PROGRAMME ON NUCLEAR POWER PLANT SAFETY, ANNUAL PLAN 2003

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VTT PROCESSES

National research programme

Safety case and licensing evidence

Utility research

Regulatory inspections

and reviews

Regulatory research

SAFIR

2003-2006 Safety of nuclear power plants – Finnish national research programme

SAFIR, THE FINNISH RESEARCH

PROGRAMME ON NUCLEAR POWER PLANT SAFETY, ANNUAL PLAN 2003

Author: Eija Karita Puska

Publicity: Public

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The use of the name of the Technical Research Centre of Finland (VTT) in advertising or publication in part of this report is only permissible by written authorisation from the Technical Research Centre of Finland

Project manager Contact person

Eija Karita Puska Eija Karita Puska

[email protected]

Diary code (VTT) Order reference

KTM

Project title and reference code Report identification & Pages Date

12SAHA PRO1/P7007/03

22 p. + App.

15.5.2003

Report title and author(s)

SAFIR, The Finnish Research Programme on Nuclear Power Plant Safety, Annual Plan 2003 Puska, E. K.

Summary

SAFIR 2003 – 2006 is the Finnish public research programme on nuclear power plant safety, launched and administrated by the Ministry of Trade and Industry (KTM).

The steering group of SAFIR was nominated by KTM. VTT Processes was elected as the coordination unit of SAFIR for the period 2002-2004 on the basis of an open call for proposals. The projects for the SAFIR programme for the year 2003 were also selected on the basis of an open call for proposals. The main funding sources of the programme are KTM, Technical Research Centre of Finland (VTT), Radiation and Nuclear Safety Authority (STUK), and the power companies Teollisuuden Voima Oy (TVO) and Fortum Oyj. These parties are also represented in the steering group of the programme along with the representatives of TEKES and Helsinki and Lappeenranta Universities of Technology.

SAFIR research programme consist currently of 19 research projects in the six research areas. The volume of the projects in 2003 varies from some person months up to several person years, and the planned total duration from one to four years. In 2003 the planned volume of the programme is 31 person years and 4 million €.

The research areas and focus of the programme have been defined in the proposal for the content and organisation of the new research programme. These six key research areas are reactor fuel and core, reactor circuit and structural safety, containment and process safety functions, automation, control room and IT, organisations and safety management and risk-informed safety management.

The execution of the actual programme is based on this annual plan 2003 prepared by the programme staff in consultation with the main funding partners of the programme and accepted by the steering group.

Distribution Publicity

VTT Processes, Library, P.O.Box 1604, FIN-02044 VTT, Finland Phone: +358-9-4565004, Telefax: +358-9-4565000

Public

Project manager Reviewed and approved by

Eija Karita Puska

Senior Research Scientist Risto Sairanen Group Manager

Seppo Vuori Research Manager

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MTI T. Turunen J. Manninen J. Aurela A. Väätäinen TEKES M. Äijälä R. Munther STUK J. Laaksonen H. Koponen L. Reiman S. Suksi T. Okkonen S. Salomaa M. Ojanen K. Valtonen O. Vilkamo J. Marttila R. Virolainen J. Hyvärinen H. Heimbürger P. Salminen M-L. Järvinen J. Hinttala R. Rantala H. Takala R. Keskinen O. Valkeajärvi H. Ollikkala I. Niemelä S. Sinkkonen P. Tiippana S. Koivula K. Åstrand S. Metso HUT H. Kalli R. Salomaa M. Määttänen H. Hänninen LUT

L. Koskelainen J. Vihavainen H. Purhonen M. Puustinen TUT

P. Lautala

FORTUM A. Vuorenmaa H. Tuomisto J. Kohopää P. Lundström P. Gango U. Lindén M. Tiitinen J. Snellman J. Helske H. Raumolin A. Tamminen P. Siltanen A. Norta M. Winter E. Rinttilä M. Komsi J. Vaurio R. Teräsvirta M. Antila S. Savolainen K. Jänkälä O. Hietanen T. Buddas A. Neuvonen R. Paussu O. Kymäläinen A. Felin P. Kettunen E. Markkanen R. Korhonen M. Kattainen M. Välisuo P. Luukkanen TVO

M. Paavola R. Mokka A. Rastas E. Patrakka M. Mustonen M. Friberg I. Mikkola K. Ranta-Puska M. Solala S. Koski K. Taivainen L.-E. Häll R. Himanen J. Hakala E. Mannola H. Sjövall A. Reinvall

P. Smeekes M. Malinen E. Muttilainen K. Hukkanen M. Yli-Kauhaluoma J. Pesonen

VTT J. Forstén VTT Processes M. Kara L. Mattila S. Vuori P. Koskivirta T. Vanttola S. Kelppe J. Tuunanen R. Sairanen I. Lindholm O. Tiihonen M. Rajamäki M. Anttila M. Hänninen M. Manninen H. Raiko H. Räty A. Daavittila A. Tanskanen R. Huhtanen A. Tiitta

VTT Industrial Systems

R. Rintamaa J. Solin P. Aaltonen T. Laitinen P. Karjalainen- Roikonen M. Valo

L. Hansson-Lyyra P. Auerkari H. Talja

A. Saarenheimo H. Keinänen M. Sarkimo P. Kauppinen M. Vilpas J. Suokas R. Heinonen B. Wahlström P. Haapanen

L. Norros U. Pulkkinen O. Ventä K. Simola K. Ruuhilehto M. Nuutinen A. Helminen T. Reiman VTT Build. &

Transp M. Kokkala E. Mikkola

O. Keski-Rahkonen K. Orantie

SKI

C. Viktorsson G. Löwenhielm L. Gunsell R. Nyman O. Sandervåg L. Pettil

YEN/YTN/Other S. Rouhinen U. Sirkeinen U. Lähteenmäki O. Pahkala M. von Bonsdorff P. Jauho

A. Vuorinen M. Vuorio P. Kallioniemi A. Leppänen L. Nurmi T. Bennerstedt (NKS)

G.Van Goethem (EU)

I. Huhtiniemi (JRC)

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PREFACE

SAFIR, SAFfety of Nuclear Power Plants - Finnish National Research Programme, 2003- 2006, is the newest link in the chain of Finnish national research programmes in nuclear.

Organisation of public nuclear energy research as national research programmes was started in 1989, launched by the Ministry of Trade and Industry (KTM). Since then national programmes have been carried out in the fields of operational aspects of safety (YKÄ 1990-1994, RETU 1995-1998), structural safety (RATU 1990-1994, RATU2 1995- 1998), FINNUS 1999-2002 combining the operational aspects and nuclear safety and nuclear waste management (JYT 1989-1993, JYT2 1994-1996, JYT2001 1997-2001).

In parallel there has been the Finnish Fusion Research Programme (FFUSION) 1993-2002, programmes on Advanced Light Water Reactor concepts (ALWR) 1998-2002 and a project on component life management 1999-2003, partly funded by the National Technology Agency (TEKES). Currently fusion research continues in the FFUSION2 and nuclear waste management research in the KYT programme.

KTM decided to continue the national research efforts on fission reactor safety in a single research programme after completion of FINNUS. The national advisory committee on nuclear energy, commissioned by KTM, gave the mandate of planning the new programme to the FINNUS steering group that nominated the SAFIR planning group to make the proposal for the content and organisation of the new research programme [1].

The steering group of SAFIR was nominated by KTM. VTT Processes was elected as the coordination unit of SAFIR for the period 2002-2004 on the basis of an open call for proposals. The projects for the SAFIR programme for the year 2003 were also selected on the basis of an open call for proposals. The main funding sources of the programme are KTM, Technical Research Centre of Finland (VTT), Radiation and Nuclear Safety Authority (STUK), and the power companies Teollisuuden Voima Oy (TVO) and Fortum Oyj. These parties are also represented in the steering group of the programme along with the representatives of TEKES and Helsinki and Lappeenranta Universities of Technology.

SAFIR research programme consist currently of 19 research projects in the six research areas. The volume of the projects in 2003 varies from some person months up to several person years, and the planned total duration from one to four years. In 2003 the planned volume of the programme is 31 person years and 4 million €.

The research areas and focus of the programme have been defined in the proposal for the content and organisation of the new research programme [1]. The execution of the actual programme is based on this annual plan 2003 prepared by the programme staff in consultation with the main funding partners of the programme and accepted by the steering group.

This report has been prepared by the programme leader in cooperation with the project leaders and members of the programme staff.

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1 INTRODUCTION

A country utilising nuclear energy is presumed to possess a sufficient infrastructure to cover the education and research in this field, besides the operating and supervisory organisations of the plants. The starting point of public nuclear safety research programmes is that they provide the necessary conditions for retaining the knowledge needed for ensuring the continuance of safe and economic use of nuclear power, for development of new know-how and for participation in international cooperation. In fact, the Finnish organisations engaged in research in this sector have been an important resource, which the various ministries, the Radiation and Nuclear Safety Authority (STUK) and the power companies have had at their disposal [1]. SAFIR, Safety of Nuclear Power Plants - Finnish National Research Programme, 2003- 2006, is the newest link in the chain of Finnish national research programmes in nuclear energy research.

The plan [1] has been made for the period 2003–2006, but it is based on safety challenges identified for a longer time span as well, as indicated in Figure 1. The favourable decision- in-principle on a new nuclear power plant unit adopted by Parliament in May 2002 has also been taken into account in the plan. The safety challenges set by the existing plants and the new plant unit, as well as the ensuing research needs do, however, converge to a great extent.

Figure 1. Main themes of nuclear plant safety [1].

The construction of the new power plant unit will increase the need for experts in the field in Finland. At the same time, the retirement of the existing experts is continuing. These factors together will call for more education and training, in which active research

Main themes

physical barriers plant functions initiating events

how the risks are managed?

how the plant is operated?

safety management

what the plant sustains?

how the plant functions?

design

& analysis

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activities play a key role. This situation also makes long-term safety research face a great challenge.

The general plan [1] defines the important research needs related to the safety challenges, such as the ageing of the existing plants, technical reforms in the various areas of technology and organisational changes. The research into these needs is the programme’s main techno-scientific task. In addition, the programme has to ensure the maintenance of know-how in those areas where no significant changes occur but in which dynamic research activities are the absolute precondition for safe use of nuclear power.

Figure 2. Future safety challenges to be addressed at in the SAFIR programme [1].

Although SAFIR builds strongly on the FINNUS [2,3] and all the other preceeding national research programmes on nuclear safety, new areas and flexibility is sought, too, within the limits of available funding. Thus, rapid specified projects, projects running throughout the entire research programme and development work of a very long duration all suit to the SAFIR in a flexible manner.

In addition to conducting the actual research according to the yearly plans, SAFIR will function as an efficient conveyor of information to all organisations operating in the nuclear energy sector and as an open discussion forum for participation in international projects, allocation of resources and in planning of new projects. For this purpose another forum, the forum of projects administrated and performed outside SAFIR, but to be reported and discussed within SAFIR, has been established.

Future safety challenges

physical barriers plant functions

initiating events

safety management design & analysis

integrity function

operation risks

1. New fuel designs and enhanced use 2. Ensurance of integrity of

an ageing reactor circuit 3. Ensurance of containment

integrity and leak-tightness

8. Operational development with modern technology 9. Plant lifetime management 10. Development of organisational

culture and safety management 12. Risk-informed safety

and operational management 11. Risk analysis of external effects

6. Automation modernizations 7. Control room modernizations

4. New types of nuclear power plants 5. Uncertainties associated with process safety functions

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2 PROGRAMME CONTENTS AND STRUCTURE

The SAFIR programme has been divided into six research areas:

1. Reactor fuel and core

2. Reactor circuit and structural safety 3. Containment and process safety functions

4. Automation, control room and information technology 5. Organisations and safety management

6. Risk-informed safety management

Figure 3 illustrates the division of the twelve major future safety challenges into these six research areas of SAFIR. Figure 3 indicates also the recognised fact that many of the safety challenges have in fact overlap over the various research areas.

Figure 3. Division of the twelve major future safety challenges into the six research areas of SAFIR [1].

There are currently 19 research projects and the administration project going on during the year 2003 in the programme. The titles of the projects and their division into the six research areas have been illustrated in Table 1. The extent of the projects vary from a few man months into several man years. Most of the projects have been planned to continue throughout the entire four-year span of the SAFIR programme. Detailed research plans of the projects have been included in Appendix 1 and corresponding tables on expenses and financing in Appendix 2.

Grouping of challenges

1 2

3 5 6 4

1. New fuel designs and enhanced use

9. Plant lifetime management 2. Ensurance of integrity of an ageing reactor circuit

3. Ensurance of containment integrity and leak-tightness 5. Uncertainties associated with process safety functions 8. Operational development

with modern technology 6. Automation modernizations 7. Control room modernizations

12. Risk-informed safety and operational management 11. Risk analysis

of external effects 10. Development of organisational

culture and safety management 4. New types of

nuclear power plants

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Table 1. The research projects of SAFIR in 2003.

Group Project Acronym Funding

k€ Volume

person months 1.

Enhanced methods for reactor analysis EMERALD 525 48

High-burnup upgrades in fuel behaviour modelling KORU 210 20,5 2.

Integrity and life time of reactor circuits INTELI 1057 77,9 Ageing of the function of the containment building AGCONT 13 1,1 Participation in the OECD NEA task group concrete ageing CONAGE 9,48 0,6 Concrete technological studies related to the construction,

inspection and reparation of the nuclear power plant structures

CONTECH 100,6 8

3.

Wall response to soft impact WARSI 137,7 13

The integral code for design basis accident analyses TIFANY 206,8 17 Thermal hydraulic analysis of nuclear reactors THEA 187 12 Severe accidents and nuclear containment integrity SANCY 315,6 17 Fission product gas and aerosol particle control FIKSU 98,2 10,5

Emergency preparedness supporting studies OTUS 50 4,5

Archiving experiment data KOETAR 60 6

Condensation pool experiments POOLEX 152,5 15

4.

Interaction approach to development of control rooms IDEC 140 12 Application possibilities of systematic requirements

management in the improvement of nuclear safety in Finland APSREM 50 3,8 5.

Organisational culture and management of change CULMA 205,7 17 6.

Potential of fire spread POTFIS 158 12

Principles and practices of risk-informed safety management PPRISMA 258,55 24,4

0. SAFIR Administration and information (2002-2003) SAHA 116,7 9

Total 4051,5 329,3

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2.1 REACTOR CORE AND FUEL

The area covers reactor physics, reactor dynamics and fuel behaviour analysis. The research is done solely with the help of calculational tools, partly with sophisticated tools developed at VTT and partly using tools developed elsewhere. The area has living contact to the experimental work via international connections, such as the OECD Halden Reactor Project.

In 2003 there are two projects in this area, the Enhanced methods for reactor analysis (EMERALD) dealing with reactor physics and dynamics and High-burnup upgrades in fuel behaviour modelling (KORU) dealing with the fuel research. In both projects, education of the new generation has an essential role. Figure 4 illustrates the research field.

Figure 4. Research themes in reactor fuel and core area[1].

2.1.1 Enhanced methods for reactor analysis (EMERALD)

New fuel and reactor designs, new loading strategies and the continuing trend towards higher fuel burnups make it necessary to further improve as well as validate VTT Processes´ code system for reactor analysis into a really unified, up-to-date and flexible entirety consisting of both programs acquired from elsewhere and programs that are the result of own development. It should be possible to follow the whole life cycle of the nuclear fuel from a reactor physics point of view until its final disposal. In today´s situation, when the use of nuclear power is increased at the same time as the nuclear experts of the present generation are gradually retiring from work, it is of special importance to maintain competence by training new personnel and documenting the methods and results carefully. Research in reactor physics and dynamics is very much dependent on international co-operation, where

1. Reactor fuel and core

CRITICALITY SAFETY

FUEL

ICFM = IN-CORE FUEL MANAGEMENT

ICFM CORE

SEVERE ACCIDENTS

ACCIDENTS

TRANSIENTS

COOLANT FLOW STABILITY

OPERATIONAL SAFETY LIMITS

IRRADIATION:

MATERIAL DAMAGES

& ACTIVATION

HIGH

BURNUP CLADDING

INTEGRITY CORROSION WATER

CHEMISTRY TRANSPORTATION

STORAGE FINAL

REFUELING SHUTDOWNS SPENT FUEL

REPOSITORY STORAGE

THERMAL MECHANICS

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the results of the work are presented e.g. in the form of benchmark studies and other comparisons of measurements and computational results, at conferences and in international publications. The project is planned to continue during the four years of the SAFIR program with new or updated goals to be specified on a year-by-year basis.

2.1.2 High-burnup upgrades in fuel behaviour modelling (KORU)

A national capability of making independent fuel performance evaluations will be preserved via this project. The modelling in fuel behaviour codes in use at VTT will be upgraded to meet the requirements that result from evolving fuel design and operational data – notably higher burnup goals – and from revised guidelines for applying the licensing procedures.

Emphasis will then be on the performance of fuel in postulated accident conditions, where uncertainties are being reduced by leaning on results from ongoing international experiments, and with parallel model development and validation. For steady-state conditions, the well-established codes need partial renovation, too, in particular as regards the descriptions of fission product swelling and release and detailed mechanical response.

The future licensing evaluations may not be conclusive without application of probabilistic methods that will be taken in use in transient fuel behaviour codes. Education and training of the next generation of experts is another big challenge in the project.

2.2 REACTOR CIRCUIT AND STRUCTURAL SAFETY

The area covers the studies on the integrity and life time of the entire reactor circuit and the studies of containment building construction, inspection, ageing and repairing. The reactor circuit studies are all included in one very large project, Integrity and life time of reactor circuits (INTELI), and the containment is studied in three separate projects, Ageing of the Function of the Containment Building (AGCONT), Participation in the OECD NEA Task Group Concrete Ageing (CONAGE), Concrete Technological Studies Related to the Construction, Inspection and Reparation of the Nuclear Power Plant Structures (CONTECH). In this area, the work done outside SAFIR, both in Finland and in several EU- projects will be reported in some extent. Figure 5 illustrates the research area.

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Figure 5. Research themes in Reactor circuit and structural safety area [1].

2.2.1 Integrity and life time of reactor circuits (INTELI)

The main objective of the project is to assure the structural integrity of the main components of the reactor circuit of the nuclear power plant and to study the typical ageing mechanisms affecting the integrity of main components during the life-time of the reactor. The main components included in the scope of the project are:

− Reactor pressure vessel with nozzles and internals

− Piping of reactor circuit

− Other components (steam generators, pumps, valves, pressurizer, heat-exchangers) The overall objectives of the research work related to these components are following:

To understand and model the ageing mechanisms of the reactor pressure vessel including safe-end nozzles and internals. The target is predicting the development of the ageing and estimate the effects of ageing and the need for corrective actions and possible repairs.

To develop improved methodology for the assessment of the embrittelement of reactor pressure vessel. To verify the transferability of the material data based on the Master Curve to the structural analysis of real components.

Requirements – design – manufacturing Inspection – characterisation – predictions

Reactor circuit pipelines Loadings Material properties Geometry

Environment Cumulative effects Operating experiences LBB

Water chemistry effects

Base material Oxide film Flow conditions Impurities Reactor pressure vessel with joints and internals Re-embrittlement

Loadings

Data

Material properties

Integrity Criteria Methods

Faults

Risk-informed safety management

Plant lifetime management Reactor circuit integrity

Operational development with modern technology

Other reactor circuit components

Loadings Materials Failure types Faults Criteria

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To improve the reliability of nondestructive evaluation methods used to detect, characterise and monitor defects in different areas of RPV.

To develop reliable methods for the assessment of bimetallic welds of nozzles and their loading conditions. Especially, the assessment methods of residual stresses and thermal loads will be improved.

For the reliable and quick analysis of defects and damages new methodologies and tools will be developed based on multitechnical analysis software and expert networks.

To develop methods for the measurement and prediction of material properties of reactor internals during service. The key-elements of this work will be the modelling of failure mechanisms, identification of loads affecting the structures and the technology related to the application of miniature test samples.

To apply risk-informed methods to the life-time management of piping. Risk-informed methods will be used to optimise the inspection practises applied to piping.

To develop and take in use new, more realistic material models basing on more accurate material data and improved understanding of phenomena affecting the material. Material models and realistic modelling of residual stresses and loads are necessary for the numerical simulation of the behaviour of piping.

To create model describing and predicting the transfer of active particles in primary water.

By modelling the oxide film the effects of material properties, water chemistry, stresses and flow conditions on the properties of the oxide film can be described.

2.2.2 Ageing of the function of the containment building (AGCONT)

The goal of the project is to ensure the safety of the containment building and other massive concrete structures such as sea water tunnels and the substructures of the turbines both in new and in ageing power plants.

The result will be a State of the Art -knowledge of the threatening factors involved in ageing of the containment building and the behaviour of essential materials in accidents of the ageing power plants.

2.2.3 Participation in the OECD NEA task group concrete ageing (CONAGE)

This small project covers participation in The OECD NEA Task Group on Concrete Ageing that reports the ageing effects and the determination, monitoring and reparation methods of ageing effects on concrete structures (concrete, nonprestressed reinforcement, prestressed reinforcement, steel sealing plates etc.) as well as the effects on service life.

2.2.4 Concrete technological studies related to the construction, inspection and reparation of the nuclear power plant structures (CONTECH)

The goal of the project is to gain knowledge of the structural and durability behaviour of both prestressed and nonprestressed concrete structures. The results of the project will be

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2.3 CONTAINMENT AND PROCESS SAFETY FUNCTIONS

The area covers simulation of nuclear power plant processess, calculational and thermal hydraulics using both CFD-tools and APROS-code, experimental thermal hydraulics at Lappeenranta University of Technology (LUT), severe accidents studies, where both experimental and calculational work is included and emergency preparedness studies.

Altogether 8 projects have been included into this area. They are the Wall response to soft impact (WARSI), The Integral Code for Design Basis Accident Analyses (TIFANY), Thermal Hydraulic Analysis of Nuclear Reactors (THEA), Severe Accidents and Nuclear Containment integrity (SANCY), Fission product gas and aerosol particle control (FIKSU), Emergency preparedness supporting studies (OTUS), Archiving experiment data (KOETAR) and Condensation pool experiments (POOLEX). In this field, training of new personnel as well as information on international research programmes have vital roles, too. Figure 6 illustrates the research field.

Figure 6. Research themes in containment and process safety functions area [1].

2.3.1 Wall response to soft impact (WARSI)

The main aim of the project is to develop and take in use methods for predicting response of reinforced concrete structures to impacts of deformable projectile that may contain combustible liquid ("fuel"). Loading, structural behaviour, like collapsing mechanism and the damage grade, will be predicted by simple hand calculations and using non-linear FE- method. Non-linear material parameters needed for these analyses will be studied. Also methods to evaluate the spread (e.g. release, penetration and dispersion) of fuel will be

3. Containment and process safety functions

External threats

Potential releases, environmental effects Containment

Process safety functions

Passive systems

Balanced safety

Integrated assessment of containment and process safety functions

Severe accidents

shutdown conditions

coolability long-term aspects

plant-specific questions chemistry fission products

Effects of ageing on containment function Integrity Leak-tightness

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studied. Suitability of selected simple empirical and semi-empirical based methods of predicting deflagration overpressure is assessed.

2.3.2 The integral code for design basis accident analyses (TIFANY)

The objective of the project is to create and validate a generic integrated APROS calculation model for design basis accident (DBA) analyses. Integrated in this context means that the calculation model includes both the primary circuit and the containment of the generic nuclear power plant without any restrictive boundary conditions to control the thermal hydraulics in the interface.

The resulting extension of the APROS simulation program will have generic models for typical boiling water reactor (BWR) and pressurized water reactor (PWR) containment. With this approach, it should be easy and fast for the end user to adjust the generic model to the specific geometry and configuration in interest.

The generic models will be validated against experimental results. The project also includes checking and correcting of the ranges of the correlations used in APROS to cover the needs of the DBA analyses.

2.3.3 Thermal hydraulic analysis of nuclear reactors (THEA)

The objectives of THEA project are to support experimental work in Lappeenranta University of Technology, to develop new or improved methods for thermal hydraulic analysis, especially for calculating multidimensional two-phase flows, and to increase the knowledge and understanding of reactor thermal hydraulics.

The major tools used in the project are APROS simulation programme and the CFD-code Fluent 6. The participation into international research programmes OECD/SETH, OECD/PSB-VVER and USNRC/CAMP is also included into this project.

2.3.4 Severe accidents and nuclear containment integritY (SANCY)

The objective of the SANCY-project is to reduce the remaining uncertainties of severe accident phenomena that are important to Finnish nuclear power plants. These issues comprise the core melt coolability, severe accident management in the long time range, severe accident phenomena under plant shutdown conditions, plant specific in-vessel melt progression issues, in particular the behavior of lower core support structure and melt discharge rate from the pressure vessel. A general review of remaining severe accident uncertainties will be made, with the specific goal to explore if any issues important for Finnish nuclear power plants have been left out in the past and on-going research.

Furthermore, the follow-up and participation of major international research projects in the area of severe accidents will carried out as part of the project. The project is planned to run for three years (2003-2005).

2.3.5 Fission product gas and aerosol particle control (FIKSU)

Phebus FP is a large scale facility, where phenomena taking place in a severe nuclear accident can be experimentally studied in realistic conditions. In this project the work done at the Phebus FP program is followed up. VTT will also continue to contribute into the

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Ruthenium chemistry is studied in oxidising atmosphere. The conditions will represent accidents during the plant shutdown. The objective is to determine, whether gaseous ruthenium could be released into the containment building.

2.3.6 Emergency preparedness supporting studies (OTUS)

The objective of the OTUS project is to improve emergency preparedness at the Finnish nuclear power plants by 1) studying radiation levels in rooms and in the vicinity of the power plants and accessibility of them in case of a severe accident during maintenance and refuelling outage, and by 2) studying the effect of land and sea breeze on the dispersion of atmospheric discharges of the power plants during an accident and methods to assess the dispersion during such weather.

In case of a severe accident during maintenance and refuelling outage radioactive substances can be transported into different parts of a nuclear power plant and its environment because the lid of a reactor pressure vessel and different material transfer gates of a reactor containment building may be open. Radiation levels in the rooms of the Finnish nuclear power plants and their vicinity have not been studied earlier in such accidents. Results from the PSA2 study of the refuelling and maintenance outage of the Olkiluoto power plant are available for the assessment of radiation levels. The corresponding PSA2 study of the Loviisa power plant has not yet been finished but similar assessment is expected to be possible.

Each of the Finnish nuclear power plants has a meteorological tower with temperature and wind measuring instruments. The power plants are prepared to make dispersion assessments of atmospheric discharges based on simple Gaussian dispersion models during an accident.

The event of land and sea breeze has not been taken into account in the procedures for the assessment of dispersion. Land and sea breeze in Finland has been studied at Finnish Meteorological Institute (IL) and University of Helsinki (HY).

2.3.7 Archiving experiment data (KOETAR)

The objective of the KOETAR project is to save, check, and archive data and documents of the thermal-hydraulic experiments performed at Lappeenranta University of Technology during the past 28 years. Most of the data and the documents are on media that are not compatible anymore with the hardware and software in use today.

The work will be done following the plan written and approved in the previous FINNUS research program. The checked data and documents will be archived in the STRESA database maintained by the Nuclear Safety Research Unit at Lappeenranta University of Technology and in CD-ROM disks.

2.3.8 Condensation pool experiments (POOLEX)

The main goal of the project is to increase the understanding of the phenomena in the condensation pool during steam injection. These phenomena could be connected to bubble dynamics (bubble growth, upward acceleration, detachment, breakup), pool swell, pressure oscillations and distribution, condensation rate (magnitude determined by means of visual bubble-volume estimates) and vibrations of the vent pipe. To achieve this understanding

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the needs for instrumentation and visualization has to be defined. For example, to estimate the loads on the pool structures by condensation pressure oscillations the frequency and the amplitude of the oscillations have to be measured. However, measuring and visualizing all the above-mentioned phenomena requires sophisticated measuring techniques. The needs and the resources for these techniques will be reviewed in the early phase of the project.

The final result of the project will be the database, which can be used for testing and developing the computing methods used for nuclear safety analysis.

2.4 AUTOMATION, CONTROL ROOM AND INFORMATION TECHNOLOGY

The area includes currently two research projects: Interaction approach to development of control rooms (IDEC) and Application Possibilities of Systematic Requirements Management in the Improvement of Nuclear Safety in Finland (APSReM). Additionally, the work done beyond SAFIR in Finland in the projects related to the renewal of existing control rooms and in connection of the new unit as well as work in some international projects will be discussed and reported in suitable extent. Figure 7 illustrates the research field.

Figure 7. Research themes in automation, control room and IT area [1].

2.4.1 Interaction approach to development of control rooms (IDEC)

The project aims at formulating a scientifically founded approach to the development of human-technology interaction (HTI) in control rooms and operating centres of complex industrial systems. A set of criteria for the validation of the design process and its results will be created. An interaction perspective, the so-called ecological approach is adopted in the construction of the metric. We argue that artefacts should not be evaluated independently of the practices of their use. Thus, in connection with defining criteria for a good control room

4. Automation, control room and information technology

Automation

Requirements engineering

Qualification process

Design evaluation

Lifetime management

Control room

Changes in control room work Verification and validation process Human system interaction (interface design)

IT

Reference model for the work processes Configuration and requirements management

Knowledge and information management

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means of translating and representing the demands in the evaluation metrics of the HTI of complex systems, on the other, are the central research problems in this study. The approach and metrics will be developed in connection to NPP control room design cases.

2.4.2 Application possibilities of systematic requirements management in the improvement of nuclear safety in Finland (APSReM)

The main objective of the work is to study and to show a summary of good Requirements Management practices and application possibilities in the selected application areas. The most important application areas are the authority activities (STUK) and modifications and procurement of nuclear power plants.

Detailed objectives of the work are:

- to find out the state-of-the art practices and research situation of the Requirements Management on the important application areas

- to map the needs for Requirements Management and the application possibilities of authority activities and other application areas

- to study more deeply the practices of Requirements Management against the emerged needs

- to assess on-going research of Requirements Management practices - to identify necessary research topics

- to present the results of the work to wider audience

2.5 ORGANISATIONS AND SAFETY MANAGEMENT

Currently the work in SAFIR in this area is performed in the project Organisational culture and management of change (CulMa). In this area the work done beyond the SAFIR both in Finland and in international projects will be discussed and reported in suitable extent. Figure 8 describes the research area.

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Figure 8. Research themes in organisations and safety management area [1].

2.5.1 Organisational culture and management of change (CulMa)

The aim of the project is to enhance the understanding of the effects of organisational culture and different ways of organising work on the safety of nuclear power production. Goal is to develop methods and models with which to take into account organisation's culture e.g. in change situations so that all of the criteria for effective organisation (safety, productivity and health) are adequately considered.

2.6 RISK-INFORMED SAFETY MANAGEMENT

The research area includes currently two projects Potential of Fire Spread (POTFIS) and Principles and Practices of Risk-Informed Safety Management (PPRISMA). In this area, too, the work done beyond the scope of SAFIR both in Finland and in international projects will be discussed and reported in appropriate extent. Figure 9 illustrates the research field.

5. Organisations and safety management

understanding cultural aspects

implementation of changes

changes in age structure

improved productivity

& efficiency development

of technology changing

procedures and habits

maintaining knowledge and expertise management

and decision making

work load and wearout bringing new

technology into operation

preventing routine effects

Theoretical development

Practical problems

Pressure for change

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Figure 9. Research themes in risk-informed safety management area [1].

2.6.1 Potential of fire spread (POTFIS)

According to overall plan of SAFIR [1] PSA-models should be as complete as possible, and include external effects such as floods and fires. Fires alone present a very demanding research problem; there a dynamic approach is essential and needed. The central goal of POTFIS project in fire research is continuing the avenue opened during FINNUS to develop deterministic and stochastic submodels to the same level as other branches of PSA. The major strategic problem during SAFIR is the ability to predict potential of fire spread in given scenarios. To be able to utilize already available knowledge and tools in practise several parallel development lines are needed: (a) input data and its reliability for fire-PSA, (b) ignition and flame spread models of fires, (c) reliability models of active fire protection, (d) assessment of operative fire protection, and (e) special fire themes related to passive systems.

2.6.2 Principles and practices of risk-informed safety management (PPRISMA)

Risk-informed safety management means use of information from probabilistic safety assessment (PSA) to support decision making in various contexts. Generally, the project deals with the whole scope of risk-informed methods and application areas related to safety of nuclear power plants. The main objectives are:

− to develop risk-informed decision making methods that integrates results from risk and reliability analyses with other expertise in the problem domain

Development of risk analysis Development of risk analysis

Level 1 PSA:

plant design Level 1 PSA:

plant operation Level 1 PSA:

plant operation Level 2 PSA: severe accident management

Level 2 PSA: severe accident management

Fire safety and risk analysis

Dynamic reliability and risk models Risk-informed

maintenance strategies Structural reliability

and failure models

Cost-benefit evaluations in design, operation and monitoring Multidisciplinary, risk-informed practices involving different technical areas

Multidisciplinary risk analysis related to plant lifetime

Risk analysis of external effects

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− to develop assessment methods for nuclear power plants operation and maintenance in order to enhance risk-informed ways of planning of activities and acting in situations

− to develop methodologies in the problem areas of PSA

− to advance skills in nuclear risk analysis, assure the competence transfer to the new generation and to participate in international co-operation.

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3 COSTS AND FUNDING

The total cost of the programme in 2003 is planned to be € 4.05 million. The major funding partners are KTM with € 1.11 million, VTT with € 1.11 million, STUK with € 0.79 million, TVO with € 0.44 million, NKS with € 0.13 million, Fortum with € 0.17 million, TEKES with € 0.10 million and other partners with € 0.20 million. The volume, funding and costs of SAFIR in 2003 have been illustrated in Table 2. The total extent of the programme in 2003 will be approximately 31 man years. The share of personnel costs is nearly 80 % of the yearly expenses, as illustrated in Figure 10.

Financing in 2003

0 200 400 600 800 1000 1200

KTM VTT STUK Fortum TVO NKS Other

kEuro

Expenses in 2003

0 500 1000 1500 2000 2500 3000 3500

Personnel Mat&suppl Travel Extern serv. Other

kEuro

Figure 10. Financing and Expenses of SAFIR in 2003.

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Figure 11 illustrates the distribution of funding and person years between the six research areas of SAFIR. The most “nuclear-specific” research areas 2: Reactor circuit and structural safety, 3: Containment and process safety functions and 1: Reactor fuel and core have the largest shares, whereas the three remaining areas with more connections and applications beyond the nuclear field, namely 4: Automation, control room and information technology, 5: Organisations and safety management and 6: Risk-informed safety management total into 21 % of the entire programme volume.

Distribution of funding in SAFIR research areas in 2003

19 %

30 % 30 %

5 % 5 %

11 %

1 2 3 4 5 6

Distribution of person years in SAFIR research areas in 2003

21 %

27 % 31 %

5 % 5 %

11 %

1 2 3 4 5 6

Figure 11. Distribution of funding and person years in SAFIR research areas in 2003.

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22 Table 2. Volume, expenses and financing of the SAFIR projects in 2003. erillinen exel-taulukko SAFIR Projects 2003 with KTM 90 k EUR reservation included in Expenses and Financing20.05.2003 Changes in funding in THEA,POOLEX and CULMA from tables of 12.03.2003E.K. Puska ExpensesFinancing VolumePersonnelMat&supTravelExt servOtherTotal 2003KTMSTUKFortumTVONKSEUVTTOtherKTM res. pers monthkeurokeurokeurokeurokeurokeurokeurokeurokeurokeurokeurokeurokeurokeurokeuro Enhanced Methods for48489020511525907003050320010 Reactor Analysis (EMERALD) High-Burnup Upgrades in Fuel20,517203404210557500004040 Behaviour Modelling (KORU) Integrity and life time of reactor circuits77,9758,866103,21209105725526460502503007825 (INTELI) Ageing of the Function of the Containment1,11300001360000070 Building (AGCONT) Participation in the OECD NEA Task Group 0,67,901,58009,4823,5000040 Concrete Ageing (CONAGE) Concrete Technological Studies Related to the Construction, Inspection an Reparation 897,10,50,820,2100,6018,50000082 of the Nuclear Power Plant Structures (CONTECH) WAll Response to Soft Impact (WARSI)13652,72671137,737,7400600000 The integral code for design17204,80200206,8005252000102,8 basis accident analyses (TIFANY) Thermal hydraulic analysis of12133016335187471111110010205 nuclear reactors (THEA) Severe Accidents and nuclear17183,66,521,9101,91,7315,6655,3810,76120,330099,13015 containment integrity (SANCY)