last barrier against the release of radioactive fission

products into the environment. To assess the radiological source term for releases from

the containment (leakages, intentional pressure suppression or containment failure),

knowledge of the fission product behaviour and time-dependent fission product

distribution in the containment is necessary.

In Germany, this issue is primarily dealt with at the multi-purpose ThAi experimental

facility, where experiments are performed or provided at the pilot plant scale, e. g. on

the fission product behaviour in the containment, on the iodine transport behaviour, on

aerosols and hydrogen, and on the thermal hydraulics under typical accident

conditions.

Release of fission products and aerosols into the containment

During severe accident in light-water reactors with damages to the fuel-rod cladding

and release of fission products from the fuel rods, radioactive material is transferred to

the primary circuit which, in case of leakages, may enter the containment. Moreover,

core melt may also enter the containment in case of severe accidents. Here, the

release of fission products/aerosols and core melt into the containment have a decisive

influence on the further accident sequence. The results of corresponding experiments

(e. g. from the PHEBUS reactor) are used for the validation of COCOSYS/ASTEC.

Fission product and aerosol release from the core melt

For the determination of the fission product behaviour in the containment and their

distribution in the containment, and thus the determination of the assessment of the

radiological source term for releases from the containment, knowledge on the release

of fission products and aerosols from the core melt is necessary. The release of gases,

51

aerosols and fission products from the core melt is decisively influenced by the further

melt behaviour, in particular by the concrete-melt interaction. For COCOSYS and other

codes there is still considerable demand regarding the simulation of fission product and

aerosol release from the core melt.

Impact of transient events on the fission product and aerosol behaviour

The scarce information from experiments on a series of phenomena (high-pressure

melt ejection (HPME), direct containment heating (DCH), hydrogen combustion, dry

resuspension) are to be used in future for an analytical description. Before conducting

major and consequently technically challenging experiments, parameter studies should

be performed to quantify, at least, the order of magnitude for a potential impact of these

transient events.

Release of fission products from the sump

Analytical model approaches on the release of fission products suspended in liquids as

function of gas amounts transferred to the sump were developed e. g. in context with

the RECOM code development. However, in view of the insufficient scaling of the

experimental studies conducted in this field until now, these model approaches are not

fully corroborated. Due to the decisive impact on the source term in the long-term

phase, more experimental and analytical activities are necessary.

Transport and deposition behaviour of aerosols and fission products

Detailed knowledge on the transport and deposition behaviour of aerosols and fission

products is necessary to determine the time-dependent fission product distribution in

the containment. On this issue, a series of experiments, as e. g. VANAM, KAEVER,

and the extensive ThAI experiments and analytical work on code development

(e. g. FIPLOC, COCOSYS) were conducted. Further experiments at the ThAI facilities

and analytical studies are required.

52

Chemical and transport behaviour of iodine

The relevance of iodine during severe accidents is given by its high volatility, its

chemical activity and its radiotoxicity. The iodine aerosols released into the

containment are deposited on the surfaces or enter the sump. Via chemical reactions,

also under the influence of the high-radiation field, gaseous iodine is released

additionally into the containment in the longer term and can be absorbed at the

surfaces or enter the sump with aerosols or droplets. Here, the transport processes

between sump and containment atmosphere and to the surfaces or away from the

surfaces play an essential role for the source term. Experimental and theoretical work

over many years led to the development of the iodine accident codes, such as IMPAIR.

At present, the transport behaviour (phase and mass transfer) of iodine is analysed at

the pilot plant scale (ThAI experiments). Further experimental and analytical demand

for studies exists in the fields of “iodine behaviour in case of a superheated

atmosphere”, “iodine behaviour under the influence of ozone”, “BWR-specific iodine

chemistry” and “interaction of iodine with aerosols”.

Impact of spray systems on the aerosol and fission product behaviour, in particular on iodine

The spray systems often installed in nuclear power plants serve to reduce the pressure

and temperature loads occurring under accident conditions and to reduce the source

term released into the environment by wash-out of aerosols and fission products. In

order to enable the simulation of these processes, spray models like IVO (Finland) and

MARCH (USA) were developed. These models have already been implemented and

compared to each others at, among others, in the German COCOSYS containment

code. Their reliability still has to be verified by corresponding calculations for which

appropriate experimental data can be delivered, among others, by the ThAI facility.

Interaction between other technical systems

As with spray systems, other technical systems may also have an impact on the

containment atmosphere. Here, e. g., the impact of H2-recombiners on the aerosol and

iodine behaviour (e. g. iodide decomposition, generation of volatile iodine, change of

53

the aerosol size spectrum) has to be analysed. The design of the ThAI facility enables

the conduction of corresponding experiments on the validation of COCOSYS.

4.3.5 Definition of primary source terms

A comprehensive assessment of the source terms in the containments under

consideration of core melt, fission products, hydrogen and steam is necessary, e. g., as

input parameter for COCOSYS. The central interactions for the overall development of

accident consequences and retention measures are to be identified and modelled.

4.3.6 Development of methods for the assessment of the reliability of computer codes also with regard to the applicability to real plants

In Germany, the SUSA method was developed to quantify uncertainties (see 4.2.11).

Until now, it has primarily been applied to the issue of thermal hydraulics in the reactor

cooling system.

Now, the level reached regarding the simulation of phenomena and processes in the

containment also allows the corresponding application of SUSA. In the medium term, it

is planned to apply SUSA also for the coupled system ATHLET-CD/COCOSYS.

The assessment of a problem-oriented modelling (necessary degree of detail,

extrapolability, required calculation time, user friendliness) is particularly required in

connection with the work on the German-French integral code ASTEC. On this issue,

fundamental contributions are to be expected from the EU-sponsored EVITA project

and its continuation within the 6th EU Framework Programme.

4.3.7 Accident instrumentation and diagnosis methods

In the case of an accident or severe accident it is necessary that the operating

personnel has a detailed picture of the respective status of the event sequence to

54

enable reliable prognoses on the safety of the barriers and the quantification of the

source term. The experiences gained from realistic experiments with measurement and

diagnosis methods shall be evaluated regarding the application in reactors and

implemented for the specification of an accident instrumentation.

55

4.4 Development of methods for probabilistic safety analyses, for instrumentation and control, as well as for the assessment of the human factor


“With regard to the improvement of the tools for identifying deficiencies in the plant

design and operating procedures, probabilistic methods must be developed further and

the existing assessment uncertainties must be reduced.“





4.4.1

4.4.2 Level 3 Development of models for local weather forecasts FZK, GRS Competence Pool

Atmospheric dispersion of radionuclides FZK

4.4.3 Level 1 and Level 2 CCF models GRS

Fire GRS, FZK Competence Pool

Earthquake (incl. Paleoseismology) PT R, GRS

Reliability of active components under accident conditions GRS

Extended modelling of uncertainties GRS, PT R (HTWS) Competence Pool

Dynamic PSA GRS

4.4.4 Model evaluation by means of exemplary PSA GRS

4.4.5 Development of methods for instrumentation and control

Safety assessment of computer-based systems (software and hardware) PT R (ISTec)

Reliability of computer-based systems for PSA PT R (ISTec)

4.4.6 Human behaviour Probabilistic approach (assessment of human reliability) PT R (TUB), GRS

Systematic approach (organisation, structures, implicit standards) PT R (TUB)

4.4.7 Technical systems to support human performance

Task distribution man / machine (degree of automation) PT R

Prognosis tools (focal points: test control room, simulators) PT R, GRS

Further development of diagnosis methods FZR (work almost completed) Competence Pool

PT R (ISTec, HTWS)

Technical Field - Development of Methods for Probabilistic Safety Analyses, for Instrumentation and Control, and for Human Factor Assessment

Development of methods for probabilistic safety analyses

Table 4: Main Research Topics of the Research Institutions in the Field of Probabilistic Safety Analyses 4.4 Themen-PSA- 02.12-e.xls 1 v 1


Fig. 10: Personnel Prognosis for Topic Areas in the Field of Probabilistic Safety Analyses

Probabilistic Safety Analyses, Instrumentation and Control and Human Factor

0

5

10

15

20

25

30

35

40

45

2002 2003 2004 2005 2006

Man

Yea

rs

Level 1 and Level 2 Level 3Human behaviour Development of methods for instrumentation and controlTechnical systems to support human performance


Fig. 11: Personnel Prognosis for the Technical Field of Probabilistic Safety Analyses (EVK: Evaluation Commission; Mean Value 1996-1998)

Probabilistic Safety Analyses, Instrumentation and Control, and Human Factor

0

10

20

30

40

50

60

70

80

90

EVK 2002 2003 2004 2005 2006

Man

Yea

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56

4.4.1 Development of methods for probabilistic safety analyses

Within the framework of a PSA, all relevant information on plant design, operating

modes, operating experience, component and system reliabilities, operator actions and

generic safety-related influences are analysed and comprised in an overall assessment

of a plant. A PSA allows to assess the balance of the available safety systems and

equipment, to identify potential deficiencies, to show possibilities of their removal and

to assess the efficiency of accident management measures.

The aim or research work is to further develop the methodical bases and tools for the

performance of PSA and to quantify its reliability. On this issue, further aspects

(common-cause failures (CCF), operator actions, accident management (AM)

measures, ageing and maintenance, plant dynamics etc.) have to be included, new

technical developments (passive components, digital instrumentation and control etc.)

have to be considered and uncertainty and sensitivity analyses have to be performed

on significant parameters.

4.4.2 Level 3

Level 3 of a PSA analyses damage consequences due to accidental releases of

radionuclides. Such analyses were performed within the framework of the “German

Risk Study Nuclear Power Plants, Phase A“ (1979) and in the “Risk-Oriented Analyses

of the SNR-300“ (1982). Today, most of the PSAs are limited to Levels 1 and 2 which

analyse accident sequences within the plant.

Development of models for local weather forecasts

The RODOS system applied for the development of methods for the assessment and

mitigation of radiological consequences due to reactor accidents implicitly includes

models for local weather forecasts.

In order to further increase the reliability of prognoses and the closeness to reality of

the model chains on the atmospheric dispersion of radionuclides, concluding work and

57

an additional validation of the complex mesoscale model chains are necessary. An

adaptation of the model chain actuation with data of the local model - the new weather

forecast model of Germany’s National Meteorological Service DWD - is planned. The

model chain adapted this way shall be validated with further data from field

experiments and verified with agreed criteria for quality assurance. Further experiences

have to be gained that enable an assessment of the remaining uncertainties in the

model calculations. Due to the prototype character of the model chain, these

experiences are directly considered in existing decision-support systems (disaster

response) as e. g. RODOS/RESY, or can be considered in the current model

development of the DWD within the framework of the Precautionary Radiological

Protection Act.

Atmospheric dispersion of radionuclides

The development of dispersion models within the framework of the RODOS decision-

support system has largely been finalised. Further improvements of model contents,

user interfaces, data sets and presentation techniques are realised in close co-

operation with the users. The RODOS will be installed in several European states.

4.4.3 Level 1 and Level 2

The majority of PSAs performed in Germany until now have been Level-1 PSAs

(accident sequences at the plant up to core melt). In the “BWR safety study“ (1993)

and in the “Konvoi PSA“ for GKN-II (2001), the processes within the plant after core

melt were also investigated (Level-2 PSA). These two studies also analysed incidents

(in Level 1 of the PSA) that may occur during shutdown condition, whereas the majority

of PSAs exclusively deal with incidents during power operation.

The methods of Levels 1 and 2 of the PSA are proven and suitable to deliver

corroborated results. However, there are methodical problems limiting the

completeness of the PSA which considerably contributes to a limited reliability of the

results. It is reasonable and necessary to further develop the methods on the basis of

58

the level reached to enhance the reliability of the PSA. This applies, above all, to the

following areas:

Common Cause Failure (CCF) models

The non-availability of redundant systems with highly reliable components is inevitably

dominated by CCF. Due to corresponding countermeasures CCF occur seldom so that

the quantification of their occurrence probability can only be based on operating

experience to a limited degree. Thus, the development and verification of adequate

models for the assessment of CCF are of particular importance.

Fire

In this area, further developments are required, above all, with regard to analyses of

impacts of fire on systems. To some extent, it is difficult to integrate fire-induced

component failures in existing fault trees. In particular, it has not been clarified until

now in which way the different fire-induced failure possibilities of an I&C cable

(interruption, signal change, overvoltage input) with their different consequences can

be considered in the PSA. The simultaneous occurrence of faulty signals in two

redundancies of the measured values acquisition, taking place in the extreme case of

fire propagation, with the potential consequence of faulty actions of the reactor

protection system requires improved analysis methods.

The COM3D code applied for the calculation of hydrogen combustion processes can

also be used for the calculation of fires and fire-extinguishing processes.

59

Earthquake (incl. paleoseismology)

Starting from the USA, where records on earthquakes date back only a few centuries,

paleoseismic methods are currently being developed in several countries. The aim of

these methods is to assess the probability and maximum magnitude of local strong

earthquakes via the determination of strength and time of historical and prehistoric

earthquakes. The applied analysis methods reach from the evaluation of drillings and

excavations to airplane- and satellite-based remote sensing. The analysis concentrates

on the reliability of the results of paleoseismic methods for application in PSAs and the

question whether the seismic hazard in Germany might have to be reassessed.

Reliability of active components under accident conditions2

Data on the reliability of active components under accident conditions are required,

above all, for the probabilistic assessment of accident management measures and the

assessment of system reliabilities in Level 2 of the PSA. Since such data at best can

be obtained to a limited extent directly from operating experience, model developments

- and in critical cases also experimental studies - are required on this issue.

Extended modelling of uncertainties

In Germany, the SUSA method is successfully applied for the performance of

uncertainty analyses. The conventional uncertainty analysis performed with SUSA

identifies the influence of state-of-knowledge uncertainties in parameters, model

assumptions, phenomena and, in case of application of numerical algorithms, on the

result of calculation models. For the results of probabilistic safety analyses, however,

there is a distinction to be made between uncertainties due to stochastic variability

(aleatory) and knowledge uncertainties (epistemic). A concept was developed for an

approximate analysis of the epistemic uncertainties of a probabilistic dynamics

calculation with a stochastics module. The application-level design and realisation of

2 Reliability of passive components; see Chapter 4.1.1

60

this concept seems to be promising concerning a better possibility for the assessment

of the results in future PSAs.

In addition, the potential of knowledge-based methods of information processing (fuzzy

set theory, artificial neural networks etc.) shall be pointed out for the PSA for the

treatment of fuzziness and examined exemplarily.

Dynamic PSA

For several years, emphasis has been laid on the further development of methods for

the performance of dynamic probabilistic safety analyses. Until the end of 2001, the

MCDET (Monte Carlo Dynamic Event Tree) analysis tool had been developed which

enables the total consideration of the interaction between the dynamics of an event

sequence and the stochastic influences within the framework of a PSA, and which

delivers dynamic event trees as a result developing along a time axis. In order to allow

for a quantitative assessment of operator actions in connection with the dynamic event

sequence, presently a tool is missing for the consideration of the influence of human

interactions on the dynamic event sequence of , e. g., an incident. The HRA (Human

Reliability Analysis) methods currently applied in PSAs do not fulfil this requirement. In

the next years, methodical tools have to be developed by means of which actions of

operators at nuclear power plants can be registered, modelled and integrated into

dynamic event trees.

4.4.4 Model evaluation by means of exemplary PSA

Recent PSAs showed that the absolute frequencies of damage states are very low.

This increases the relative significance of very seldom events that have not been

considered in PSAs by now or only in form of estimates.

The essential methodical problems currently limiting the reliability of the analyses can

be summarised as follows:

The spectrum of the “initiating events” analysed until now is incomplete, particularly as

far as accidents are concerned that may result from external impacts (e. g. fire,

61

earthquake, aircraft crash, extreme weather conditions, flooding, lightning) Until now, it

has not been possible to make a reliable assessment whether such events lead to

significant contributions to the accident risk. Moreover, only accidents during power

operation have been considered in Level 2 of the PSA until now.

With regard to the assessment of the reliability of safety-relevant systems there are still

methodical problems in various areas that have to be solved or at least reduced

(software-based instrumentation and control, knowledge-based operator actions,

organisational influences, failure of large passive components).

In some areas (effects of deboration accidents, phenomena after core melt), the

methods for the simulation of accident and severe accident sequences have to be

further developed.

The following methods have to be upgraded:

• Determination of the frequencies for major releases of radionuclides also in

case of very improbable initiating events due to extreme high water and

weather conditions and the failure of containers with high energy content, and

• assessment of the influences from plant management and organisation on the

reliability of components and operator actions.

The advanced methods shall be tested within the framework of a Level-2 reference

PSA.

In addition to accidents during power operation (nominal load), the PSA should also

consider partial load operation, start-up and shutdown processes and shutdown plant

states.

Further, the PSA should both consider internal and external impacts affecting several

systems (internal: fire, flooding; external: aircraft crash, earthquake, flood, lightning).

62

4.4.5 Development of methods for instrumentation and control

Safety assessment of computer-based systems (software and hardware)

The application of computer-based digital I&C systems at nuclear power plants offers

the possibility of a comprehensive information processing for the shift personnel and of

an improved self-monitoring and diagnosis function, and can thus contribute to

increase nuclear safety.

At present, it can be noted that these systems are also increasingly applied in the field

of safety instrumentation and control. Methods and tools for the qualification of modern

hard/software systems have to be adapted to the fast technical progress regarding

software-based I&C systems. In addition, new general issues arising with regard to the

safety assessment of these systems, as e. g. computer network technologies, have to

be dealt with. For the qualification and safety assessment of these systems, basic

methodical R&D work has to be performed.

Reliability of computer-based systems for PSA

The PSA was successfully established as essential method for the overall safety

assessment of nuclear power plants. Currently, it complies with the state of the art to

deliver the reliability proof for software-based systems by means of qualitative

methods, since modelling of failure mechanisms of software as prerequisite for the

determination of the quantitative reliability coefficients is still only possible in

exceptional cases. In addition, some aspects of the architecture of software-bases

instrumentation and control differ from those of conventional instrumentation and

control which leads to new challenges for the probabilistic reliability assessment. A

PSA, however, requires characteristic parameters for all system components. For this

reason, it is necessary to work on the modelling of processes of occurrence and

removal of software errors and on the specification of limit values (bounds) for the

postulated failure of computer-based systems.

63

4.4.6 Human behaviour

Probabilistic approach (assessment of human reliability)

The relatively course methods on “human reliability” have to be advanced for the

application in PSAs for nuclear installations in order to reduce the uncertainty bands

that exist today. In particular, it is necessary to corroborate the available models and

data, most of them still from the non-nuclear industries, to a larger degree by means of

experiences in the field of nuclear technology. Supplementary to the methods available

so far, which mainly concern “rule-based actions”, methods for the assessment of

“knowledge-based actions, including performance failure, also have to be developed

and tested.

Systematic approach (organisation, structures, implicit standards)

Three Miles Island and Chernobyl showed the decisive contribution of man to safety

and reliability of nuclear installations. Since then, the human factor regarding the

course of events (not only in the field of nuclear technology) has been discussed under

the keyword “safety culture”. While the theoretical clarification of the concept has made

progress in the meantime, there is a lack of practicable instruments for the assessment

of the respective quality of safety culture and adequate methods for its targeted and

sustainable implementation and promotion.

The situation is more or less the same regarding conceptualisation and design of

institutional forms of safety management systems that are oriented towards

organisational structure and management practices to enable the examination of

safety-oriented processes and to ensure their continuous optimisation. Moreover, it

concerns their integration with existing management systems, such as the quality

management system.

The age structure of personnel working in the nuclear section will lead to a loss of

highly qualified experts in the foreseeable future. This raises the question how their

implicit knowledge acquired over decades can be maintained and secured for future

experts. In this respect, appropriate structures of knowledge management have to be

further developed and implemented institutionally.

64

With regard to the three topics mentioned, targeted practice-oriented fundamental

research is required.

4.4.7 Technical systems to support human performance

Task distribution man / machine (degree of automation)

Various reported events at German nuclear power plants caused by software-based

systems and equipment (e. g. blocking of rod insertion, crash of a fuel element cask)

show that malfunctions and failure of software-based technology are often related with

operator actions. In order to better understand the problems associated with the

application of this technology, R&D work on the questions of the relationship

man/machine in automated processes are to be examined fundamentally.

Prognosis tools (focal points: test control room, simulators)

With regard to the examination of measures related to accident management measures

and accident diagnosis and prognoses there are restrictions concerning:

− the simulation and automation of procedures,

− the simulation velocity, and

− the modelling of simulators.

For this reason, it is necessary to further upgrade ATLAS 2000 and similar simulation

tools to enable the realistic examination of the interaction between operator actions and

plant behaviour (e. g. procedure, AM measures).

The tools play an important role as basis for support systems for operators and in case

of emergencies, as well as for improved reliability analyses (Level-2 PSA).

The aim of further development shall be the further increase of reliability of the

analyses and to extend the scope of application with emphasis on three focal points.

65

The first focal point is the extended diagnosis of malfunctions and failures occurred.

The diagnosis can then be used in analyses to identify the initial and boundary

conditions, to accelerate the time-consuming adaptation of the code to the failure, and

to increase the reliability of the results.

A second focal point refers to the acceleration of the computer codes used in ATLAS.

Finally, the third focal point is the development of additional auxiliary tools for analyses

that make the handling of ATLAS more reliable and effective according to users’

experience.

Further development of diagnosis methods

The further development of information processing methods (soft computing) and the

use of digital instrumentation and control opens up the potential for enhancement of

safety of the nuclear power plants in operation by quality improvement in the fields of

modelling, simulation, monitoring and control. For the development of fault diagnosis

and isolation (FDI) methods for the monitoring of sensors, measuring chains and

instrumentations, research and development is required, in addition to the classical

signal processing methods (e. g. frequency analysis, spectral analysis and statistic

analyses), on the application of so-called soft-computing methods with the aim of early

warning signals to increase the safety level. In this respect, multi-sensorial monitoring

methods have to be developed for sensors, measuring chains and instrumentations

that have to be verified in experiments and under real application conditions.

66

4.5 Know-how transfer regarding safety assessments of Eastern reactors


„The improvement of the safety of nuclear power plants of Soviet design is one of the

most pressing tasks to be coped with in co-operation with the Central and Eastern

European Countries. In this respect, the ageing effect due to neutron embrittlement of

the reactor pressure vessel (RPV embrittlement) is of special significance. Western and

in particular German support is indispensable due to the outstanding knowledge

available in Germany in the field of systems engineering.“




Chapter Topic Areas Research Institutions Remarks

4.5.1 Code adaption and development

• ATHLET FZR / GRS

• ATHLET-CD GRS

• COCOSYS (DRASYS), ASTEC GRS

• Coupling of ATHLET with 3D neutron kinetics codes

- ATHLET / DYN 3D FZR

- ATHLET / QUABOX – CUBBOX GRS

- other neutron kinetics codes GRS

FZR

GRS

4.5.2 RPV ageing by neutron embrittlement

FZR

Technical Field - Know-how Transfer regarding Safety Assessments of Eastern Reactors

Development of interactive accident simulators

Exchange of test specimens

Reactor dosimetry

Reactor physics / core data

Specific Topics

(scientific-technical co-operation with Russia, the Ukraine, Bulgaria, Czech Rep., Slovak Rep., Hungary)

Adaptation and verification for VVER and RBMK

(analyses of specimens from Greifswald)FZR

Table 5: Main Research Topics of the Research Institutions in the Field of Safety Assessments of Eastern Reactors 4.5 Themen-Ost-02- 02.12-e.xls 1 v 1


Fig. 12: Personnel Prognosis for Topic Areas in the Field of Safety Assessments of Eastern Reactors

Know-howTransfer Regarding Safety Assessments of Eastern Reactors

0

5

10

15

20

25

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2002 2003 2004 2005 2006

Man

Yea

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RPV ageing by neutron embrittlement Code adaptation and development


Fig. 13: Personnel Prognosis for the Technical Field of Safety Assessments of Eastern Reactors (EVK: Evaluation Commission; Mean Value 1996-1998)

Know-how Transfer Regarding Safety Assessments of Eastern Reactors

0

10

20

30

40

50

60

70

80

90

EVK 2002 2003 2004 2005 2006

Man

Yea

rs

FZR, TU Dresden, HTWS Zittau GRS, TU München

67

The personnel estimate for know-how transfer to Eastern Europe mainly refers to the

co-operation with Russia and the Central and Eastern European countries.

For many years, experiences and results from German reactor safety research have

been used within the framework of scientific-technical co-operation with Russia and the

Central and Eastern European countries planned in the long run. The results of this co-

operation establishes the basis for safety-related studies on VVER- and RBMK

reactors and for the statements on the safety of these plants derived from these

studies. The co-operation has to be continued with the aim to contribute to an improved

safety level of these plants.

Within the framework of WTZ projects (scientific-technical co-operation), criteria and

combustion models for the description of hydrogen behaviour during potential severe

accidents are developed and applied to VVER reactors. The criteria are integrated into

the 3D program GASFLOW. Corrosion analyses are performed on steel in liquid lead

or lead-bismuth, and handling technologies and methods for surface treatment are

developed to reduce the susceptibility to corrosion. Further, experiments on oxidation

kinetics and degradation behaviour of B4C control rods are conducted and analysed

within the framework of work on QUENCH.

4.5.1 Code adaptation and development

Until now, the co-operation has been concentrating on the adaptation and extended

validation of the calculation codes developed in Germany for the design-specific

conditions of Russian reactor types and on the performance of generic accident

analyses. These activities have to be continued with the aim to make the continuously

developed code improvements within the framework of German reactor safety research

available to the partner countries, to increase the scope of validation of the codes

especially for VVER, to extend the coupling of the code and plant models of the partner

countries, and to use the experience feedback from code application in Germany.

68

ATHLET

The thermal-hydraulic code ATHLET remains to be the basis for joint upgrading and

validation activities for accident analyses in the cooling circuit of VVER and RBMK

reactors. In future, the best-estimate- code ATHLET will be of special importance as

reference code for the Russian code development KORSAR. The German methods for

the quantitative determination of the reliability of calculation results will be adopted and

applied by the partner countries, and a considerable experience feedback is expected.

German contributions are delivered by reference calculations with ATHLET and

participation in the development of a concept and performance of counterpart tests at

the ISB and PSB facilities. The corresponding instrumentation for these tests will be

provided. In the years 2002 to 2004, the work will also be supported by the EU project

“Improved Accident Management for VVERs“.

ATHLET-CD

Focal point is the development of methods for the simulation of the melt convection in

the lower plenum and for the calculation of the load on and damage extent of the RPV.

It is planned to apply the methods developed to VVER conditions.

In order to enable the comprehensive use of ATHLET-CD for the simulation of

accidents with core damage in VVER reactors, further model adaptations are

necessary, in particular on heat transfer, on the material-specific melt-down behaviour,

on the melt cooling in the lower plenum, and on RPV damage.

COCOSYS (DRASYS), ASTEC

The code system COCOSYS is applied in several countries of Central and Eastern

Europe for the simulation of accident and severe accident sequences in the

containment and the confinement of VVER reactors. The DRASYS model for the

simulation of dynamic condensation processes was already adapted to the special

conditions involved with the pressure suppression systems of VVER-440 and RBMK

reactors and validated in parts. This work, partly supported by the EU, has to be

69

continued by using new experimental results. Coupled versions of ATHLET/COCOSYS

are of special interest for Central and Eastern European (CEE) countries.

The German-French integral code ASTEC for the evaluation of the source term in case

of core melt accidents has to be validated in co-operation with partners in Russia and

the CEE countries. Here, support by the EU is to be strived at within the 6th Framework

Programme.

Coupling of ATHLET with 3D neutron kinetics codes

A more detailed analysis of different reactivity-initiated accidents requires the coupled

simulation of circuit thermal hydraulics with 3D neutron kinetics. ATHLET is already

coupled with various kinetics codes for the hexagonal fuel element geometry of the

VVER reactors (e. g. DYN3D, BIPR8). A useful experience feedback is expected from

the generic applications of new coupled version but, in particular, from the validation of

ATHLET/BIPR8 by means of measured transients at Russian VVER-1000 plants.

In parallel, the code complex DYN3D/RELAP, developed at the Institute of Physics and

Power Engineering in Obninsk, will be validated by means of operational transients and

by comparative calculations with DYN3D/ATHLET. In addition, analyses of postulated

accident scenarios at VVERs with MOX and CERMET (ceramic-metal) core loading

shall be performed. The assumptions on the thermal-hydraulic boundary conditions of

these transient analyses shall be corroborated by mixture tests at the ROCOM facility

and by the calculation of boron concentration profiles in case of boron dilution

transients.

For the analysis of reactivity transients in RBMK reactors with square fuel assembly

geometry, ATHLET is coupled with the 3D kinetics program QUABOX/CUBBOX. This

combination is successfully applied, to some extent with the support of the EU.

In future, backfitting measures as well as new fuel element designs and new control

rod constructions at RBMK, will require further model adaptations in order to be able

dealing with safety relevant issues such as reactivity behaviour and efficiency of the

secondary shutdown system.

70

This concerns the extension and adaptation of the cross-section library, core loading

with new fuel elements with higher enrichment, the modelling of new rod constructions,

as well as the plant-specific adaptation of the complex reactor control and protection

system and the modelling of the control rod cooling circuit. The models applied are to

be verified by comparison with plant measurements and results achieved by Russian

codes.

Reactor physics/core data

For the generation of cross sections in VVER-1000 fuel elements with increased

enrichment, the KENOREST burn-up program has to be adapted to VVER conditions.

Comparative calculations of the same transients with differently available cross-section

libraries lead to large differences in the results. The causes for this situation have to be

clarified.

Development of interactive accident simulators

Analysis simulators with extensive plant-specific modelling of I&C and systems

engineering, graphical and interactive user interfaces also became accepted for VVER

reactors. On the basis of the simulation software ATLAS, the simulator for VVER-

1000/V-320 is to be upgraded. In addition, an analysis simulator for a VVER-440/V-230

plant has to be developed in co-operation with the Russian simulator manufacturer

VNIIAES. This simulator is particularly required for the analysis of accident

management measures.

71

4.5.2 RPV ageing by neutron embrittlement

Exchange of test specimens (analyses of specimens from Greifswald)

It is planned to determine the degradation of different alloys in dependence of the

neutron fluence and to clarify the microstructural changes. On the basis of the results

achieved, models are to be developed that describe the relationship between radiation

exposure, structural changes and the fracture-mechanic properties.

The taking of specimens from the RPV of the Greifswald reactors and their analysis is

of outstanding significance to the assessment of the radiation-induced material

embrittlement of the VVERs in Eastern Europe. It is strived at taking the specimens

within the framework of an EU project. This shall be preceded by the compilation of

data on radiation exposure and on the operating history within the framework of a WTZ

project.

Reactor dosimetry

It is planned to calculate the neutron and gamma radiation exposures of the RPV and

the accelerated irradiation specimens at VVER-1000, as well as to adjustment of

calculations according to fluence measurements. This shall result in a recommendation

on the routinely application of fluence calculations and measurements for VVER-1000.

72

4.6 Innovative concepts


“The competence existing in Germany should continue to be used in the future with the

aim to further enhance the standards of nuclear safety.”




Technical Field - Innovative Concepts

Chapter Topic Areas Research Institutions Remarks

4.6.1 FZK, FZJ, GRS

• Plant-related safety FZJ,THA, IKE

• Research and further development FZJ

Loss of compentence

FZR

FZK

Loss of compentence

FZK, FZJ, THA, GRS

4.6.2 Avoidance and reduction of plutonium and minor actinides GRS, FZJ, THA

4 1 4

GRS, FZJ, THA

FZJ, THA, IKE

FZK, FZR

4.6.3 Safety of subcritical systems for transmutation FZK, FZJ

FZK, FZJ

Material properties and corrosion behaviour FZK

Thermal hydraulics of heavy-metal coolant FZK, FZJ, FZR

Proton window (e. g. MEGAPIE) FZK

International developments (e. g. Generation 4)

Specific Topics

Significance for strategies and safety

High-temperature reactors

Reactors with fast neutrons

Core design (solid, liquid targets)

Overall concept

Analysis of the technical concept and the safety of innovative international concepts

Reduction strategies for plutonium and proliferation resistance

Advanced fuel strategies for light-water reactors (e. g. uranium-free matrix or thorium matrix in light-water reactors)

Advanced fuel strategies for HTRs (e. g. very high burn-ups, thorium-based fuel)

Plutonium reduction in new critical reactors (e. g. CAPRA, MSR)

Critical molten-salt-fuelled reactors

Light-water reactor moderated with supercritical water (HPLWR)

Lead-cooled fast reactor (BREST 1200)

Table 6: Main Research Topics of the Research Institutions in the Field of Innovative Concepts 4.6 Themen-INNO-02- 02.12-e.xls 1 v 1


Fig. 14: Personnel Prognosis for Topic Areas in the Field of Innovative Concepts

Innovative Concepts

0

5

10

15

20

25

30

35

40

45

2002 2003 2004 2005 2006

Man

Yea

rs

Analysis of the technical concept and the safety of innovative international conceptsSafety of subcritical systems for transmutationAvoidance and reduction of plutonium and higher actinides


Bild 15: Personalprognose für den Fachbereich innovative Konzepte (EVK: Evaluierungskommission; Mittelwert 1996-1998)

Innovative Concepts

0

10

20

30

40

50

60

70

80

90

EVK 2002 2003 2004 2005 2006

Man

Yea

rs

FZK, Karlsruhe University FZR, TU Dresden, HTWS Zittau GRS, TU München FZJ, TH Aachen e.g. Industry, TÜV, Universities

73

Innovative safety concepts are dealt with at many research institutions world wide. It

has to be examined whether these concepts include elements that can be used to

improve the safety level of reactors operated in Germany, or that may contribute to the

safe disposal of radioactive waste by transmutation of long-lived radionuclides.

Thus, German activities on these issues funded by the Federal Government solely

cover the participation in the further development of safety requirements and safety

technologies which are to consider new findings from operating experience and from

research, as well as the development of basic technologies for the minimisation of

radioactive waste (e. g. by transmutation).

German nuclear research and development also has a high level of competence in

these fields which is integrated in international (IAEA, OECD) and European (EU)

networks in order to further improve nuclear safety within and across national

boundaries also in future, and to develop scientifically substantiated repository

concepts for long-lived radioactive material in co-operation with other European

partners. An objective of the work is, among other things, the harmonisation of safety

requirements and safety criteria in Europe. In view of the reactors operated in the

candidate countries, this issue is of particular importance.

Scientists of German research institutions shall participate in international

developments also in future to examine whether these can be used for safety

improvements of German installations, and to contribute relevant safety knowledge

existing in Germany.

For young scientists, activities on innovative concepts are of special interest because

they include scientifically challenging topics that are dealt with in international co-

operation and integrated into international projects. They contribute to the urgently

necessary maintenance and further development of nuclear safety competence in

Germany.

4.6.1 Analysis of innovative international concepts

The aim of the work is, among other things, the assessment of different transmutation

concepts regarding technical feasibility, the safety characteristics and the achievable

transmutation rates. Neutron-physical analyses on the assessment of strategies for the

reduction of radiotoxicity of the waste by transmutation are performed, and safety-

74

related questions on transient and accident behaviour of transmutation facilities are

dealt with.

The work on critical molten-salt-fuelled reactors (MSR), representing a possible

alternative to accelerator-driven systems (ADS) and other concepts for the

transmutation of plutonium and other long-lived actinides, shall be pursued in

international projects of the European Union and the “International Scientific-Technical

Centre”.

The aim of the work on high-temperature reactors (HTR) is to maintain a minimum level

of competence for the observation, assessment and exertion of influence of the safety

philosophy on HTR projects, currently being performed in other countries. Scientific

contributions are made to use the inherent safety characteristics of the modular HTR

with support by the industry.

4.6.2 Avoidance and reduction of plutonium and minor actinides

In international co-operation, studies are performed to minimise radioactive waste

already during operation of nuclear reactors by the use of various fuels (e. g. thorium-

based systems and inert fuel matrices).

At the reactors operated today, the reduction of plutonium can be realised by the use of

plutonium MOX fuel based on thorium or on inert matrices much faster than it is

currently possible in conventional uranium-based reactors. In this respect, safety

analyses are performed for pressurised water reactors. The aim of the work is to

reduce the amount of plutonium, maintaining or even improving the safety-related

characteristics of the reactor. Alternative fuel strategies, as e. g. the combustion of

minor actinides (americium, curium, neptunium) in LWRs for the reduction of long-lived

radiotoxicity, shall be examined more detailed in future regarding their conversion

efficiency and their impacts on the fuel cycle (fuel production, reactor operation,

reprocessing, disposal of waste) in international co-operation.

75

4.6.3 Safety of subcritical systems for transmutation

The activities on subcritical systems for the reduction of radiotoxicity are integrated in

the 5th and the 6th Framework Programme of the EU and specified in close co-operation

with the European partners regarding the long-term strategy of these activities. They

have to be continued to evaluate and demonstrate the technical feasibility of

transmutation in the medium term and in international co-operation.

The work on concept development is oriented towards the superordinate strategy of the

“European Roadmap for Nuclear Waste Transmutation” of the Enlarged Technical

Working Group which analyses a nuclear fuel cycle under consideration of accelerator-

driven subcritical systems (ADS). The scientific-technical contributions to the analysis

and assessment are dealt with under the aspects of safety, technology, neutronics,

thermal hydraulics and material within the framework of various EU projects and

networks.

The spallation target is an essential component of an accelerator-driven system (ADS)

Within the framework of the international MEGAPIE (Megawatt Pilot Experiment)

initiative of the French Commissariat à l'Energie Atomique (CEA), the Swiss Paul

Scherrer Institute (PSI) and the Forschungszentrum Karlsruhe (FZK) an ADS-typical

lead-bismuth spallation target is being designed and constructed for the Proton

Accelerator of PSI. Commissioning is scheduled for 2005. Further partners are

research institutions from Belgium, Italy, Japan, Korea and the USA. The work is

supplemented by the EU project MEGAPIE-TEST.

5 International co-operation

German research activities performed within the framework of international co-

operation are presented in the respective subject chapters of this report. Emphasis is to

be laid on joint research work with scientists from Central and Eastern Europe and

Russia within the framework of bilateral contracts of the Federal Government or

German research institutions on scientific-technical co-operation.

The necessity of maintaining German competence with regard to nuclear safety issues

has been recognised by all those involved. In the past years, the funds for reactor

safety research have been reduced in many countries. As a consequence, test facilities

76

had to be decommissioned. In Germany, these were the HDR, BMC and UPTF

facilities. Due to the concern that the lack of important test facilities might lead to

scientific deficits in the safety assessments of nuclear installations, German

participation in research projects of the OECD/NEA plays an particular role. It promotes

the participation in the further development of international safety requirements in a

qualified manner, and efficient exertion of influence by Germany on the safety-related

enhancement of nuclear technology, and thus of the safety level of Germany’s

neighbouring countries.

The OECD projects have in common that there is an interest of many countries in the

solution of jointly identified safety-related issues and in the efficient shared use of the

experimental resources still available internationally by joint financing of the necessary

investigations.

The OECD projects render a scientific added value. Experts from different countries

jointly discuss the planning of the projects down to technical details, influence their

realisation according to the specified objectives and interpret the results with regard to

their safety-related impacts.

Nearly all of the OECD projects are agreed upon internationally, using a standard

contract form. This contract includes the technical proposal for the work programme,

the financial obligations of the participating countries, as well as the regulations for the

realisation of the project and the access to its results.

In general, the costs for the project are borne to about 50 % by the country providing

the experimental facility. The remaining costs are shared among the other participating

OECD countries. The respective shares are calculated according to the proportions of

the gross national products.

The financial participation of Germany is realised, in most cases, as projects within the

framework of project-funded reactor safety research of the BMWA.

Germany participates in all current projects of the OECD-NEA and intends also to

participate in further planned projects.

For many years, so-called international standard problems (ISPs) have been performed

on behalf and with the technical support of the OECD/CSNI (Organisation for Economic

77

Co-operation and Development / Committee on the Safety of Nuclear Installations). An

ISP comprises the comparison between analytical (calculated) and experimental

(measured and balanced) data. The experiments that serve as basis for ISPs relate to

the different fields of nuclear safety research. In general, they represent the respective

state of the art in science and technology with regard to configuration and performance

and, above all, instrumentation.

When calculating analytical data, computer codes shall be applied that are both used in

research and by the experts consulted by the licensing authority or the industry for

safety assessments and the design of nuclear power plants. Here, the best-estimate

calculations are to be performed and, preferably, without knowing the results of the

experiments.

Thus, international standard problems are suitable for validation (comparison between

calculation result and experiment) and verification (comparison between the calculation

results obtained with different models), for an in-depth discussion about the

phenomenological understanding and implementation in form of corresponding models,

as well as for the identification of deficits. For these reasons, many ISPs performed in

the past, today are an indispensable element of the basis validation of the relevant

computer codes.

In the past, Germany rendered considerable contributions to ISPs, be it by their

performance, i. e. provision of experiments as well as technical performance and

documentation of comparative activities, or by participation of various German

institutions. Especially in view of the increasingly limited experimental possibilities in

Germany, the necessity is emphasised to maintain this general attitude with regard to

ISPs also for the future.

Within the framework of Co-ordinated Research Programmes (CRPs) of the IAEA, two

programmes are in progress. The first programme concentrates on R&D activities for

the demonstration of the technological feasibility of transmutation of long-lived

radioactive waste and its impacts on the reduction of radiotoxicity. Methods are

developed for the description of the dynamics of the transmutation systems and the

future research demand is defined. The second programme deals with safety issues

and development of methods in connection with transmutation of plutonium in a

BN-600 reactor in Russia.

78

HALDEN

Participants: 13 countries

Performed by: Institute for Energy Technology, Kjeller and HALDEN, Norway

Duration: 01.01.2003 - 31.12.2005

Topics of research: In-pile fuel/materials investigation, man-machine interaction

MASCA (Material Scaling)


Performed by: Russian Research Centre Kurchatov Institute, Moscow

Duration: 01.07.2000 - 30.06.2003

Topic of research: Phenomenological studies on the behaviour of the reactor pressure

vessel during severe accidents under thermal loading induced by chemical reactions

and by fission products in the core melt

RASPLAV


Performed by: Russian Research Centre Kurchatov Institute, Moscow

Duration: 01.07.1997 - 30.06.2000

Topic of research: Experimental investigations of convection in a melt pool with

prototypical core melt behaviour to assess the possibility of core retention in the reactor

pressure vessel

79

OLHF (OECD Lower Head Failure)


Performed by: Sandia National Laboratory on behalf of the USNRC

Duration: 01.09.1998 - 31.08.2001

Topic of research: Experiments on the creep behaviour of models of the lower plenum

to assess potential failure modes of the lower RPV dome plate

SETH


Performed by: Framatome ANP (Germany) on PKL, Paul Scherrer Institute

(Switzerland) on PANDA

Duration: 01.04.2001 - 30.06.2005

Topic of research: Experimental tests in the primary circuit and in the containment for

the prevention or control of accidents at the test facilities PKL (Germany) and PANDA

(Switzerland)

MCCI (Melt Coolability – Concrete Interaction)


Performed by: Argonne National Laboratory on behalf of the USNRC

Duration: 01.01.2002 - 31.12.2005

Topic of research: Experimental investigation of cooling mechanisms for the control of

the core melt spreading in the containment and on long-term 2D molten core/concrete

interaction

80

CABRI-WL

Participants: currently 13 countries

Performed by: Institut de Radioprotection et de Sûreté Nucléaire (IRSN)

Duration: 2001 - 2008

Topic of research: Tests at the CABRI research reactor on the behaviour of high burn-

up and MOX fuel under the conditions of reactivity-initiated accidents

81

Abbreviations

ADS Accelerator Driven System

AM Accident Management

AMES Ageing Materials Evaluation and Studies

ARTIST Aerosol Trapping in Steam Generator

ASTAR Advanced 3D Two-Phase Flow Simulation Tool for Application to Reactor Safety

ASTEC Accident Source Term Evaluation Code

ATHLET Analyse der Thermohydraulik von Lecks und Transienten - Analysis of the Thermal Hydraulics of Leaks and Transients

ATHLET-CD Analysis of the Thermal Hydraulics of Leaks and Transients - Core Degradation

BGR Bundesanstalt für Geowissenschaften und Rohstoffe - Federal Institute for Geosciences and Natural Resources

BMBF Bundesministerium für Bildung und Forschung - Federal Ministry for Education and Research

BMC Battelle Model Containment

BMF Bundesministerium für Finanzen - Federal Ministry of Finance

BMU Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit - Ministry for the Environment, Nature Conservation and Nuclear Safety

BMWA Bundesministerium für Wirtschaft und Arbeit - Federal Ministry of Economics and Labour

BWR Boiling Water Reactor

CABRI Reactivity-insertion-accident (RIA) Test Reactor of CEA at Cadarache (France)

82

CAPRA Consommation Accrue de Plutonium dans les Réacteurs Rapides - Increased Plutonium Consumption in Fast Reactors

CATHARE French Thermal-hydraulics Code (CEA)

CEA Commissariat à l’Énergie Atomique - French Atomic Energy Commission

CFD Computational Fluid Dynamics

COBRA Model for Thermal Hydraulics Inside a Fuel Element

COCOSYS Containment Code System

COMAS Corium on Material Surfaces

COMET Experiments on Analyses of Core Melt Accidents

CORESA Corium on Refractory and Sacrificial Materials

CSNI Committee on the Safety of Nuclear Installations

CUBBOX Approximation by Cubic Local Polynomials

DEMONA Demonstration nuklearen Aerosolverhaltens - Aerosol Experiments

DCH Direct Containment Heating

DWD Deutscher Wetterdienst - German National Meteorological Service

DYN3D 3-Dimensional Reactor Dynamics Program

EVK Evaluierungskommission - Evaluation Commission

EOL End of Life

83

FH Fachhochschule - University of Applied Sciences

FIPLOC Fission Product Localisation (Computer Program)

FOREVER Failure of Reactor Vessel Retention

FZJ Forschungszentrum Jülich - Research Centre Jülich

FZK Forschungszentrum Karlsruhe - Research Centre Karlsruhe

FZR Forschungszentrum Rossendorf - Rossendorf Research Centre

GASFLOW 3 D CFD-Code

GRS Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH

HTR High Temperature Reactor

IAEA International Atomic Energy Agency

I&C Instrumentation and Control

ICARE French Core-melt System Code

IKE Institut für Kernenergetik und Energiesysteme - Institute for Nuclear Technology and Energy Systems

IRSN Institute de Radioprotection et de Sûreté Nucléaire (France)

ISB Russian Thermal-hydraulic Facility; Small LOCAs for VVER-440

84

ISP International Standard Problem

ISTC International Science & Technology Center

IZFP Institut für zerstörungsfreie Prüfverfahren der Fraunhofer Gesellschaft - Institute for Nondestructive Testing of the Fraunhofer Society

KAEVER Kernschmelzaerosolverhalten - Core-Melt/Aerosol Behaviour

KATS Karlsruher Thermit-Schmelzen (Experimente zur Ausbreitung und Kühlung der Kernschmelzen - Karlsruhe Thermite Melts (Experiments on the Spreading and Cooling of Core Melt)

KESS Kernschmelz-Systemcode - Core-melt System Code

KV Kompetenzverbund -Competence Pool

LAVA Core-melt Spreading Program

LOCA Loss of Coolant Accident

LWR Light-water Reactor

MARCH US-American Thermal-hydraulic Module of the Source Term Code Package

MASCA Material Scaling

MC3D Model on Melt-Coolant Interaction (CEA Grenoble)

MEGAPIE MEGAwatt Pilot Experiment (Pilot Spallation Target)

MELCOR Integral Severe Accident Analysis Code (NRC)

85

MOX Mixed-oxide Fuel (Uranium/Plutonium or Thorium/Plutonium)

MPA Staatliche Materialprüfungsanstalt, Universität Stuttgart State Materials Testing Institute, Stuttgart University

MSR Molten-salt Reactor

NEA Nuclear Energy Agency (of the OECD)

OECD Organisation for Economic Co-operation and Development

PANBOX Neutron-kinetics Code (Framatome ANP)

PANDA Passive Residual Heat Removal and Depressurisation Test Facility (Switzerland)

PHEBUS Facility for Core Analyses at Cadarache (France)

PKL Primärkreis Loop - Primary Circuit Loop (Thermal-hydraulic 4-Loop Test Facility)

PSA Probabilistic Safety Analysis

PSB Russian Thermal-hydraulic Facility for VVER-1000

PT R Projektträger Reaktorsicherheitsforschung (GRS) - Project Sponsor Reactor Safety Research (GRS)

PT WT E Projektträger für Wassertechnologie und Entsorgung - Project Sponsor for Water Technology and Waste Management

PTS Pressurised Thermal Shock

PWR Pressurised Water Reactor

86

QUABOX Approximation by Quadratic Local Polynomials

QUENCH Experiments on the Quench Behaviour of Fuel Rod Bundles

RALOC Radiolysis and Local Concentration (Containment Code)

RBMK Reaktor Bolschoi Moschtschnosti Kanalny - Soviet-designed Graphite-Moderated Pressure Tube Reactor

RECOM Code for the Simulation of Radionuclide Resuspension at Bubbling Water Pool Surfaces

RELAP American Accident Analysis Code

RIA Reactivity-initiated Accident

ROCOM Rossendorf Coolant Mixing Model

RODOS/RESY Real-time Online Decision Support/Computer-based Decision Support System

RPV Reactor Pressure Vessel

RUB Ruhr-Universität Bochum

SCDAP Severe Core Damage Analysis Package, American Core-melt System Code (NRC)

SUSA Software System for Uncertainty and Sensitivity Analysis

ThAI Thermal-hydraulics-Aerosol-Iodine (Experimental Facility)

TOPFLOW Transient Two Phase Flow Test Facility

TUM Technische Universität München

87

UPTF Upper Plenum Test Facility

UPTF-TRAM UPTF-Transient and Accident Management

VANAM Experiments on the Depletion Behaviour of Nuclear Aerosols in a Multi-compartment Containment Geometry

VVER Russian Pressurised Water Reactor

WTZ Scientific-Technical Co-operation (with Central- and Eastern European Countries)

88

Literature /EVK 00/ Nuclear Reactor Safety and Repository Research in Germany;

Report of the Working Group Convened by The Federal Ministry of Economics and Technology (BMWi) (Evaluation Commission);21 January 2000

/AtG 02/ Gesetz über die friedliche Verwendung der Kernenergie und den Schutz gegen ihre Gefahren (Atomgesetz - AtG) vom 23. Dezember 1959, Neufassung vom 15. Juli 1985, letzte Änderung durch Gesetz vom 22. April 2002 (BGBl. I 2002, Nr. 26)

/KVK 01/ Vorauseilender Ausstieg aus der Kernenergie an den deutschen Hochschulen und Forschungszentren; P. Fritz et. al., atw Jg. (2002) Heft 2, Februar 2001, Seite 88 ff.

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