j. smit (editor)
TRANSCRIPT
SEPTEMBER 1993 ECN-93-015
energy innovation
ECN NUCLEAR ENERGY A bird's eye view
J. SMIT (EDITOR)
llllllllllil KS00162887S R: FI DE005S61973
The Netherlands Energy Research Foundation ECN is the leading institute in the Netherlands for energy research. ECN carries out basic and applied research in the fields of nuclear energy, fossil fuels, renewable energy sources, policy studies, environmental aspects of energy supply and the development and application of new materials.
ECN employs more than 900 staff. Contracts are obtained from the government and from national and foreign organizations and industries.
ECN's research results are published in a number of report series, each series serving a different public, from contractors to the international scientific world.
This green series contains ECN rs corporate documents - about e.g. budgetting. organization, programming -which do not belong to the scientific or technological output.
Het Energieondeizoek Centrum Nederland (ECN) is het centrale instituut voor onderzoek op energie-gebied in Nederland. ECN verricht fundamenteel en toegepast onderzoek op het gebied van kemenergie. fossiele-energiedragers. duurzame energie. beleids-studies, milieuaspecten van deenergievoorziening en deontwikkeling entoepassing van nieuwe materialen.
Bij ECN zijn ruim 900 medewerkers werkzaam. De opdrachten worden verkregen van de overheid en van organisaties en industrieen uit binnen- en buitenland.
De resultaten van he*. ECN-onderzoek worden neer-gelegd in diverse rapportenseries. bestemd voor ver-schillende doelgroepen, van opdrachtgevers tot de internationale wetenschappelijke wereld.
Jeze groene serie bevat de algemene ECN- docu-menten - over bijvoorbeeld begrotingen, organisatie, programmering - die niet tot de technisch-weten-schappelijke output behoren.
Netherlands Energy Research Foundation ECN P.O. Box 1 NL-1755ZG Petten the Netherlands Telephone : +31 2246 49 49 Fax :+31 2246 44 80
Energieonderzoek Centrum Nederland Postbus l 1755ZG Petten Telefoon : (02246) 49 49 Fax : (02246)44 80
© Netherlands Energy Research Foundation ECN © Energieonderzoek Cei is um Nederland
ECN NUCLEAR ENERGY A bird's eye view
J.SMIT(EDrrOR)
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This report is drawn up for an audit of ECN Nuclear Energy to be performed by the Massachusetts Institute of Technology (NIT).
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CONTENTS Page
1. INTRODUCTION 5
2. SAFETY OF NUCLEAR REACTORS 6
2.1. Introduction 6 2.2. Neural Networks 8 2.3. Early fault detection 10 2.4. Dynamic signal analysis 12 2.5. Radiobiology 15 2.6. Coordination project with institutes in Czech and Slovak republics 18 2.7. Uncertainty analysis of probabilistic accident consequence codes 20 2.8. Documentation systea power reactors 21 2.9. Risk and safety of thermal reactors 23 2.10. Core-concrete interactions 25 2.11. The thermochemical database of ECN Nuclear Energy 28 2.12. Thermochemistry of source term materials 29 2.13. The chemical behaviour of tellurium during reactor accidents 31 2.14. Volatile organometallic compounds for nuclear technology 32
3. NUCLEAR FUEL CYCLE 33 3.1. Introduction 33 3.2. Research for nuclear safeguards 34 3.3. DEBORA 38 3.4. EVEREST 39 3.5. The 600 D borehole project 40 3.6. PROSA 42 3.7. Participation in the HAW project 44 3.8. Direct disposal of spent fuel elements (DOS) 48 3.9* Radiation damage in NaCl 49 3.10. Radio-ecology 52 3.11. Thermal conductivity of high-burnup U02 fuel 54 3.12. Transmutation of nuclear waste (public information) 55
4. ADVANCED REACTORS 56 4.1. Introduction 56 4.2. Reactivity effects in advanced reactors 58 4.3* Nuclear data for fission reactors 60 4.4. HTR power transients and reactivity 62 4.5* Fracture resistance and creep fatigue damage of irradiated
austenitic stainless steel plate and weld materials 64 4.6. Programme to enhance nuclear competence (PINC) 66 4*7* Dynamics of power reactors 69 4.8. Creep experiments structural steels 71
5. ENGINE PROGRAMME 73 5.1. Introduction 73 5.2. Low-activation construction materials 75 5.3« Fuels for innovative reactors 77 5.4. Advanced fuel cycles and non-proliferation 78 5.5. Studies on transmutation of nuclear waste 8l 5.6. Reactor physics aspects of PRISM 83 5»7» Technological research RAS 84
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6. FUSION TECHNOLOGY 86 6.1. Introduction 86 6.2. NET related technology prograaae 87 6.3- Solid breeder blanket prograne 90 6.4. Long ten prograwse 91 6.5. JET and NET contracts 93
7. APPLICATIONS 94 7.1. Boron Neutron Capture Therapy
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1. INTRODUCTION
The nuclear research is an important part of the activities of ECN Nuclear Energy, one of the six Units of the Netherlands Energy Research Foundation (ECN) at Petten. The activities are subsidized by the Ministry of Economic Affairs and divided in a basic part of about 5 Billion dollar and a program part of about 9 »illion dollar. The total subsidy is about 35% of the yearly turn over of the Unit. The total staff of the Unit amounts about 220 of which 65 have a university degree and about the same number followed college of advanced technology. About 40 person years are hired froa other units, mainly in the area of engineering. The criteria for basic subsidy are: - research within the mission for which no external financing is available; - build-up of knowledge and facilities; - doctor studies. Part of this basic subsidy is designated for a common Unit prograa called Engine. This program is focused on a sustainable energy supply system. The nuclear part is described in chapter 5- The rest of the basic subsidy is used for: - build up of knowledge in the nuclear field; - participation in the long term European fusion program; - research and development or new applications for the High Flux Reactor =*s-aad—thajtot Cell Laboratories. The program subsidy is used for four main programmes: 4 safety of nuclear reactors; -f fuel cycle; - advanced reactors; I - ITER related fusion research. / 'These programmes are guided by the Ministry and have a four years time frame with yearly adjustment. For many projects additional financing from the European Communities has been obtained. An important fact for the programming is the availability of the High Flux Reactor« a 45 MW pool type material test reactor, and a hot cell laboratory. The experimental part of the program is mainly oriented on the use of these facilities.
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2. SAFETY OF NUCLEAR REACTORS
2.1. Introduction
The safety related activities of the Unit are centred around the following areas: - surveillance by signal analysis for prevention of accidents; - risk and safety analysis; - thermochemistry for severe accidents; - support of Eastern European Reactor Safety; - radiobiology for neutron radiation effects on humans.
In the field of dynamic signal analysis the major purpose is to develop stand alone (PC based) systems for surveillance, real-time core barrel motion analysis and vibration analysis of power reactors, the instrumentation might also find customers elsewhere: fuel cell and wind turbine surveillance.
With OECD Halden an Early Fault Detection system is developed that will be demonstrated in the Borssele PWR. Further the application of neural networks in the areas are pursued. In the area of signal analyses the emphasis lies on commercialization of the apparatus and software developed in and outside the Netherlands.
In the area of risk and safety the main target is conducting analyses for utilities and licensing bodies. A suite of computer programs and data bases must be maintained under very strict Quality Assurance requirements for installation, updating, assessment and storage. The know-how and expertise of the analysts is improved by participation in international benchmark exercises. A special project is devoted to the uncertainty of probabilistic accidents consequence codes. In cooperation with KfK in Karlsruhe (Germany) and other European organizations, expert judgements will be used. A small but important activity is the more systematic document organization for power reactors built and on the drawing boards. The accessability must and can be improved by retrieval systems on the computer network.
The thermochemistry work in the Unit is aimed both at experimental and analytical work. In collaboration with AEA, Harwell, with CEC support, the molten core concrete thermodynamic models will be supplied with extended input of thermodynamic properties. In this way the models will become more realistic. The accessability of the existing data base will be improved under cooperation with software engineering experts. Commercialization of the resulting data bank is the target. The thermochemistry of source term materials will be investigated with the emphasis on the behaviour of Te, which could contribute very strongly to early fatalities in severe accidents,
Selected topics of the Unit activities have been incorporated in a collaboration project with the Czech and Slovak Republic. This project addresses items which have been identified by the IAEA as urgent problems to be solved for WER's.
In conclusion there is a project that studies the effects of low energy neutron radiation effects on humans. Using the ECN Low Flux Reactor the
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experimental work is conducted in cooperation with laboratories in the UK, Belgium and Poland in the framework of a CEC programme.
The activities in ti:.:.- field of nuclear reactor safety are at present largely financed by the Netherlands government and CEC. It is the intention of the Unit where possible to increase the commercial turnover. This percentage will never be 1002. but for all areas, with the exception of radiobiology, a market exists from core barrel analyzers via consequence analyses to thermochemical data. The future direction of projects will also be determined by its commercial revenues, but in general the shift in emphasis of the program will be controlled by the governmental bodies. Viewpoints of national licensing authorities, NRC and IAEA will influence the direction.
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2.2. Neural networks (project id no. 1300)
Introduction
A large number of process paraaeters and system interactions may require operator intervention especially in abnornal operational conditions. By aeans of operator support systems a number of uncertainties during decision-making situations are removed. Introducing real time diagnosis methods it is possible to increase the reliability and safety of nuclear reactors. In this way failures which lead to reactor trips or other failures which may diminish the reactor safety can be avoided. As a new technology, neural networks offer an information processing technology for the implementation of real time diagnosis methodologies in a nuclear power reactor.
Project definition
The reliability of the information which is made available to the reactor operator depends on the reliability of signals which are coming from the sensors. The effective utilization of this information is dependent on the presentation of this information in a pragmatic format. Neural networks showed that they can be of support in the relevant problems L1~3L
Project description
The following subjects form the parts of the proposed research.
a. Diagnosis: status identification of the reactor Status identification concerns two different type of failures in a reactor. These are sensor failures and system failures. Identification of both type of failures is essential task in an operating reactor. Neural networks provide effective and efficient (real-time) methodologies for failure detection and diagnosis.
b. Noise Data-Base of Borssele nuclear power plant Noise characteristics of Borssele NPP is known up to now along 19 fuel cycles. The data of the 19th fuel cycle can be studied by neural network for seeing if any extra information is to obtain in relation to deviations from normal state [3]. There is a substantial interest in the area of transient identification and classification.
c. Parameter identification Mathematical model of the system is used often with numerical algorithms for the identification of parameters by means of which optimal adaptations of models from the data can be obtained. In the number of cases the above mentioned methods fail to obtain physical information in the case the data are not in the validity range of the model. Neural networks are probably more reliable than the conventional methods to identify the parameters. Neural networks are very important to identify the system parameters in time and frequency domain. One of the interesting application is the estimation of the feedback parameters in pressure water reactor such as moderator reactivity temperature coefficients. These coefficients are computed from CPSD (cross power spectral densities) between colant temperatures and fluxes from in- and ex-core neutron detectors. This study is important for Borssele
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operation during the start-up of the reactor with high boron concentration. Other applications lie within the interests of boiling water reactors.
d. Plant-vide monitoring The techniques of the signal validation can be extended to the plant-wide monitoring with the use of an autoassociative neural network where the input and output of the network are provided with the same variables. The network learns over a range of operational status (startup, normal operation, shut-down and classified disturbances), where the same data are used as input-vector as well as the desired output vector, during the training. After the training, monitoring of the differences between the estimations and the actual values obtained from the sensors of present method for the identification of 'drift' or instrument damage (real-time application). This application is already in tested in 1991-
The method of neural network and the relevant software will in the first place made ready for off-line case studies. In a later study real-time applications will be performed as demonstration of the reliability of the neural network methods. In the course of time, different type of algorithms will be developed and tested in our on-line and real-time environment.
The first application is in the direction of monitoring of the steam generators and both the heat-rate monitoring of the steam generators and the heat-rate monitoring of Borssele NPP.
In relation to neural network studies and applications, a larger neural network is devised. Participants of this network are in cooperation with ECN: University of Tennessee, JAERI, ITU and in the Netherlands: SNN, PAON, TUD, KUN.
Project control
Time table and milestones - Survey applications of neural networks (a,b,c and d): 1st quarter of
1993) - Application of real-time SG (steam generator) monitoring: 2nd and 3rd
quarter of 1993. - Final report: 4th quarter of 1993-
The project will be completed at the end of 1993 after 2.4 manyear/yr.
References
[1] E. Eryurek and E. Turkcan, ECN-R--9I-OO7.
[2] O.Ciftcioglu and E.Turkcan and S.Seker, ECN-RX—91-044. [3] E.Turkcan et al., ECN-CX—90-028. [4] E.Turkcan et.al., ECN-RX--91-057-
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2.3. Early fault detection (project id no. 1279)
Introduction
The Early pault Detection System (EPD) is a model based failure detection system which is developed in the frame of OECD Halden project [1-8], An EFD-system for an industrial process contains a number of independent, mathematical models which represent different components of the process. The model works in real-time in parallel with the process . The model gives as process status output variables which are to be compared with the measured values. When the difference between the measured and computed values exceeds the present limit values, EFD- alarm is activated. For surveillance purposes process models are so designed that they represent even the smallest details of the process. This allows the detection of disturbances in a very detailed level. Also disturbances in the instrumentations can be detected.
Project definition
In the first place, the EFD-method will be implemented for Borssele nuclear power plant (NPP) signals through the ECN real-time monitoring system at ECN [9#10]. The primary component from which signals are obtained for investigations is the steam generator of the Borssele NPP. In the second place, application of the methodologies and the implementation of the software at ECN site will be performed. The software will be tailored to the network's operational conditions at the Borssele NPP and the EFD method will be implemented at the power plant site. In the third place, the method will be tested for the pressurizjr of the primary system of the plant.
Project description
The method will be applied only for the prediction of the feed-water flow [11]. The EFD method will be adjusted for the actual signals of the Borssele NPP. In this configuration the method will be tested for different operational conditions of the nuclear power plant. In the following year the following activities will be performed:
a. The feed-water flow prediction will be implemented and reported. b. The failed model will be studied and corrected and the sensitivity
analyses will be performed. c. System software will be improved. d. The present application will be reported and a demonstration will be
given to the operational personnel of the Borssele NPP. e. The (new) Borssele network system of the main process computer will be
studied and the necessary signals for the modelling will be selected from the network system. Hence by the implementation in Borssele more signals will be available as well as the pre-heater and feed water pump signals.
f. Application of the developed EFD programs will be carried out in the power plant Borssele with the help of its network facilities.
g. Completed EFD-system will be reported and a proposal will be prepared aiming at the purchase of the system by Borssele.
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Project control
The project will be completed at the end of 1993 after 1 manyear/yr.
January/March 1993- improvement and adjustment of the model leading to an internal demonstration
April/June 1993 ' further experience by means of both modelling and studying the Borssele computer facilities.
August/September : adaptation to the network system in Borssele which will expectedly leads to a commercial order. Besides, conclusive discussion with Halden about the present and further joint work.
December 1993 • Completion of the project and documentation.
References
[I] Proposals for cooperation between ECN and Halden (March 5. 1987). [2] Early Fault Detection system Running on a live power plant, HWR-261
(1990) Halden, Norway. [3] Application of EFD for the advanced ALMR design, HWR-264 (1990),
Halden, Norway. [4] Detailed Diagnosis Based on EFD, HWR-267 (1990), Halden, Norway [5] A. Sorensen, EFD at the Loviisa NPP simulation method (1990). [6] T.Suzudo , O.Berg, EFD size estimation and prognosis using process
models and improved presentation technique (1990). [7] Principles of EFD and Signal Validation.HWR-279 (1991), Halden, Norway [8] Implementation and First experience with signal validation system flow
sensors, HWR-280 (1991), Halden, Norway. [9] O.Ciftcioglu and E. Turkcan, ECN-RX—91-026. [10] E.Turkcan et al., ECN-RX--91-057. [II] E.Turkcan,et al., memo NFA-DSA-91-17.
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2.4. Dynamic sigr.*l analysis (project id no. 1'403)
Introduction
On-line dynamic signal analysis with the aid of computer systems is a potential powerful methodology for performing surveillance of complex industrial processes. At ECN such a system is developed for surveillance of nuclear reactors (Borssele and HFR ), where application field can be beyond the nuclear field. The developed measurement and analysis system gives the possibility of utilization of these methods in order to identify the dynamic behaviour of the system whose dynamic characteristics are required to be known.
Problem definition
New application possibilities, especially for externally developed methods in the field of on-line monitoring techniques for surveillance of nuclear power reactors as PWR, BWR will be studied. Also, direct connection with the above-mentioned goals vill be strived. Optimization of the present hardware and software system will be carried out and using the experience gained new developments will be initiated or up to date followed.
Project definition
The realization of more compact on-line measurement system based on a PC-system will be carried out. By this th-2 requirements of potential prospective customers will be met.
The realizability of non-nuclear applications will be investigated separately. There are contacts with the Industrial Board for Oceanography and international offshore organizations (a.o.,ELF-Petroland). In order to enhance the commercial chances further, various industrial fairs will be participated together with optional demonstration of the equipment.
Applications in general:
- The surveillance and control of processes: change of "noise signatures" can be an indication of a disturbance or a change in the process which requires operator intervention.
- A special case of this is the in-situ test of measuring channels and sensors, which are not accessible otherwise (e.g.,. in nuclear reactor or petro-chemical industry).
- The estimation of physical parameters of a process by means of models (e.g., autoregressive) which are adapted to the signals. Examples: step-response , time-constants, derived physical parameters, state estimation and transfer-functions.
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New developments will mainly lie in further automation and user friendly development of the methods:
- Signal validation {referred to as verification of signal output). - Further development of the pattern recognition of signatures of system
parameters. - Interactive computer use. - Knowledge-based expert systems for routine use. - Integration of surveillance system in process measurement system ( e.g.,
in control-rooms).
Project description
The hardware and the software of the system must be subjected to maintenance. The hardware maintenance includes:
- Borssele on-line instrumentation system. 32 channels »AC/DC converters, micro-processor.
- Data communication Borssele-Petten. - Main data processor system at Petten; front-end system (PDP 11/24), host
computer (VAX 4200); array processor (FPS-5105), work-stations (WS 2000, WS-3100, DEC-Alpha) for real-time analysis.
- HFR-system; 16-channels mobile-system for various applications through telephone lines.
- PC on-line measurement system; menu oriented real-time 4/8-channels signal analysis system equipped with DSP-card and build-in ADC. Three systems are available: a. Dodewaard stability monitoring system. b. General signal analyses system (system for Indonesia or HFR). c. Core barrel motion and primary system integrity analysis system for
PWRs.
Software maintenance includes the system software maintenance and the optimization and maintenance of several computer programs (about 120) used for special applications.
An important part is the design of display facilities and integrated user interfaces for user friendly programming. Here it is noteworthy to mention of two software packages named DATAVIEW and UNIRAS.
Other ECN projects will be participated when advanced signal analysis techniques are needed, a.o.:
- HFR stability and boiling detection for the purpose of reactor operation and special experiments.
- Measurements of the dynamic characteristics of the fuel-cells. - Real-time applications for wind energy.
This work can be extended outside of the ECN, e.g., for request from KFD, Dodewaard reactor, off-shore experiments.
International jointwork connections will be established in order to investigate new techniques. The utilization of neural networks in combination with signal processing techniques will be investigated in the cooperative work with the University of Tennessee (TU), Istanbul Technical University (ITU) and the Unit Informatics at ECN. The scientific results
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will be reported in the form of publications and presentations at international conferences.
a. - Improvement of the present on-line computer system by connecting two real-time systems directly to VAX 4200 and connecting two array processors to the main frame.
- Goal-oriented on-line signal analyses system development where alarm procedures ave included.
b. Completion of 8-channel PC-based data acquisition system for commercial use: - Stand-alone surveillance system. - Stand-alone real-time core barrel motion analyzer. - Stand-alone surveillance system for vibration analysis.
Project control
In 1993 it requires 2.4 manyear/yr.
References
[1] 0. Ciftcioglu and E. Ttirkcan, Sequential decision reliability concept and failure rate assessment: application to nuclear power plant surveillance instrumentation. Report presented at the IAEA specialists* meeting on analysis and experience in control and instrumentation as a decision tool, Arnhem, 16-19 October 1990, ECN-RX—90-085.
[2] 6. Ciftcioglu and E. TUrkcan, Failure detection by adaptive lattice modelling using Kalman filtering methodology: Application to NPP. Paper presented at the Symposium on nuclear reactor surveillance and diagnostics, Gatlinburg, 19-24 May 1991. ECN-RX—91-025.
[3] 0. Ciftcioglu and E. TUrkcan, Sensor failure detection in dynamical systems by Kalman filtering methodology. Paper presented at the International conference on dynamics and control in nuclear power stations, London, 22-24 October 199i. ECN-RX—91-026.
[4] T.T.J.M. Peeters, 0. Ciftcioglu and E.Ttirkcan, An innovations-based method for DC-signal failure detection: application to NPP. Contribution to the symposium on nuclear reactor surveillance and diagnostics, Gatlinburg, 19-24 May, 1991, ECN-RX—91-043-
[5] 0. Ciftcioglu, E. Ttirkcan and S. Seker, Failure detection studies by layered neural network. International ASME conference neural networks, 29-31 May 1991, San Diego, USA, ECN-RX--91-044.
[6] E. Eryurek and E. TUrkcan, Signal processing and neural network applications in pressurized water reactors, ECN-R—91"007«
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2.5. Radiobiology (project id nr 1^10 and 5240)
Introduction
The importance of the Hiroshima bomb survivor data with respect to neutron radiation risk analysis is well understood. Recent observations concerning thermal neutron activation analysis of minerals and metals in Hiroshima suggested that the DS 86 calculations for low energy neutrons are in error and are two to ten fold higher than previously reported. Furthermore, these low energy neutrons (E-0.2 MeV) were reported to be at least as effective in producing cytogenetic damage as the medium energy fission neutrons (E-l NeV) more commonly used in radiobiological experiments. The effects of dose protraction or fractionation with high LET radiation are as yet unclear. The majority of tumour induction studies in experimental animals showed no effect of dose protraction with high radiation. However, in some studies, it has been shown that for carcinogenesis and life shortening, and for in vitro oncogenic transformation, protraction of a given dose of medium LET radiation produced increased biological effects relative to single acute exposures* Additional knowledge about the biological effectiveness of neutrons for radiation induced carcinogenesis in relation to dose rate and neutron energy is needed for an adequate risk analysis after exposure to neutroas.
Because of its sensitivity, the radiation induced dicentric in human lymphocytes has been used to probe the effect of radiation at low doses. When scoring of aberrations is shared by several laboratories a low dose limit for low LET radiation of a few tenths of mGy has been achieved. Similar information concerning high LET radiation is lacking.
Problem description
The problems dealt within the projects are: a. Does protracted exposure of mice to neutrons result in differential
tumour incidences or reduction of lifespan? b. What are the cytogenetic and molecular mechanisms involved in radiation
myeloid leukaemogenesis? c. Is genetic predisposition a major factor for radiogenic leukaemia? d. Extrapolation of high doses for cytogenetic effects. e. The biological efficacy of low energy neutrons.
The objectives of the projects are: a. To get information about the effect of dose protraction with high LET
neutron irradiation on the induction of acute myeloid leukaemia and other tumours, and lifeshortening in mice.
b. The principal objective of this part of the project is to gain a detailed understanding of the complex mechanisms that underlie the radiation induction of acute myeloid leukaemia (AML) in the mouse.
c. The objective is to determine which of the four genotypic variants for telomere-like sequence arrays present in the CBA/H inbred strain is predisposing for the development of radiation induced AML in CBA/H mice.
d. To determine whether linear extrapolation from high doses to low doses is possible or that a threshold exists.
e. To gain information in cell culture systems on the relative biological effectiveness (RBE) of moderated fission neutrons.
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Project description
a. For the induction experiments a mouse model has been chosen that showed an extremely low incidence of spontaneous AML (about 1 in 2000 animals) Thus each observed AML (maximal 20-25%) can be regarded as true being radiation induced. Randomly chosen male CBA/H mice will be exposed to either 2.0 Gy X-rays at 3 mGy/min (n=200) and 600 mGy/min (n^9Q) or 0.4 Gy fission neutrons at 0-5 mGy/min (n=125) and 3n0 mGy/min (n=125). As control groups 200 mice will be included. For the non-orotracted exposures AML yield will be 18% and 11% for X-rays and fission neutrons respectively. These will be histopathologically characterized and their malignant characteristics will be determined using in vivo passage procedures. By keeping the irradiated animals during their lifespan, valuable information will be obtained concerning the effects of dose protraction with low and high LET radiation on the induction of tumours other than myeloid leukaemia and will determine whether tumour types show a differential response to dose protraction. Life shortening will be analyzed using Kaplan-Meier procedures. This project is being carried within the framework of a CEC-project together with Dr. R. Cox (NRPB, United Kingdom) and Dr. M. Janowski (VITO, Belgium).
b. DNA plugs from primary AML's and normal host tissue (brain), and bone marrow smears will be obtained from the experiment described above. The DNA plugs will be processed, digested, ard subjected to electrophoresis. DNA fragments will be transferred to membranes which will be send to the NRPB for analysis of interstitial telomere lik^ repeat sequence (TLR) restriction length polymorphisms (RFLT) at to determine the phenotype of the induced AML's in comparison with that of the host tissue. Blood smears of AML inflicted animals will be reprocessed and send to NRPB for cytogenetic analysis to obtain information about the fragile chromosomal sites involved in radiation induced leukaemogenesis.
c. Animals from the CBA/H colony in Petten will be screened by the NRPB for TLR sequence RFLP variants. Animals with the appropriate phenotype will inbred to get a true breeding TLR polymorphism substrain. Animals front this substrain will be irradiated with X-rays in order to confirm the anticipated increase in leukaemogenic radiosensitivity associated with TLR polymorphisms. The AML's and host tissues will be analyzed by NRPB as described above. In addition mice with the appropriate phenotype will be crossbred with C57*>1 mice which is genetically divergent enough to distinguish parental genetic contributions in their Fl offspring. Specific DNA lesion in Fl neoplasms induced by radiation will be characterized by NRPB by los3 heterozygosity for a set of linked markers to identify specific chromosome 2 deletions in AML's induced in hybrid mice.
d. Human lymphocytes will be exposed to graded doses of fission neutrons (doses ranging from 1-1000 mGy) to establish dose response relationship for chromosomal aberrations. At a later stage chromosome painting using fluorescent in situ hybridization techniques (FISH) will be involved. Irradiated cells will be scored by a number of European cytogenetics laboratories including NRPB (Dr. A. Edwards) and Sylvius Laboratories (Prof.Dr. A. Natarajan). From these experiments, information will be obtained how sensitive the methods are for low dose exposures with respect to in vivo dosimetry after accidental exposure and whether a threshold dose exists for the induction of chromosome aberrations.
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In collaboration with the University of Cracow Poland, Tradescantia plants and human lymphocytes will be exposed to fission neutrons in order to obtain dose effect relationships for mutations and chromosomal aberrations respectively. This part 01" the project is a framework of a CEC-project.
e. Fission neutrons generated in the 235U converter of the BIBOP facility of the Low Flux Reactor will be moderated using polyethylene to lower mean energies. Using the moderated neutrons, V79 cells will be exposed and assayed for colony formation. Biological efficacy of low energy will be compared with that obtained with X-rays. The biological efficacy of the various neutron spectra will be expressed as the percentage kerma from the 0.1 to 1.0 MeV energy interval. This approach could provide quantitative information for neutrons of various mean energies and is independent of microdosimetry methods.
Project control
The project parts a-c are part o;~ c large CEC project where the Radiobiology and Radioecology group of ECN will carry out all animal experimentation and preparative molecular biology. Actual probing and development of probes will be carried out by NRPB and VITO. The whole project is planned to be finished in 1998. First results concerning AML typing will be available at the end of 199^• Overall analysis of tumour incidences and survival analysis and publication of results will be completed late 1998- Manpower involved is at present 2 manyears/year and will increase to 2.5 manyears/year in 1994.
Project part d. involves only irradiation and dosimetry whereas actual scoring of the biological endpoints will be carried out by other laboratories. Experiments will be completed early 1995« Publication of results is expected to be early 1996. Manpower presently involved is 0.1 personyear/year.
Project part e. will start late 1993 with neutron spectrum and kerma measurements of fission neutrons with various degrees of moderation. Late 199** f"rst biological experiments will be carried out which should be completed mid 1995' Publication of results is envisaged early 1996. Manpower involved is expected to be 2 personyears/year.
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2.6. Coordination project with institutes in Czech and Slovak Republics {project id no. 1303)
Introduction
The safety of nuclear reactors in Eastern Europe receives a lot of international attention. According to Western organisations the safety of the Russian designed reactors are unacceptable and need to be improved. The bilateral contacts between research institutes in Rez and Bratislava in former Czechoslovakia and ECN has resulted in projects to improve the safety of the nuclear plants in the Czechian and Slovakian republics which in cooperation can be executed. These contacts were established in 1991. Application for financing of these projects is requested from the European Community (EC) and interested national organisations but have up to now not been successful, also not all these projects are rated as most urgent as indicated by the IAEA ranking (1). In the mean time projects are defined by the Czechian authorities for the Bohunice nuclear powerplants in Slovania which will be financed by the EG budgets under the Phare program. International tenders will be requested to execute these projects. It is the intention of ECN to tender for these projects which in part are similar to the project defined in the bilateral contacts. The experience received from this project helps to increase the chances for receiving these contracts.
Problem definition
This project, which at least partially contributes to solve the IAEA defined most urgent safety problems, can also be considered as a starting phase for the expected EC projects The project consists of four parts: - Installation of a corebarrel analyzer - Training of a reactor physics specialist - Metrology - Process analysis This project will be executed by the Reactor physics and Process analyses department in cooperation with the Czechian and Slovakian institutes.
Project description
The objective of the project is to transmit the ECN acquired knowledge by executing reactor physics- and safety analysis jointly and to train a Czechian reactor physics expert at ECN. In addition ECN will make developed equipment available for installation in, by preference, the Bohunice nuclear power plant in order to receive selected information during reactor operation. With this information the safe reactor operation can be supported. The following project parts will be executed.
a. Installation of a core barrel motion analyzer During this project part a similar analyzer available will be made suitable for use in a pressurised water reactor. Together with the Czechian organisation the reaccor plant will be selected and the equipment installed. The experience received from this project can be used to install similar equipment in other East-European reactors.
19
b. Training of reactor physics expert In this project uncertainties of reactor physics calculations concerning reactor damage surveillance will be determined. The intention is to make the ECN knowledge available which can be used for similar calculation of pressurised water reactors in Czech republic. The training programme at ECN will take six months.
c. Metrology The embrittlement of the reactor vessel as resulting from radiation threatens the safety of the nuclear plants especially in Eastern Europe It is the intention of this project part to increase the knowledge about the factors contributing to increased embrittlement.
d. Process analysis For the safety analysis ECN and the Czech institutes use amongst others the RELAP computer program. As initial phase of a complete analysis an input deck for the WER reactor will be produced. The experience ECN obtained performing similar analysis will be used. ECN will receive detailed knowledge about the WER nuclear power plant. This experiences can contribute to receive safety analysis contracts for the EC programs.
Project control
The four parts of the project started in the second quarter of 1993 end will be completed in the last quarter of 1993« The manpower is 0.5 manyear/year.
References
1. IAEA-TECDOC-640. Ranking of safety issues for WWER-440 model 230 nuclear power plants. February 1992.
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2.7. Uncertainty analysis of probabilistic accident consequence codes (project id no. 1409)
Introduction
The risks of a nuclear power plant are assessed in a probabilistic safety assessment (PSA). Part of that assessment is the analysis of the consequences of accidents for people living around the plant. Tools for this part of the assessment are probabilistic accident consequence codes. Knowledge of the uncertainties associated with the results of these codes has an important role in the effective allocation of research and development effort toward the reduction of risks.
Problem description
Fairly comprehensive assessments of the uncertainties have already been made. Fundamental to these assessments were estimates of uncertainty or probability distributions of values for each of the important model parameters. The major criticism is that these estimates were largely made by those who developed the accidents consequence codes as opposed to experts in each of the many different scientific disciplines involved.
Project definition
The objectives of the project are to develop and apply expert judgement elicitation techniques in estimating the uncertainties associated with the predictions of probabilistic accident consequence assessment codes, and to investigate the use of the results of these studies as input to uncertainty analyses of such codes.
Project description
ECN collaborates with organisations similarly involved in the study, currently envisaged to be KfK, SRD and NRPB. ECN contributes to the study in the following ways: 1. To assist in the selection of appropriate experts in specific subject
areas of consequence assessment (see below). 2. To provide experts both in the general area of accident consequence
assessment and in particular modelling aspects (see below), who will interact with and advise the chosen experts in other fields.
ECN will contribute, as identified in 1. and 2. above, to the following three subject areas: - Early health effects (with support from KfK). - Behaviour of deposited material and calculation of related doses (in
support of SRD). - Late health effects (in support of NRPB).
Project control
ECN is involved in the project from mid-1992 to mid 199**.
The EC-funding (50JK) for the participation of ECN is 15.000 ECU. The mtuipower is about 0,2 manyear/year.
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2»8. Documentation system power reactors (project id no. 1^99)
Introduction
A tremendous amount of documents is being published in the field of nuclear energy. Although much information can be retrieved by library services, a systematic survey of the existing power reactors and the principal documents could be an asset for the ECN-Nuclear Energy unit. It is expected that a systematic description of required information (on Power Reactors in this project) will enhance the service to the potential users of the documents and at the same time reduce the costs substantially.
Problem description
Reactor Assessment requires an overview of existing reactor technology and regulations, and has to anticipate on future problems. Generally an adequate amount of information related to the design and application of nuclear power plants is received by ECN Nuclear Energy. However, this information is not ready available for reacting on actual news items, or for use in the ongoing projects. Ad hoc requests for documents are by nature often incomplete (which complicates the care for quality control). A systematic documentation can be a helpful means to ease the execution of various projects, including PR activities.
Project definition
Objective The objective of the project is to provide direct access to relevant reports on all Power Reactors, and on current research and development activities. The services of library will be used to the extent possible. It will serve as the focal point for questions regarding nuclear energy; both for technological and scientific evaluation, and for support of Public Relations activities.
Results - A list of power reactors (in operation, suspended or decommissioned)
showing their main characteristics). - A survey of the kind of required documentation on power reactors (safety
reports, probabilistic safety analyses etc.)/an up-to-date view of the available documentation on power reactors, un up-to-date list of required documents.
- An up-to-date survey of the rules, regulations and standards as released by: IAEA, OECD/NEA, USNRC, EC/ an up-to-date list of the required documents/an up-to-date list of the available documents.
- An up-to-date survey of the nuclear laws/practice in Germany, The Netherlands/an up-to-date list of the required documents/an up-to-date list of the available documents.
- A view on selected specific subjects (e.g. ground contamination). - An up-to-date survey of public information reports, leaflets etc. (in
the field of power reactors) by: . National governments/ministries. . National and foreign research institutes (e.g. ECN, GRS, KfA, KfK, Halden, Studsvik).
. Professional organisations (Netherlands Nuclear Society, European Nuclear Society, American Nuclear Society, etc.).
22
. Non-professional organisations (Natuur & Milieu, US Council for Energy Awareness, etc.).
. Utilities and distributors (Borssele, Dodewaard and foreign)-
. Industry (Canada, France, Germany, Japan, Sweden, USA, etc.)« Relevant newspaper articles etc. are believed to be covered already.
Limitations Detailed information on Reactor Physics, Reactor Chemistry, Reactor Materials etc. is not included in this project.
Project description
It is believed that interviews with colleagues in and outside the section will lead to consensus on the characteristics of the kind of the required information. The library service is already well equipped to look up the required documents in the relevant data base collections.
Project control
The project started in 1993- The commitment is 0.^ manyear per year in 1993 and 199** • For the years 1995. 1996 and 1997 0.1 manyear per year is foreseen.
Reference
1. Memorandum 9009/BvdS/MH-333A (in Dutch)
23
2.9. Risk and safety of thermal reactors (project id no. 1^60)
Problem description
The risks associated with the operation of thermal nuclear power plants must be limited to acceptable levels. Quantification of these risks is possible by means of computer programs which simulate the progression of an hypothetical accident, computer programs by which the frequency of occurrence of these accidents can be estimated, as well as computer programs by which the radiological dose and resulting risks to the public can be calculated. Proper use of these computer codes requires general know-how in the field of reactor safety as well understanding of the models which are incorporated in the different computer programs. Computer programs and know-how are also necessary for the assessment of possible measures which can be taken during the course of an accident, both inside and outside of the plant, to limit the consequences as much as possible.
Project definition
The objective of this project is to maintain general know-how in the field of reactor safety and to obtain and update a package of state-of-the-art computer programs necessary to perform quantitative assessment of the consequences of postulated accidents. This package includes computer programs on: - system analysis to determine the accident frequency - process analysis to predict the accident progression - risk analysis to compute the risks for the environment The computer programs must be installed, maintained, assessed and applied according to quality insurance requirements. Assessment of computer programs is achieved by participation in code user groups and code benchmark exercises.
The result will comprise a qualified package of computer programs installed and maintained at the ECN-computer systems. Computer programs developed at ECN will be qualified and documented according to quality assurance requirements.
Project description
The project consists of a large number of activities which can be grouped into: - Participation in different working groups related to thermal reactor
safety of international organizations as OECD/NEA, EC, the IAEA and the USNRC, as well as participation in computer program user groups as CAMP and MCAP.
- Installation, testing and qualification of the computer programs obtained from third parties.
- Development, qualification and documentation of ECN computer programs. - Participation in international standard problem exercises for purpose of
training and assessment of the computer programs. - Research activities for specific safety related subjects in order to
improve ECN competence on these subjects and to contribute to the resolution of unresolved safety issues.
24
Project control
Planning This project extends from 1993 to 1996. The activities to be performed will be defined at the beginning of each calender year. For 1993 these activities are listed in attachment 1. The work will be executed by the sections Process Analysis, Risk Analysis and Reactor Assessment of the group Nuclear Technology of ECN Nuclear Energy. The results obtained will be reported to the Ministry of Economic Affairs on a yearly basis.
The manpower in 1993 amounts to 4.9 manyear.
References
1. E.J. Velema; International Standard Problem 29 (Hydrogen distribution inside a PWR-containment under severe accident conditions). ECN-BUNE memo NP-R&V-92-02.
2. H.A. Roodbergen; International Standard Problem 31 (CORA 13 experiment) on Severe Fuel Damage. ECN-BUNE memo NP-R&V-92-01.
3. H.A. Roodbergen, L. Winters; Berekeningsresultatcn Internationaal Standaard Probleem 31 (CORA 13 experiment). ECN-BUNE memo NP-R&V-92-O3.
4. H.A. Roodbergen, L. Winters; ECN contribution to International Standard Problem 33. ECN-BUNE memo NT-PA-92-09.
5. A. Woudstra, J.P.A. van den Bogaard, P.M. Stoop; Assessment of RELAP5/M0D2 against ECN-reflood experiments. ECM-C—92-008.
6. E.J. Velema; Description of the CONTAIN input model for the Dodewaard Nuclear Power Plant. ECN-I--92-004.
7. L. Winters; Hydrogen generation following a surge line break in the Borssele Nuclear Power Plant. ECN-CX—92-026.
8. P.M. Roelofsen, A.D. Poley; Richtlijnen PSA-3, onderzoek naar methoden en modellen voor het uitvoeren van een probabilistische consequentieanalyse. ECN-C—92-040.
9. A. Woudstra, L. Winters. RELAP5/M0D2-berekeningen ter evaluatie noodkoelsysteem kerncentrale Dodewaard. ECN-CX—92-069«
10. P.M. Stoop, H. Gruppelaar; Project proposals for improving the safety of WER 440/230 reactors in Czechoslovakia. ECN-BUNE-NFA-Alg-92-04.
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2.10. Core-concrete interactions (project id no. 1469)
Introduction
The assessment of the consequences of severe accidents in nuclear reactors involving molten core-concrete interactions (MCCIs) requires estimates of the quantities and physico-chemical forms of the radioactive s oies released from the melt into the containment atmosphere. Such oimates in turn require a detailed knowledge of the complex chemical in tactions which would occur between the fission products, fuel and the components of the core structural materials and the concrete. In recent years much effort has been put into the thermodynamic characterization of these processes. The results of such studies are important for predicting several aspects of MCCIs, such as: fission product release, and solidus and liquidus temperatures.
The fundamental requirements for the thermodynamic studies are reliable thermodynamic and phase diagram data coupled with the calculation of multi-component, multi-phase equilibria.
Problem description
Thermodynamic models to describe the condensed phase equilibria for MCCIs have been systematically developed over the recent years [1-3]« Models now exist for the six-component system U02-Zr02-CaO-Si02-Al203-MgO. Extensions of this model have to be made by adding BaO-SrO-La203-Ce203 and FeO. The oxygen dependency of U. Zr, and Si should also be considered. These extensions involve many systems of which no, or hardly any, thermochemical data are available (e.g. Ce203-Si02)[4].
The central problem in the calculation of phase diagrams is to obtain a representation of the Gibbs energy of the total system. From this representation the chemical activities (G" ) of the various components can be extracted which are indispensable in thermodynamic calculations of non-ideal systems. If the activities of the various components are known as a function of temperature, pressure, and composition, the chemical behaviour of these can be estimated more accurately.
Project definition
Knowledge of the activities of the condensed species is important for an assessment of the amount of fission products released in an MCCI-accident, since the condensed species determine the activity in the various solid solutions. The actual release is determined by the combination of the thermodynamic properties of the condensed and the gaseous phases. Although most'.y the properties of the less volatile gaseous compounds are treated as "sufficiently known", this is not always the case [5~7]. Much effort should therefore be made to determine insufficiently known or lacking properties.
Project description
In order to obtain accurate thermodynamic properties of the relevant pure substances a critical literature assessment to be made. For the modelling of the binary systems a literature assessment of the phase relations occurring has to be performed. In combination with those found in the
26
critical assessment mentioned earlier, the phase relations can be modelled using appropriate computer models. For the modelling several types of computer models are being applied. As a basic model the associate model is used whereas the sublattice model is being used to model the various ranges of solid solutions occurring.
A ternary, or higher-order system can be calculated from their constituent binary systems. For instance, if a ternary system has to calculated, three binary systems have to modelled with an appropriate thermodynamic model. With the coefficients thus obtained the ternary system can then be calculated. If a quaternary system has to be calculated, six binary systems need to be modelled to obtain three ternary systems which in turn are combined to give the quaternary system. Theoretically, to model a nine-component system, every lower-order system has to be modelled first. In practice, the influence of the systems with an order higher then three are negligible, since two and three component interactions are most likely to occur.
The chemical activities calculated in the multi-component system, will be checked with mass-spectrometrical measurements up to 2400 K. By measuring the intensities {= vapour pressure/T) of chemical species in multi-component mixtures the activities can be found since:
&t = {Intensity of (i) in multi-component system}/{Intensity of pure (i)}
The results obtained in these experiments will be used to check the soundness of the multi-component calculations and to test the assumptions which have be made to arrive at this point.
Project control
The project is part of the MCCI (Molten Core-Concrete Interactions) program of the CEC (Brussels); a financial contribution of 30% will be given by CEC.
The total amount of binary systems of importance in core-concrete interactions, is: 40. Per binary system an average of 90 hours is required. It is estimated that it will take approximately 6 months to perform mass-spectrometric experiments, including discussion of the results.
The work will be performed in collaboration with AEA-Harwell in the period 1 January 199^ - 1 January 1996; the total personnel effort is 2.8 manyear.
References
1. Ball R.G.J., Mignanelli M.A., Proc. of Second OECD (NEA) CSNI Specialists Meeting on Molten Core Debris-Concrete Interactions 1-3 April, p. 257 (1992).
2. Chevalier P.Y., J. Nucl. Mater. 186, 212 (1993).
3. Ball R.G.J., Mignanelli M.A., Barry T.I., Gisby J.A., J. Nucl. Mater. 201, 238 (1993).
27
4. Cordfunke E.H.P., Huntelaar M.E , Thermochemical data for the sillicates and zirconates of barium, strontium, lanthanum and cerium, ECN-C—91~ 017.
5. Jackson D.D. (1971). Report UCRL-51137.
6. Krikorian O.H. (1982), High Temp.- High Press. 14, 363.
7. Glushko V.P., Gurvich L.V., Bergman G.A., Veyts I.V., Medvedev V.A., Khachkurukov G.A., Yungman V.S. (1981), Termodinamicheskie Svoistva Individual 'nykh Vestchestv. Tom III, Nauka, Moskva.
2B
2.11. The thermochemical database of ECN-Nuclear Energy (project id no. 1*153)
Problem description
the advancements in the field of thermochemical modelling have been considerably in recent years. Nowadays there are reliable computer codes for the calculation of the chemical equilibria in multiphase/multicomponent systems (e.G. Solgasmix, chemsage)• The accuracy of such calculations is determined by the accuracy of the input data.
Thermochemical equilibrium calculations are being used in many areas of nuclear technology such as reactor safety, nuclear waste or corrosion resistant coatings. During recent years, the nuclear chemistry group has generated a vast amount of evaluated thermochemical data, which are being stored in a databank. In order to obtained make a more effective use of the data, a user-friendly database is required.
Problem definition
During recent years, the nuclear chemistry group has generated a vast amount of evaluated thermochemical data, which are being stored in a databank. The objective of the present project is the generation of user-friendly retrieval software with interfaces to equilibrium codes, phase-diagram-calculation programs and spreadsheet software. In addition, the content of the database is optimized by updating old chemical systems and inclusion of new ones.
Project description
- In collaboration with the software engineering group, platform-independent software tools have been selected for the development of the retrieval software and a evaluation of the user need? has been made. The retrieval software is being developed at present.
- The contents of the database is critically examined .- the number of elements is being extended with the actinides and lant'ianides.
- Updates have been made for a number of chemical systems treated in earlier studies.
Project control
The personnel effort is 0.7 manyear in 1993'
References
1, E.H.P. Cordfunke and R.J.M. Konings (eds), Thermochemical Data for Fission Products and Reactor Materials, North Holland, Amsterdam, 1990.
2, E.H.P. Cordfunke and R.J.M. Konings (eds), Thermocheniical Data for Fission Products and Reactor Materials: the ECN Database, ECN-RX--93-009 (J. Phase equilibria, in press).
29
2.12. Thermochemistry of source-term materials (project id no. 2211)
Problem description
The amount of radioactive material that is released to the environment during a reactor accident is determined by the many chemical equilibria that might occur in the various stages of an accident. Nowadays, advanced computer codes are available to calculate the behaviour and release of the fission products during normal and accident conditions. The programs (e.g. VICTORIA, VANESA) use extensive thermochemical databases for condensed and gaseous phases as well as solutions. Since the reliability of the results depends heavily on the input data, a considerable effort is made to ensure that the "best data" are available.
Problem definition
The objective of the project is to understand the chemical behaviour of the most hazardous fission products by means of thermochemical modelling. To this purpose a thermochemical database is compiled [1] and thermochemical experiments are made to determine unknown or poorly known quantities. In addition, some small scale interaction experiments are made to study the reaction of fission products and reactor materials.
Project description
The attention is mainly focused on fission products, such as Cs, I, Te, Ba, Sr, Sb, Mo, and Ru and the reactor materials U, Zr and Si. The activities can be divided in the areas: 1. Fission product behaviour in the fuel. 2. Fission product behaviour in the primary cooling system. 3. Core-concrete interactions.
In each case, the most important chemical species have been identified and their chemical properties are critically evaluated. These evaluations frequently show which key parameters are poorly known and an experimental programme is defined on that basis to determine these quantities. Unique facilities are available for performing these experiments, such as enthalpy of solution and drop calorimetry, high-temperature infrared spectroscopy and phase diagram determination.
The work is being done in close collaboration with partners in Europe (CEC, UK-AEA, Siemens, Kernforschungsanlage Karlsruhe) and in the United States (NRC, Battelle (Columbus), Sandia National Laboratory).
Project control
The personnel effort is 5 manyears per year. Contracts with US-NRC (Washington) and CEC (Brussels) cover the period 1992-1996.
References
1. E.H.P. Cordfunke and R.J.M. Konings (eds), Thermochemical Data for Fission Products and Reactor Materials, North Holland, Amsterdam, (1990).
30
2. E.H.P. Cordfunke, R.J.M. Konings, The release of fission products from U02 fuel: thermochemical aspects. J. Nucl. Mat. 201 (1993) 57.
3. R.G.J. Ball, B.R. Bcwsher, E.H.P. Cordfunke, S. Dickinson, R.J.M. Konings, Thermochemistry of selected fission product compounds. J. Nucl. Mat. 201 (1993) 81.
4. M.E. Huntelaar and E.H.P. Cordfunke, The ternary system BaSi03-SrSi03-Si02. J. Nucl. Mat. 201 (1993) 250.
31
2.13» The chemical behaviour of tellurium during reactor accidents (Project id no. 1372)
Problem description
The tellurium is a hazardous fission product since it contributes considerably to the early fatalities in case of severe reactor accidents. Noreover, many tellurium isotopes decay to iodine. The chemical behaviour of tellurium is still poorly known. Interaction with surfaces of the reactor core and the primary system are of prime importance for the understanding of the tellurium behaviour since these interaction can lead to a considerable retention. For example, it has been suggested that the interaction of Te and zircalloy can lead to the formation of ZrTe2, but the formation of SnTe has been proposed also. Similar reactions may occur with reactor materials such as Inconel or stainless steel.
Problem definition
In this study the interaction of tellurium and several reactor materials will be investigated and a chemical model will be developed for its release during severe reactor accidents. The results will be compared with calculations using current models the VICTORIA code.
Project description
The interaction of tellurium and reactor materials will be studied by small scale laboratory experiments. In addition, phase diagrams and thermochemical measurements will be made to support the development of a chemical model.
Project control
The personnel effort is 1.0 man year during the period 1993~1997«
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2.1*t. Volatile organometallic compounds for nuclear technology (project id no. 1372)
Problem description
Organometallic compounds are being used Frequently as precursor materials in Chemical Vapour Deposition technology for the development of corrosion-resistant coatings. The physico-chemical properties of such precursors are needed to select the best process conditions in CVD reactors. Organic tellurium compounds are not only relevant to CVD processes but also to the reactor safety. In the reactor containment, the fission product tellurium can react with organic fragments which are formed through pyrolysis of plastics and epoxy paints. Organic tellurium compounds are not soluble in water and will therefore not be retained in the filter systems thai; are being used for containment venting.
Problem definition
The project has the following objectives: (1) The determination of the ihysico-chamical properties of organometallic
compounds of silicon, titanium ai.r" tellurium. (2) Evaluation of the relevance of the potential formation of organic
tellurium compounds to the reactor safety. (3) Modelling of the chemical processes for CVD of hafnium carbide as a
coating material.
Project description
Vapour pressure, infrared spectroscopy and calorimetric measurements will be performed to determine the relevant physico-chemical properties of the selected compounds. In-situ analysis of thermal decomposition processes will be done for silicon and titanium precursors. Interaction of tellurium and organic compounds in strong gamma-radiation fields will be studied.
Project control
The personnel effort is 1.0 man year during the period 1991~1995-
References
1. M.G.M. van der Vis, E.H.P. Cordfunke, R.J.M. Konings, The thermodynamic properties of tetraethoxysilane (TEOS) and an infrared study of its thermal decomposition. Proceedings EURO CVD NINE, August 1993, Tampere, Finland.
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3. NUCLEAR FUEL CYCLE
3.1. Introduction
Several parts of the fuel cycle are of direct importance in the Dutch f'uclear Programme, and are as such represented in the R & D programme of the unit Nuclear Energy. For instance, knowledge of the fuel, and the bahaviour of the fuel pin is of extreme importance in advanced reactors where the fuel will be used at still increasing burnups.
Although no reprocessing of spent fuel is done in the Netherlands, the problem of radioactive waste is nevertheless an important item. For this reason, ECN contributes to the national programme OPLA in which the safe storage in geological formations (NaCl) is investigated. A dominant question to be studied is the access and the possible removal of this waste in the long term.
The transmutation of long-lived actinides and fission products is part of the European EFFTRA-programme, and of the national RAS-programme. The objective is *;o contribute - via participation in international programmes - to the development of methods to reduce the amount and longevity of high-radioactive nuclear waste.
Finally, participation in an IAEA- programme on safeguards, with emphasis on the determination of the enrichment in enrichment plants, and in the European ESARA programme, should be mentioned.
3*»
3»2. Research for nuclear safeguards (project id no. 2170)
Introduction
The future acceptability of nuclear power gene ation depends on three aspects: non-proliferation, reactor safety, and nuclear waste [1]. About 17 per cent of world electricity is generated by nuclear power plants. Host of these plants are designed, build, and they operate safely and reliably thanks to international exchange of technology, equipment, and materials. Responsible governments don't allow this kind of exchange, unless there are sufficient guaranties that the co-operation shall exclusively serve non-military, peaceful purposes [2]. Within the European Communities, international control on all non-military nuclear activities has been founded on the Euratom Treaty [3]. Worldwide the International Atomic Energy Agency (IAEA)* performs nuclear safeguards inspections, eg in connection with the Non-Proliferation Treaty (KPT) [4]. These independent international inspections support the international confidence about the peaceful character of nuclear activities, which is an essential precondition for the international nuclear trade.
Problem description
The Kingdom of the Netherlands is Party in eg the Euratom Treaty (one of the three treaties on which the European Communities are founded), the NPT, and Member of the IAEA. Consequently there is a need for the Netherlands Government to judge the performance of the organisations created by these international commitments. For the judgement of the technical aspects of safeguards a technical competence, and a good knowledge of the actual developments is necessary. Occasionally some small, well defined projects are executed for the Ministry [5,6,7]. Safeguards are implemented in the Netherlands, on basis of the Euratom Treaty, by the Euratom Safeguards Directorate, and on basis of the Non-Proliferation Treaty by inspectors of the IAEA. The inspectorates have a need for research and development of special safeguards techniques, applicable in the Netherlands situation, and if possible of a more general nature. In particular the IAEA is strongly dependent on technical support from its member states. A proposal for a dutch support programme to the IAEA was made some years ago. The execution is pending, being dependent on an additional external funding from the ministry of foreign affairs. The management of the facilities wants to be advised on the technical most desirable way to implement those safeguards inspections, in respect of their particular requirements. Since 1985 work is done under contract of the Netherlands centrifuge enrichment industry UCN/Urenco. In the last years proliferation relevant issues are getting more attention of the media. The complex interplay between treaties, international inspections, and national ambitions in relation to nuclear weapon proliferation has to be explained in simple and both political and technical correct language.
Project definition
Contributing to the improvement of safeguards effectiveness and efficiency by development and field test of non-destructive measurements (in particular related to safeguarding of the ultra-
35
centrifuge enrichment installations in Almelo), improvements and application of analytical chemistry and mass spectrometry. Maintaining a general competence in the broader area of safeguards application by international co-operation within ESARDA (European Safeguards Research and Development Association), IAEA, and other international contacts. Systems analysis to place the technical work in the proper context of the safeguards systems, their effective and efficient implementation, and adaptions in view of economic, and political developments.
Activities in the area of safeguards started in the late sixties with experimental work and technical advise on safeguards matters to governmental bodies. The project functions as an umbrella under which related tasks are performed, and a long standing nationally and internationally recognized competence is to be maintained.
Results Contributions to international agreement on safeguards relevant issues [8,9.10,11]-Continuing contribution to ESARDA in different committees [12,13,14] of which the convenorship of the ESARDA NDA working group [15,16,17,]. Contributions to consultants and advisory groups since 1971 [18], of the IAEA, and advisory work.
Publications [19,20], conference papers [21], contract reports, internal reports, and memo's give an account of the regular results of the research work.
Project description
Working method It has been a project internal policy to avoid as much as possible the withdrawal of the creative forces that are e sential for the research into administrative matters. Diversity of subjec, cannot be described in one methodological approach. For instance for the enrichment facilities a certain enrichment measurement method in the cascade area has been investigated and internationally accepted [10], but needed further development in respect of some cases where technical complications avoided a straightforward application. Present work promises a successful application [21].
Activities Research and field test of non-destructive measurements (in particular related to safeguarding of the ultra-centrifuge enrichment installations in Almelo) both as free research and related to small projects defined for UCN/Urenco [22,23,2*1,25], and probably in the near future also for the support programme to the inspectorates. Systems analysis in relation to specific objectives eg formulated in two projects for the Ministry of Foreign Affairs. Co-operation within ESARDA, and its working groups. Preparing publications [19,20,7].
Project control
Quarterly progress review by management, regular contacts with customers for the progress in detailed subjects. The quality control is primarily based on the motivation and competence of the persons. Reports are carefully drafted, and subjected to project internal and external review by
36
competent scientists. External financing demands gradually more formalized control methods for relatively smaller research tasks. It is essential to draw a borderline to halt the process of fragmentation.
The main project accounts for the annual total of about two manyear.
References
[I] G.H. Brundtland, Our Common Future. [2] United States of America, Nuclear Non-Proliferation Act of 1978",
Public Law 95-242-Mar.lO. 1978. [3] "Treaty establishing the European Atomic Energy Community", Roma, 25
March 1957. [4] The Treaty on the Non-Proliferation of Nuclear Weapons, signed in
London, Moscow and Washington on 1 July 1968. Entered into force on 5 March 1970.
[5] K. van der Meer, R.J.S. Harry, Safeguards in nuclear weapon states, possible approaches for reduction of inspection effort, ECN April 1991, ECN-CX—91-034 Confidential.
[6] R.J.S. Harry, The starting point of IAEA safeguards, 15th symposium on safeguards and nuclear material management, Augustinianum, Vatican City, Roma, Italia, 11-13 May 1993.
[7] R.J.S. Harry, Safeguards to detect undeclared nuclear activities, INMM 3̂ th annual meeting proceedings, Scottsdale, Arizona, 18-22 July 1993-
[8] Commission Regulation (Euratom) No 3227/76 of 19 October 1976 concerning the application of the provisions on Euratom safeguards.
[9] Convention on the Physical Protection of Nuclear Material, IAEA Legal series No. 12, IAEA in Austria, December 1982.
[10] Hexapartite Safeguards Project Overview, ESARDA bulletin nr. 5 (October, 1983), PP 6-7.
[II] Report of the Chairman and Recommendations of the Technical Committee on Physical Protection of Nuclear Material, 2k April - 5 May, 1989, Vienna, Austria, IAEA Headquarters, and Meeting of experts with the task of drafting recommendations for facilitating co-operation with regard to the implementation of the Convention on the Physical Protection of Nuclear Material, held at the IAEA Headquarters in Vienna, 18-20 June 1990, Report by the Chairman, Vienna 21 June 1990.
[12] R.J.S. Harry, Practical aspects of nuclear material accountancy and their application to safeguards, in Practical applications of R&D in the field of safeguards. Proceedings of a symposium, sponsored by the European safeguards research and development association, Rome, March 7-8, 1974, p. 97-1M.
[13] R.J.S. Harry, Analysis of R&D Activities in the field of NDA, ESARDA Bulletin, Nummer 21, december 1992 ESARDA en CEC, GCO-Ispra, ISSN 0392-3029, p 5-7-
[14] H. Lefevre on behalf of ESARDA co-ordinators, Analysis of ESARDA partners safeguards R&D activities, 15th symposium on safeguards and nuclear material management, Augustinianum, Vatican City, Roma, Italia, U-13 May 1993-
[15] R.J.S. Harry, The ESARDA approach to international standards, invited paper, Proc. second annual symposium on safeguards and nuclear material management, Edinburgh, Scotland, 26-28 March I98O, ESARDA 11, p. 258, CEC JRC Ispra (1980).
37
[16] R.J.S. Harry, Enrichment standards for gamma-spectrometry, proceedings sixth annual symposium on safeguards and nuclear material management, Venice. Italia, 14-18 May 1984. CEC JRC (1984), ESARDA 17. p. 187-194.
[17] R.J.S. Harry, PIDIE, Plutonium Isotopic Determination Intercomparison Exercise, ECN-RX—90-044. July 1990, INMM, 31st annual meeting proceedings, Los Angeles, California 15-18 July 1990t PP- 216-225-
[l8] V. Bragin, L.D.Y. Ong, Report on the advisory group meeting on safeguards assessment and evaluation methods*, 17-21 May 19931 IAEA-STR-, Vienna, June 1993-
[19] R.J.S. Harry, Splijtstofbewaking en non-proliferatie van kemwapens, Energiespectrum, l6e jaargang, nr. 9. september 1992, pp. 199-205.
[20] R.J.S. Harry, Splijtstofbewaking, controle op vreedzaam gebruik, Natuur en Techniek, 6le jaargang, juni 1993« PP- 456-467-
[21] V.A. Wichers, J.K. Aaldijk, P.A.O. de Betué, R.J.S. Harry, Computer optimized detection geometries for uranium enrichment verification in centrifuge plants, 15th symposium on safeguards and nuclear material management, Augustinianum, Vatican City, Roma, Italia, 11-13 May 1993-
[22] J.K. Aaldijk, Fast neutron measurements at UF6 desublimers in UCN laboratory, ECN-CX—92-078, ECN Petten (1992).
[23] J-K. Aaldijk, W.F. Freudenreich, Determination of UF6-inventory in desublimers by fast neutron measurements, confidential report under review, to be printed.
[24] V.A. Wichers, R.J.S. Harry, Improved Design of TGM Collimators for small diameter Cascade-to-header Pipes, ECN-CX-92-049. ECN Petten (1992).
[25] V.A. Wichers, P.A.C. de Betué, J.K. Aaldijk, "Experimental calibration of improved TGM collimators for small diameter cascade-to-header pipes, to be published.
38
3.3. DEBORA (project id no. 1218)
Introduction
DEBORA is the acronym for DEvelopment of a seal for a BOrehole with RAdioactive waste.
A disposal mine in a salt formation consists of a system of horizontal galleries in which vertical emplacement boreholes are drilled. After the outplacement of the waste the boreholes are sealed, the galleries are backfilled with crushed salt, a system of dams is constructed and the shafts are backfilled and sealed. In the long term performance of this closed and sealed repository the behaviour of the boreholes seal is of crucial importance.
Problem description
To be able to make a good design of the borehole seal the requirements and operating condition must be very clear. There are requirements which are in principle contradictory such as: the seal must be a firm barrier between intruding groundwater and the waste but on the other hand it must not be gastight in order to avoid a high pressure of the gas that might be produced in the borehole.
Project definition
The major objective of the project is to evaluate the present knowledge with respect to relevant loading conditions and constitutive behaviour of borehole seals. In cases were important variables have been identified, model calculation will be performed to estimate the short and long time behaviour of the system borehole-boreholeseal-rocksalt. For the determination of insufficiently known parameters proposals will be made for later investigations [1]. Such investigations are not to be performed in this project.
Project description
The ECN participation in this project consists of: A sensitivity study of the thermo-mechenical loading on the borehole seal. The analysis of the mechanical interaction of the borehole and the gallery. The analysis of the mechanical interaction of the borehole seal and the gallery.
Project control
The project started in 1990 and will be finished in 199^. About 2 manyear will be spent on this project.
Reference.'
[1] Spies, Th., Prij, J., Rothfuchs, T. (1993). Sealing of HAW boreholes in salt formations: Objectives and first results of the DEBORA project. In: PEGASUS project EUR 14816 EN.
39
3.4. EVEREST (project id no. 1219)
Introduction
EVEREST is the acronym for Evaluation des Elements Responsables de 1'equivalent de dose associé å un STockage.
In the field of Management and Storage of Radioactive Waste the European Community plays an important role in the development and harmonisation of methodologies for the evaluation of the post closure safety. In the eighties two leading safety studies have been performed, PAGIS and PACOMA. In these studies different host rocks and different disposal techniques were considered. General trends in the results of these studies are the low to extreme low dose-rates and the very far future in which these exposures might occur.
Problem description
Lue to the long time between the disposal of the waste and the exposure in the biosphere, in most cases more than 100.000 years, the results of the post closure safety studies inevitably have a large uncertainty. To be able to optimize the design one must be able to identify those elements in the geological disposal system which have the largest influence on the resulting exposure.
Project definition
The Communities program EVEREST is aiming to develop a common methodology for the evaluation of the sensitivity and uncertainty in the results of the post closure dose calculations. This methodology has to be applied on different host rocks and different disposal concepts. The methodology has to include model as well as data uncertainty. The project is carried out by teams from Belgium (CEN/SCK), France (ANDRA and IPSN), Spain (ENRESA), Germany (GRS and GSF) and The Netherlands (ECN).
Project description
The ECN participation in EVEREST consists of the following items: Contribute to the harmonisation of the scenarios to be used in the safety study. Perform a probabilistic uncertainty and sensitivity analysis of the subrosion scenario. Perform a deterministic sensitivity study of the groundwater intrusion scenario. Organize a sensitivity study on the model for convergence and compaction of backfilled openings.
Project control
The project started in 1990 and will be finished in 1994. About 1.5 raanyear will be spent on this project.
40
3.5. The 600 m borehole project (project id no. 2254)
Introduction
For the safe disposal of radioactive waste in rock salt formations it is required that deep boreholes will be drilled from galleries with limited height and without the use of drilling liquid. For the safety assessment of such disposal sites it is required that the thermomechanical behaviour of rock salt can be predicted with sufficient accuracy. This means that the short term (elastic) qnd long term (creep) behaviour of the rock salt can be analyzed and that the results of these analyses describe the actual behaviour as measured in situ.
Problem description
Up to now the value of the modulus of elasticity used in mechanical analyses of rock-salt was based on the values as measured in laboratory experiments. In the COSA project [1] it appeared that constitutive parameters based on these laboratory scale experiments do not accurately describe the measured behaviour. Therefore it is required that the constitutive relations will be verified with the aid of in-situ measurements. The parameters in the creep law have to be determined from in-situ measurements of the convergence of openings in the rocksalt. Moreover these convergence measurements can only be correctly used for the determination of the creep parameters once the ambient lithostatic pressure is known. The measurements of convergence performed up to now have been executed at depths where it is not certain that the disturbance of the salt dome due to the large number of excavations is vanished.
Project definition
The objectives of the project are: Prove the feasibility of the dry-drilling of deep, large diameter boreholes. Prove the feasibility of a method of on-line sampling of drilling fines and gas release. Improve the model used for the description of the behaviour of rock salt in so far this behaviour is relevant for the disposal of radioactive waste in rock salt formations.
The project will lead to a dry drilled borehole with a diameter of 600 mm and a drilling depth of 600 m. Equipment will be available for the on-line sampling of the drilling fines and the gas which is released during the drilling operation. Constitutive relations for rock salt [2,3] will be verified and corrected dependent on the modulus of elasticity and the borehole convergence at large depths.
Project description
The project will be executed in the ASSE salt mine near Braunschweig in Germany together with GSF - Porschungszentrum ftir Umwelt und Gesundheit GmbH.
GSF-activities:
41
Drilling of a 600 nun borehole up to a depth of 600 m from a gallery the floor level of which is 750 meter below the surface. Development of a monitoring device which is capable of monitoring the composition of the gas which is released during the drilling operation.
ECN-activities: Measurement of the elastic response of rock-salt with varying pressure. Measurements of the convergence of the borehole at various depths and for a sufficiently long time.
Project control
Up to now about 28 manyear has been spent on the project. For 1993 the manpower is 0.3 manyear.
References
[1] The Community Project COSA: Comparison of Geo-Mechanical Computer codes for salt. Atkins R&D, CEC, EUR IO76O EN, 1986.
[2] Hamilton, L.F.M., Prij. J., Benneker, P.B.J.M.: Evaluation of the experiments with the variable pressure device. ECN-C—92-010.
[3] Heijdra, J.J., Frij, J.: Convergence measurements in a 300 m deep borehole in rock salt. ECN-C—92-016.
42
3.6. PROSA (project id no. 2255)
Introduction
PROSA is the acronym for Probabilistic Safety Assessment of geologically disposed radioactive waste.
Radioactive waste results from applications of radioactive materials, e.g. in industry, medical care, research in several fields and in nuclear power plants. The radioactivity of some categories of waste forms a potential risk for man and environment for a very long time, up to many thousands of years, and consequently a careful isolation is needed. As man made structures can guarantee an isolation only for a period of several hundreds of years, other designs are required. Geological disposal is studied in many countries as a possible solution for this design problem. In stable geological formations repositories have been designed consisting of a multi-barrier-system. In the Netherlands the research is performed in the framework of the OPLA program and is concentrated on disposal concepts in rock salt.
Problem description
Whether the geological disposal system indeed gives the necessary isolation is the subject of safety studies. In these studies the long term performance of the total system including the different barriers has to be assessed. In the studies one has to consider all features events and processes (FEP's) which might influence the long term performance. Not only natural but also human induced FEP's have to be addressed.
Project definition
The first aim of PROSA is the determination of the (human) health risk related to the disposed radioactive waste. A second aim is the identification of safety related characteristics which can be used for further selection purposes. Special attention has to be given to a systematic treatment of uncertainties. PROSA is restricted to post closure safety.
Project description
To reach the aims a methodology has been developed consisting out of several steps:
identification of release scenarios determination of the probabilities of the scenarios determination of the (probabilistic) calculational model determination of the parameters and their distribution function calculation of the doses and risks sensitivity and uncertainty analysis
The project is carried out in close cooperation between ECN, RIVM and RGD. ECN is responsible for the scenario selection, the model and data of the salt compartment, the dose calculations, the uncertainty and sensitivity analysis and the integral project management. RIVM is responsible for the model and data of the groundwater compartment and biosphere compartment. The RGD is responsible for the geological data.
43
Project control
The project started in 1990 and will be finished in 1993- The total number of manyears is 20 from which 2 at the RGD, 8 at the RIVM and 10 at ECN.
References
[I] J. Prij, Veiligheidsaspecten bij de opberging van radioactief afval in steenzout: Een systeemanalystische benadering. Energiespectrum, januari 1987.
[2] J. Prij, Veiligheids Evaluatie van Opbergconcepten in Steenzout (VEOS). Deelrapport 1: Samenvatting en Evaluatie. ECN/RIVM, januari 1989.
[3] J* Prij, L.H. Vons, The role of dams and seals on the release of nuclides due to water intrusion into the repository. Paper for the CEC workshop on closing and sealing of a repository. Braunschweig 22-05/26-05-1989.
[4] J. Prij et al. The role of human intrusion in the Dutch safety study. Proceedings of NEA workshop 186-197, Paris, June 1989.
[5] J. Prij, Safety evaluation of disposal concepts in rock salt. Proceedings of NEA/IAEA/CEC Symposium on the Safety Assessment of Radioactive Waste Repositories, Paris, 9*13 Oct. 1989.
[6] J. Prij, Risico's van opgeborgen radioactief afval. Energiespectrum, februari 1990.
[7] J. Prij et al, Safety evaluation of geological disposal concepts for low and medium-level wastes in rock salt (Pacoma project), EUR 13178 EN, 1991.
[8] J. Prij, On the design of a radioactive waste repository. Thesis University Twente, 14-06-1991, ISBN 90-375~026l-X.
[9] J. Prij, The role of human intrusion in a waste repository in rock rock salt, International Conference Safewaste 93. organized by SFEN, French Nuclear Energy Society, Avignon 13_l8/06/1993. Proceedings Volume 2 490-501.
[10] Jan Prij et al, Convergence and compaction of backfilled openings in rock salt, 3rd Conference on the Mechanical Behaviour of Salt, September 14-16, 1993-
[II] J. Prij et al. PRObabilistic Safety Assessment, Final Report. ECN/RIVM/RGD 1993-
44
3.7. Participation in the HAW-project (project id no. 2253)
Introduction
ECN participates in the test disposal of Highly radioActive Waste (HAH) in the ASSE salt mine in Germany. The HAW-project is sponsored by the European Community. The "safety-criteria for the final disposal of radioactive waste in an underground mine" require that only approved techniques are to be applied for the construction of a repository for radioactive waste. Such approved techniques do not yet exist for the final disposal of heat producing HAW in rock salt formations. It is, therefore, necessary to construct, to operate and to approve the complete technical system of a HAW repository in a full scale pilot test.
The Asse salt mine located close to the town of Wolfenbvittel was selected for the installation of such a full scale pilot test which allows the representative testing of the complete transport and emplacement system for HAW canisters and to investigate the interaction of the HAW canisters with the surrounding rock salt.
Because it is important to keep the results of such a full scale test transferable to a different repository site the following requirements are to be considered in regard of the test at Asse:
the dimensions of the underground test field are to be as similar as possible to those of a real repository the radioactive canisters used should generate a representative surface dose rate and heat power reliable data are to be gained in regard of the consequences of the interaction between the HAW canisters and the host rock the technical components used for transport and emplacement of the radioactive canisters should also be applicable in a future repository without significant modifications.
The entire HAW-project should be performed in four phases:
Phase 1: Development of the test plan and start up of the (1982-1984) development of the transport and emplacement system
Phase 2: Continuation of the development works, procurement (1985-1989) of all technical components, installation of the test
equipment and start up of heater reference tests
Phase 3: Transportation of the radioactive canisters to the (5 years) Apse mine and emplacement in the underground boreholes and
continuous evaluation of the test results
Phase 4: Termination of the in situ test, retrieval and interim (2 years) storage of the radioactive canisters, final evaluation of
all test results, and writing of a final report.
However, due to a decision of the German government in 1992 to stop the project at the end of 199**» phase 3 and 4 will not be executed.
45
Problem description
The basic lay-out of a disposal nine for high level waste in rocksalt in the northern planes of Europe consist of deep dry-drilled holes, drilled from galleries in which the vitrified waste is disposed. The heat production, the irradiation of the vitrified waste effects the rocksalt. The heat production causes convergence of the rocksalt and leads to a pressure build up on the container with the vitrified waste. The irradiation and the heating of the rocksalt will cause gas production. This effects has to be known in order to model these processes for the safety analysis. An other important aspect of the disposal of the containers in the holes is the technical operation itself. This has to be demonstrated before a disposal mine can be operated in a safe way.
Project definition
In order to achieve the two main goals of the project the following objectives are formulated. 1. Development of a transport and emplacement system for HAW canisters. 2. Study of the release (rates and quantities) of water and gas components
as a result of heat release and gamma-radiation and the resulting rise of the gas pressure in the emplacement boreholes.
3. Study of the thermally induced stresses and resulting displacements in the boreholes, galleries and pillars with respect to the validation of computer models.
4. Development and testing of suitable measuring methods for the safety monitoring of a final repository during the construction and operating phases.
The ECN participation is focused on the third objective. The results of the thermal induced stresses and the deformation of the field can be explained satisfactorily by the models derived from the 600 m hole project (2254). The constitutive relations are used in the safety analyses of the PROSA project. The results of the release of volatile component are of a limited nature as no radioactive sources has been placed in the boreholes. A complete disposal system has been developed, tested and licensed by the German mining authority.
N.B.: The german government has decided not to continue the project after 1994. That means that the sources will not be emplaced in the boreholes and consequently the study of the release of brine and gas components as a result of heat and gamma irradiation will have a limited character.
Project description
The test field consists of two parallel running mine galleries at the 800 m level of the Asse-salt mine. In each gallery four holes are drilled. Fig. 1 shows the position and arrangement of the test galleries. It was planned to establish the maximum temperatures of l60*C, 210*C and 230*C at the wall of the six emplacement boreholes (three in gallery A and three in gallery B) with the aid of the radioactive canisters (see Pig. 2.2). Two additional electrically operated heater boreholes are to reach the maximum temperature of 230*C. A borehole spacing of 15 m is provided in the direction of the gallery and a spacing of 19 n> cross-wise.
46
The test galleries can be reached by two access galleries (also for reasons of ventilation) in the west and east. The two emplacement boreholes with identical maximum temperature are different for various reasons according to typcjs A and B (see gallery designation). In the boreholes of type A the annular gap between the liner and the borehole wall is backfilled with a porous medium consisting of ceramic aluminium beads in order to provide access to the water and gas components liberated on the entire heated and irradiated length. This borehole type represents the operation phase of a repository when the annulus between the salt and the canister stack is still open.
A1-B1 230°C electric A2-B2 230°C Cs 137 •»- Sr 90 A3-B3 180°C Cs 137 + Sr 90 A4-B4 160°CSr90
\ Data Acquisition Stationsj
Workshop W8M ""THUE
i r i r r • t - r i
Pig. 1.: Arrangement of the emplacement boreholes
In the case of the type B borehole an unrestricted convergence of the borehole wall and a creeping of the salt onto the liner is permitted due to a lack of backfilling. The effect of a probably almost tight contact of the rock salt with the canister stack on the liberating behaviour of the volatile components can thus be comparatively studied. This borehole type represents the long term phase in a repository after closure of the annulus between the salt and the canister stack.
The already mentioned lining of the emplacement boreholes is necessary to guarantee the permanent retrievability of the radioactive canisters. Therefore, the liner must fulfil the following requirements: - mechanical strength to resist the pressure load resulting from the
creeping rock mass; - gastight barrier against corrosive components released from the rock
mass. During the test duration the stability of the liners is to be controlled continuously. The measuring equipment for this purpose was developed by ECN, More than 1000 measuring points are placed in the HAW-field to monitor the temperature, strains, displacements etc.
17
The main effort of ECN is focused on the thermo-mechanical behaviour of the HAW-field.
Project control
The project started in 1985 and will end at the end of 199*1. An estimation of manyears leads to a number of about 100 over this period.
References
[1] The HAW-project, test disposal of highly radioactive radiation sources in the Asse-saltmine; EUR 14531 EN (1993).
48
3.8. Direct disposal of spent fuel elements (DOS) (project id no. 1212)
Introduction
In the national research programme on the disposal of radioactive waste (OPLA) up to now only waste streams from reprocessing plants and radioactive waste from applications in industry, medical care and research centres has been studied. Direct disposal of spent fuel elements was not included in the programme. The DOS-project is a first limited attempt to complete the OPLA program in the area of the radioactive source term. The OPLA programme deals with repositories in rock salt formations.
Problem description
Whether a geological disposal system gives the necessary isolation is the subject of safety analysis. In the DOS-project only the "flooding" scenario is studied and a comparison is made between the results of the PROSA study.
Project definition
The aim of this project is the determination of the human health risk in relation to the disposal of the radioactive waste in the "flooding" scenario.
Project description
To reach the aim of the DOS-project the following activities have been executed: 1. Quantify the radionuclide inventory. 2. Design of the disposal mine in general terms. 3> Calculate the consequences in terms of maximum individual doses as a
function of time. The project is carried out in close cooperation with the PROSA-project.
Project control
The project started in 1991 and will be finished in 1993- The total number of manyears involved in this project is about 2.5.
References
1. B.J.M. Benneker et al.. Repository lay-out and waste description, to be published.
2. W. Slagter, J. Prij, Results cf consequence analysis, to be published.
49
3.9. Radiation damage in NaCl (project id no. 2252)
Introduction
Predictions of repository containment failure due to sudden release of the stored energy produced by the gamma-rays emitted by radioactive waste canisters have been made in the Netherlands.
Problem description
Low dose rates are more efficient than high dose rates in producing radiation damage in single NaCl crystals [1]. Impurities contained in the NaCl lattice, also enhance the development of radiation damage defects in experiments performed with single crystals artificially doped [2], It is therefore posed that in radioactive waste repositories huge amounts cf stored energy will develop which could threaten the isolation of the waste. These non-quantified and alarming predictions do not take into account either the rocksalt polycrystallinity, nor its rheological properties. When these properties are taken into account, recovery of the rocksalt has to be expected to take place to some extent. Recovery will greatly reduce the stored energy produced by irradiation [3]. The respective importance of these competing processes is generally disregarded.
Project definition
Objective To combine the properties of rocksalt formations with the solid state knowledge on radiation damage in NaCl in order to allow more precise repository behaviour predictions.
Results We pretend to modify the theories and computational models used to predict radiation damage development in a repository through a comparison of these with experimental results obtained in conditions aft near as possible to those of a repository.
Limitations Safety estimations are out of the scope of this project although they are performed as well at the ECN. Analysis of the amount of colloids developed by irradiation and of the amount of brine contained in the samples are performed by our colleagues at the Barcelona University although used in our interpretations.
Project description
Working method Rock salt samples and artificially produced NaCl samples are simultaneously irradiated. Each sample differs from another in only one of the following factors: pressure during irradiation, dose rate, total dose, mineralogical composition, polycrystallinity (or monocrystallinity), original porosity, purity of the NaCl crystals, content of brine, grain size, and original substructural features.
All the experimental results, including those produced elsewhere on the same samples) are compared with each other and with the theoretically
50
predicted yields. Explanations for the obtained results are searched (and found) which gradually modify the theories. The computational models are accordingly modified.
Activities Natural samples are mined from the Asse Nine (Germany) and from the Potasas del Llobregat Mine (Spain). Artificial samples are, except the pure (low OH" content) undeformed NaCl single crystals (purchased from Harshaw), produced by us at the High Pressure and Temperature Laboratory of the Utrecht State University. They consist of cold pressed analytic quality NaCl powder, and synthetic rock salt produced from the same NaCl powder by the application of high hydrostatic pressure (500 bar) and temperature (150°C) during one month.
Before irradiation both natural and artificial samples are machined to closely fit holders specially designed for this purpose [4]. The holders allow irradiation to proceed either at atmospheric pressure or at enhanced pressures up to 200 bar. The samples are included in thick closed gold jackets (also designed by us) to avoids gas escaping during irradiation, and then included in the holders.
The loaded sample holders are placed in specially designed instrumentation assemblages named HEated Irradiation of SAlt instruments (HEISA), This instrumentation allows temperature control during irradiation. Each HEISA is placed amongst spent fuel elements of the High Flux Reactor, in the Gamma Irradiation Facility (GIF). Dose rate is controlled by selective use of the spent fuel elements to be placed around each HEISA. Dose rate is measured at regular intervals, and temperature is constantly recorded. Two soits of irradiation experiments are carried out, in respectively HEISA 1 and HEISA 2.
HEISA 1 uses fresh spent fuel elements and following the (monthly) reactor cycles the dose rate varies from 240 kGy/h to 40 kGy/h each month. Forty natural samples of the same formation (Asse Speisesalz) are simultaneously irradiated at atmospheric pressure and 100° C. Each month a sample is retrieved and substituted by a non irradiated sample. In this way accumulative doses up to 800 MGy have already been obtained.
HEISA 2 uses old spent fuel elements in order to reach a dose rate as constant as possible. 16 samples differing from each other in only one of the factors listed above are simultaneously irradiated at 100°C. Eleven irradiations at 15 kGy/h have taken place up to different total doses, and the experiments are now repeated at 4 kGy/h.
The stored energy developed in each sample during irradiation is measured by Differential Scanning Calorimetry, the micro and substructures developed during irradiation are analyzed by optical microscopy, and the amount of radiation damage defects which ought to have developed following the theories is calculated for each experiment. Jain-Lidiard model modifications according to experimental results are performed [5].
Reports [4] and publications [5.6] are produced which till now have shown that intergranular processes can account for considerable recovery of the damage produced during irradiation if they are allowed time.
51
Project control
The project started in 1986, and is planned to continue up to december 199** # where irradiations in HEISA 1 would probably have yielded a value of saturation of stored energy, and irradiations at h kGy/h in HEISA 2 will almost have been finished.
The manpower is 7*3 manyears/year.
References
[1] Van Opbroek, G. & H.W. Den Hartog "Radiation Damage of NaCl: Dose-rate Effects" J. Phys. Cl8:275 (1985).
[2] Den Hårtog, H.W. "Stralingsschade in NaCl: Eindrapportage REO-3 over fase 1 van het OPLA Onderzoek". Laboratorium voor Vaste Stof Fysica. R.U. Groningen, l*»0p (1988)
[3] Garcia Celma, A.; J.L. Urai & C.J. Spiers "A Laboratory Investigation into the Interaction of Recrystallization and Radiation Damage Effects in Polycrystalline Salt Rocks". European Communities-Commission, Nuclear Science and Technology Series. EUR-11849-EN. ISBN 92-825-9055-0 (1988).
[4] Garcia Celma, A. & H. Donker "Stored Energy in Irradiated Salt Samples". Progress report June 1990-December 1991. ECN-C-92-055.
[5] Soppe, W.J. " Computer Simulation of Radiation Damage in NaCl Using a Kinetic Rate Reaction Model. J. Phys.: Condens Matter 5:3519 (1993).
[6] Garcia Celma, A.; C. De Las Cuevas, P. Teixidor, L. Miralles & H. Donker "On the Possible Continuous Operation of an intergranular Process of Radiation Damage Anneal in Rock Salt Repositories" Proceedings of a symposium on: Geological Disposal of Spent Fuel and High Level and Alpha Bearing Wastes, Antwerp, 19-23 October 1992. International Atomic Energy Agency, Vienna, proceedings Series IAEA-SM-326 (/18): 133-lW. ISBN 92-0-000193-9.
52
3.10. Radio-ecology (project id no. 5241)
Introduction
Knowledge about levels and behaviour of radionuclides in the environment is essential for the assessment of the radiological consequences of releases of radionuclides by industrial installations during normal operation and in accident situations. In addition this knowledge is also required for the correct interpretation of measurements of radioactivity in various compartments in the environment of ECN as well as in the Dutch aquatic and terrestrial environment in general.
Problem description
The problems dealt with in the project are: a. How do radionuclide concentrations in the environment of the Petten
installations relate to real or potential sources of radioactive contamination.
b. Given the relatively high concentrations of non-anthropogenic natural radionuclides in the environment, are anthropogenic radionuclides detectable and can they be related to their sources.
c. Does the study of artificial radionuclides in the aquatic environment near Petten provide information of the transport and dilution processes from remote sources.
The objectives of the project are: a. To develop specific and sensitive analytical methods to determine low
levels of radionuclides in environmental samples, in particular in samples from the marine environment.
b. To test and apply the use of sessile marine organisms as sensitive indicators of the levels of radionuclides in sea water.
c. To obtain time series of radionuclides in sea water and bioi.idicators in order to derive quantitative relations between releases at a remote source and the levels resulting in Dutch coastal waters.
Project description
The project comprises different parts whr'.ch are summarized below. a. Samples from the marine and terrestrial environment are sampled
regularly and measured by gammaspectrometry to determine trends in concentrations. Terrestrial samples (milk, grass, soil) do presently provide the base-line levels as a result of fall out from bomb tests and from the Chernobyl reactor accident. They do provide base-line levels needed to detect and interpret any enhancement as a result of new depositions. A sensitive method to monitor the presently low levels of Cs-137 in milk has been implemented.
b. Sensitive methods to measure artificial radionuclides such as Tc-99i Cs-137 and Sb-125 in sea water have been implemented for the study of the transport of these radionuclides from remote sources. Brown seaweed was shown to be a particularly sensitive indicator for Tc-99 and is used to study the transport of the radionuclide from its point of release at La Hague (Fr). This study is carried out in the framework of a CEC-project.
53
c. Biown seaweeds as bio-indicators are used to map the spatial distributions of artificial radionuclides from nuclear industries and natural radionuclides from non-nuclear (fertilizer) industries in Dutch/Belgian coastal and estuarine waters. Preliminary results do show their usefulness to show levels related to remote sources {artificial radionuclides from the La Hague reprocessing plant) and close-in sources (natural radionuclides from fertilizer industries in Belgium and The Netherlands).
Project control
The CEC Tc-99 project is planned to be finished by the end of 1993 after which the time series on Tc~99 in brown seaweed will be continued with a lower sampling frequency. Publication of results is envisaged for beginning 1991. The bio-indicator program on radioactive contamination from nuclear and non-nuclear sources will be continued in 199^ so as to obtain a firm data base for the definite assessment of source-level relationships and publication of the results.
Further development of analytical methods in and after 1993 will focus on radiochemical analysis of transuranics in environmental materials.
The manpower involved in the project is 2.5 manyear/year.
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3.11. Thermal conductivity of high burn-up UP? fuel (project id no. 1455)
Introduction
The temperature distribution in a fuel pin depends not only on the thermal conductivity of the fuel but also on the distribution of the fission products in the fuel. Thr present tendency to increase the burn-up considerably, makes it necessary to study the temperature distribution in the fuel under these conditions.
Project definition
The objective of the project is to get a better insight in the temperature distribution of U02 fuel with a high burn-up. The influence of the diffusion of plutonium and the main fission products will be described via computer modelling. The existing codes, which use a mixture of empirical, phenomelogical and theoretical models, will be compared. In a later stage, irradiation experiments are planned in the HFR to check the predictions obtained from the computer models. Contacts are foreseen with NEA (Paris) and Halden (Norway).
Project control
The project started August 1, 1993» and the personnel effort will be 1 manyear per year in its theoretical stage.
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3.12. Transmutation of nuclear waste (public information) (project id no. 1092)
Problem description
An "Integral Research Program 1991~199^ for the study of Recycling Actinides and Fission Products", has the intention to evaluate the recycling option of the nuclear waste problem. Besides strategy and scenario studies, reactor physics research, and some technological demonstrations this "Integral Research Program" comprises this public information project.
Project definition
The goal is the creation of a vision regarding the possibilities of nuclear transformation and of the recycling of fuel. This goal is pursued through involvement in related activities, and through contributions to international meetings on the waste problem.
Project description
On request of the Dutch ministry of Economic Affairs a report on the state of the art is being issued, which is intended to be used in discussions of the nuclear energy cycle in the Dutch parliament. This report is being edited in contact with a steering committee, which was established by the Directorate Energy of the ministry of Economic Affairs, as its advisory body on the subject of waste transmutation. Several popular contributions have been given for the Dutch [1,2] and international [3,^] scientific press, and a contribution [5] was given to an international "Meeting on Accelerator-based Transmutation".
Project control
Part of the report "Transmutation of Nuclear Waste" for the ministry of Economic affairs is being rewritten for the national press. New status reports will be made in due time.
The ECN contribution amounts to 1.5 manyears per year.
References
1. K. Abrahams, Recycleren van gebruikte splijtstof, Energiespectrum 15. (1991) 96.
2. H. Gruppelaar, Actiniden recycling, Natuur en Technlek 60 (1992) 603. 3. K. Abrahams, Transmutation of long-lived nuclear waste NEA-Newsletter,
10,8, 1992. 4. H. Gruppelaar and W.M.P. Franken, Recycling of long-lived waste
components in high-flux thermal systems, Nuclear Europe Worldsr.an, June 1993.
5. K. Abrahams, Minimizing the integrated collective radiation dose and the transmutation of long-lived nuclear waste. Contribution to the OECD-NEA Specialists' Meeting on Accelerator-based Transmutation 2*1-26 March, Paul Scherrer Institute, WUrenlingen/Villigen, Switzerland.
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H. ADVANCED REACTORS
4.1. Introduction
In 1989 the Netherlands government decided to postpone the preparation for the building programme of nuclear power reactors. At the same time the interest in the development of advanced and innovative reactors came forward. The promise of higher reliability and inherent safety through simplicity fuels the attraction. Increasingly safety and low impact of severe reactor accidents on the environment hold a high place in the Netherlands public and political discussion.
The issue of developing a new generation of power reactors has therefore also a prominent place in the programme of the unit Nuclear Energy. Funded by the government several projects are underway that address problems and solutions in the development of fission power. Of course there are strong limitations in means and competence and thus a limited amount of areas can be covered at ECN.
Core of the program is the PINC project (Programme to enhance nuclear competence). This government funded project carried out by ECN, KEMA, NUCON and IRI aims at the improvement of their competence in the nuclear field. The emphasis is this program is on reactor development, improvement of tools and the promotion and education and training in nuclear science and engineering. Through this programme the Unit Nuclear Energy of ECN supports the analysis of the SBWR design. Further the development of innovative, high temperature gas cooled reactors is enhanced for selected topics. The total PINC effort in the Netherlands amounts to 27 manyear per year. ECN Nuclear Energy takes about one third of that. The cooperation with the partners in and outside the Netherlands gives much more leverage to the Unit Nuclear Energy than nine manyear per year ECN spends on it.
The other projects can be divided in two categories: - Neutron physics - Materials science and technology The neutron physics activities consist of three related topics. The two basic topics are nuclear data library evaluation and maintenance in the frame of NEA and the strengthening of the expertise and experience in the field of whole core reactor physics calculations including thermal hydraulic effects. The third topic is the analysis of the cores of advanced or innovative reactors such as the HTR-M. Transient behaviour and the calculation of reactivity coefficients are main subjects.
The materials science and technology in support of fission reactors has the HFR and the hot cell facilities to conduct experimental work on advanced alloys and ditto weldments. Post-irradiation investigations are aimed at the improvement of fracture mechanics as well as creep fatigue properties. The development of some of these alloys is in the phase of generating engineering data of life conditions in support of design and safety analysis of non-replaceable components.
The topics addressed are by no means the only few that need investigation or development. The areas for projects of the Unit Nuclear Energy have been selected with the interest for the Netherlands situation and the opportunities available. The programme on advanced reactors will be redirected in the light of changing interest of the main funder of the work
S7
and the opportunities obtair»d to cooperate with parties interested in similar subjects. The effecciveness of the present program hinges on cooperation and partnership. The present situation in this regard is much more effective than a closed programme developed and carried out by the Unit Nuclear Energy only would ever be.
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A.2. Reactivity effects in advanced reactors (project id no. 1090)
Introduction
The Tsjernobyl accident has caused an increasing interest all over the world in the construction of advanced nuclear reactors with enhanced passive safety features. If this safety is based mainly on the laws of nature, with a minimum dependency on active safety systems, the reactors are called "inherently safe". The main object is to construct the reactor in such a way that, in case of a power excursion, negative feedback mechanisms are active that bring the reactor back into a stable and safe situation, without needing the intervention of an operator or of active or passive systems with a certain chance of failure. Such reactors have a so-called "walk away" safety: the reactor can be left alone for a certain time (a number of days, weeks or even months) without the occurrence of damage to reactor and surrounding.
In the literature a large number of concepts have been presented; it is not a simple task to review the concerned safety characteristics and to make a choice about which reactor type to prefer. In the reactor physics analysis of these concepts an important role is played by the reactivity coefficients; a study of the value of these coefficients, as well during normal operation as also under accident conditions is of great importance.
Project definition
The main objective of the project is to study the physics properties of the reactor core, especially with respect to reactivity control, stability and transient behaviour. One of the purposes is the generation of a link between reactor physics and thermal-hydraulics. Also studies with respect to sensitivity and uncertainty will be part of the project. These can be used to derive the change in the multiplication factor resulting from a small change in a reactor parameter.
The project is aimed specifically on the study of the reactor piysics component in the framework of safety research. For the reactor type considered first the behaviour of the reactivity coefficients will be examined; later on these coefficients will be used in simple dynamics calculations. In this way the properties of reactivity coefficients, playing a central role in safety research, can be studied extensively. Besides the aforementioned research the gathered knowledge can be used to answer questions from government and others, also if existing reactors are concerned.
Project description
The main working method for this project is to gather insight in the subject as well by means of literature study as also by means of calculations. The calculations are performed with a code package developed and tested in cooperation with the PINC programme. One of the means to validate and verify the codes is the calculation of benchmark problems; work on this will be continued; where necessary, this will be Uone in international cooperation.
59
For a selected reactor type first introductory calculations will be done in order to get familiarised with the reactor and to acquire experience in how to model the reactor core for physics calculations. Generally these calculations will be for the stationary reactor; from geometry and fresh material composition the multiplication factor and flux- and power distribution will be calculated. Later on, in order to find a realistic material composition, burnup calculations will be needed; also a correct modelling of the control rods will need much attention. In a second phase reactivity coefficients (for fuel temperature, temperature and density of coolant and/or moderator) will be calculated; these coefficients will be used for dynasties calculations, aimed at the neutron behaviour in the core, and can also be used as input data for thermal-hydraulics studies, performed elsewhere in the Unit.
The studies on sensitivity and uncertainty will be performed with the code package used extensively in the framework of the European fusion technology programme. Modifications will be made in the codes in order to get a wide field of application.
Notes In the past few years a number of separate studies did arise from this project: analysis of a control rod drop accident for a BWR, calculation of Ooppler coefficient for the same reactor, study with respect to the positive coolant void reactivity of the CANDU reactor. There is a close relationship with the project 'Dynamics of power reactors' where the same kind of calculations are performed with a different package of codes.
Project control
Time period: 1991 - 199^ Manpower : 2 manyear per year
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4.3. Nuclear data for fission reactors (project id no. 5**06)
Introduction
Nuclear data play an important role in the reactor physics calculations for fission reactors. This project aims for extensions and improvements in quality, completeness and understanding of nuclear data files. These files are used in calculations in connection with operation, safety and economy of fission reactors and therefore the most recent and well tested data are required. In particular for the new generation of fast power reactors, such as the PRISM-reactor, improvements of nuclear data for fission products and structural materials are necessary.
This project closely collaborates with two other nuclear data projects in our group, which are executed within the European Fusion Programme, namely evaluations of cross sections relevant for the design of fusion reactors such as NET or ITER. These data libraries are the European Fusion File -EFF- and the European Activation File -EAF.
Project definition
The aim of this project is to contribute, in an international frame of NEA, to evaluations of cross sections, in particular for fission products and reactions relevant for neutron metrology. ECN is represented in the NEANSC "International Working Party on Evaluation Coordination" and in the "Scientific Coordination Group" of the JEF project (Joint Evaluated File). ECN is responsible within in the JEF-2 project for the fission-product subfile and for some of the construction materials (in close collaboration with the EFF project).
Another task is to assist in the processing of the JEF-2 data file into a multigroup structure file (performed at ECN within another project), if any problems concerning the physics and quality of data arise. The above-mentioned working library will be applied in all reactor calculations at ECN.
In order to fulfil this task it is also necessary to devote a part of the effort to follow and actively be involved in the maintenance and development of nuclear model codes.
Project description
The JEF-2 data file The main activity is test and improvement of the cross sections for the fission products [1] (in collaboration with the NEA Data Bank in Paris). The work has been recently focused on cross sections for long-lived fission products [2] in connection with studies concerning the nuclear transmutation of nuclear waste. This programme forms an important activity at ECN.
Nuclear-model code developments Multi-step direct reaction model In the last four years (1989*1992) the leading-particle statistics description for the multi-step direct reaction (MSD) has been developed and compared with other existing theories. This work resulted in a PhD graduation at the University of Groningen [3], The application of the MSD is in
61
particular important for calculation of double-differential neutron-emission distributions in the energy range from 10 to 100 MeV. The results of this research have been included in a computer code that calculates MSD cross sections.
Inelastic scattering in fission products Until recently inelastic scattering did not gain the proper attention in fission-product cross section evaluations. In many existing evaluations global spherical-optical models have been used, neglecting direct and preequilibrium effects. Present work is focused on improving anomalously large inelastic-scattering cross sections observed at low energies in the even-mass nuclei near mass A = 100, using the coupled-channel calculations
Project planning and control
This project is funded from "basic subsidy" and is currently planned for the period of 1991*1995• The volume of the available budget is reconsidered every year. In the last few years the allocated manpower amounted to about 1.5 manyear and was decreased to 0.7 for 1993-
Recent references
[1] H.A.J, van der Kamp and H. Gruppelaar, Revision of JEF-2 Evaluations for the Lcng-lived Fission Products 1-129 and Tc-99, ECN-RX—92—023.
[2] H.A.J, van der Kamp, The JEF-1 and JEF-2 Fission Product Cross-Section Files, ECN-RX—92-021.
[3] A.J. Koning, Multi-Step Direct Reactions, PhD Thesis, University of Groningen (ECN-R—92-004) and partially published in several international journals.
[4] H. Gruppelaar and A. Hogenbirk, Direct and Preequilibrium Effects in the Fission-Product Mass Range, presented at the NEANSC Specialists' Meeting on Fission Product Nuclear Data, 25~27 May 1992, JAERI, Japan and ECN-RX--92-040.
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k.k. HTB power transients and reactivity (project id no. 1489)
Introduction
As gas cooled High Temperature Reactors (HTR's) have still a number of advantages the Dutch PINC group (KEMA, GKN. NUCON. IRI and ECN) has included an exploratory study on the further development of the German HTR-Module.
Strong points of the high temperature gas cooled reactor HTR-M are: use of inert helium as coolant strong negative temperature coefficient during normal operation strong retention of all fission products inside the coated fuel particles up to high temperature even in case of loss-of-coolant accidents no need of active cooling of decay heat due to sufficient passive heat removal from the small core to the environment.
A weak point of the HTR-M is its behaviour after ingress of water or s team in the reactor core. Further the economic perspective is still unfavourable. In the frame of a cooperative agreement by Siemens/ABB, CEA (F), Framatome (F), ENEA (E) and the Dutch PINC consortium a study is undertaken to arrive at lower costs for the HTR-M.
However, in the Netherlands the nuclear knowledge on gas cooled reactors is far behind the knowledge on light water reactors and fast breeder reactors.
Therefore, as the adaptations in the design accompany with changes in the reactor physics behaviour of the reactor core and its safety related characteristics, the underlying project is being initiated to make up the arrears in reactor physics.
Project definition
The main objective of this project is to study the reactorphysics properties of the reactor core of the HTR-M, especially with respect to the safety related transient behaviour and the effect hereupon of the reactivity coefficients.
Project description
In the first phase of the study the calculation of the double heterogeneity of the spherical fuel elements has to be performed. This heterogeneity is due to the distribution of the large number of small fuel particles inside the graphite matrix of the spherical fuel elements. The second heterogeneity comes from the pile up of the many fuel elements to a complete core. Surveying calculations will be performed first assuming homogenized densities inside the fuel elements and throughout the complete core. In the second phase the calculation of reactivity coefficients of water ingress and temperature increase will be performed. Finally some reactor physical effects from the enlargement of the reactor core will be investigated.
63
During the execution of this task intercomparison of the calculational results with results from other institutes on a number of well defined computational benchmarks is foreseen.
Project control
planning : 1993-199** manpower : 1.5 manyear
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4.5. Fracture resistance and creep fatigue damage of irradiated austenitic stainless steel plate and weld materials (project id no. 1426/1432)
Problem description
The austenitic stainless steels are frequently applied as construction material for nuclear components such as PWR internals and vessels and internal structures of fast reactors. Until very recently these steels have also been considered as the primary candidate construction materials for fusion systems.
Depending on the particular applications the construction materials are exposed to neutron damage dose levels ranging from less than 0.1 dpa at 850 K for the above core structures of fast reactors to about 50 dpa at 550 K for PWR internals. This affects respectively the creep-fatigue resistance and the fracture resistance, particularly for conventional weld materials.
Data on the post-irradiation properties, in terms of new fracture mechanics parameters and creep-fatigue resistance parameters, are required for the recent structural integrity assessment procedures to improve the safety of the operating fission reactors and for the design analyses of advanced reactors. Further the improvement of the post-irradiation properties of welds, by advanced welding techniques such as electron-beam welding and automated plasma welding, has to be verified. It is particularly for this type of work that ECN has expertise and special facilities.
Project definition
The objective of the project is the measurement of post-irradiation properties of austenitic stainless steel plate and weld materials, in terms of new fracture mechanics parameters and creep-fatigue damage parameters. The dose levels are limited to 10 dpa ultimately and long-duration tests are limited about 10000 hrs time-to-rupture, both limitations mainly because of time restrictions.
The project provides engineering data on representative end-of-life conditions of the construction material of nuclear components of austenitic stainless steel. The data can be applied in structural integrity analyses to improve the safety of operating fission reactors and in design analyses of advanced reactors. The work is complementary to other ECN projects for fusion.
Project description
The project was started in I9B7, in the scope of a European cooperative advanced reactor program. The project is separated into two parts, one part (1426) concerns the degradation of the fracture resistance due to radiation hardening and embrittlement at intermediate temperatures (below 750 K) and the other part (1432) deals with the degradation of the creep-fatigue properties due to helium embrittlement at elevated temperatures (above 750 X).
The first part (1426) consists of a series of HFR irradiations and post-irradiation fracture mechanics experiments. A total number of 8 capsules with tensile specimens and small compact-tension specimens are irradiated
65
at 700 K to neutron damage dose levels ranging from 0.4 dpa up to 10 dpa. The irradiation period of the high dose {10 dpa) irradiations is about 2.5 year. The final irradiations will be completed in the first half of 1993«
The second part (1432) consists of a series of 8 successive HFR irradiation experiments (each with 3 sample holders) with creep and low cycle fatigue specimens. The irradiations last one cycle in an outer-core position. The neutron damage dose levels are about 0.1 dpa at 823 K. The final irradiations will be started in the second half of 1993-
Project control
The 'low temperature* part 1426 was started with the 10 dpa irradiation experiment at the e~i of 1986. The 'elevated temperature' part 1432 was started mid 1989- The duration of the project can be limited to a reasonable period of 10 years by strictly applying the restrictions on the ultimate irradiation dose level and duration of the creep tests.
Part 1426 was almost completed at the end of 1992. The irradiation of the final sample holder with tensile specimens and the post-irradiation testing of the samples will be completed in 1993-
Part 1432 was started in 1989 with the irradiation of 1 sample holder with 15 creep specimens. The other sample holcer followed successively until completion in the third quarter of 1993- Creep testing of the irradiated specimens was started in 1991 and in the year thereafter the low cycle fatigue experiments were started. The low cycle fatigue machine with the test equipment will be installed in a hot cell in 1993« Low cycle fatigue testing of irradiated specimens is planned to start in the third quarter of 1993.
The testing commitment for 1993 is 2.4 manyear.
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4.6. Programma to enhance nuclear competence (PINC) (project id no. 1107)
Introduction
The building of nuclear power reactors was postponed by the Netherlands government in 1989• In order to uphold the nuclear knowledge level at the same time a government funded programme called PINC was launched. Industry and research establishments cooperate in the programme to maintain and enhance their competence in the nuclear field. This approach is felt necessary to anticipate building of nuclear plants when the time has come and to closely follow developments in the nuclear industry elsewhere.
Problem description
The PINC programme is aimed at the enhancement of nuclear competence in the Netherlands research institutes and industries. In the programme four organizations participate: NUCON (Industry), IRI (University) and KEHA and ECN (research institutes). Several areas have been identified as a fertile basis for cooperation of Netherlands organizations both in and outside the Kingdom. Especially the development of advanced and innovative power reactors is prominently built into the programme, but education of young scientists and engineers plays a major role in enhancing the nuclear competence in the Netherlands for the coming years.
Project definition
The main objectives of the programme as implemented at ECN Petten are: Enhance competence by participating in selected subjects on reactor development. Improve tools for analysis. Improve and support education of young scientists into the nuclear field. Study reactor developments elsewhere in general.
The intended main results of the project are: Increased level of nuclear competence in existing workforce through dedicated participation in reactor developments and investments in fresh persons through improved education. Provide a knowledge base for decisions on nuclear power use in the Netherlands from both the r.idustry and research institutes point of view.
Project description
The project has been divided into four subprojects: Nuclear fission power generation developments. This part of the programme is devoted to studies of innovative reactor designs with improved inherent safety such as the HTR-M and MHTGR. The work is performed in close cooperation with the partners. Participation in analyses of designs of advanced reactors such as SBWR and APWR. With General Electric detailed subtasks have been defined, where ECN contributes to the analysis of the SBWR. In the field of neutron transport analysis the existing software has been updated miå greatly improved. Further commercial codes have been procured and installed for the analysis of power and test reactor cores (PANTHER).
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The education of nuclear engineers and scientists has been improved through the development of workshops, courses and cooperation with IRI (Delft University). Also theses are prepared in the framework of PINC to enhance competence.
Project control
The present project will be finished at the end of 199^ after five years at a level of nine manyear per year. At present negotiations with the Ministry of Economic Affairs are underway for the period of 199^-1996. Specific topics in preparation are:
Continuation of SBWR contribution. Participation in the development of a European advanced pressurized water reactor to be built by a consortium lead by SIEMENS and Framatome. Participation in the development of HTR-M or MHTGR. Continuation of educational part of the program. Improvement of software tools.
References
[1] J. Li, J. Slobben, H. Gruppelaar. TRX benchmark test of the PASC-3 code system and JEF-1 data library. NFA-LWR-92-01.
[2] B.J. Pijlgroms, J. Oppe, H. Oudshoorn and J. Slobben. Simplified modelling and code usage in the PASC-3 code system by the introduction of a programming environment. Contribution to "International Conference on Nuclear Criticality Safety - ICNC91", Oxford, U.K., 9-13 September 1991. Preprint ECN-RX-9I-O63.
[3] B.J. Pijlgroms, J. Oppe, H. Oudshoorn. The PASC-3 Code System and the UNIPASC Environment. Contribution to "Seminar on SCALE-4 and Related Modular Systems. OECD NEA Data Bank, Saclay, 17-19 September 1991. Preprint ECN-RX-91-O78.
[4] J. Bultman, T. Wu and C.L. Cockey. Reactor physics comparison of metal fueled and oxide fueled AMLR cores, presented at the GLOBAL '93 conference in Seattle, September 1993«
[5] Liu Guisheng, A.J. Janssen and J. Slobben. Some results of fast reactor benchmark testing calculations. NFA-ENG-91-OI.
[6] R.C.L. van der Stad, H. Gruppelaar, J. Oppe. A JEF-1 based 219-group neutron cross-section library: User manual. ECN-I-92-039.
[7] Cai Chonghai, H. Gruppelaar, J. Oppe, R.C.L. van der Stad. A JEF-1 based 219 group neutron cross section library: Plots of total elastic, fission and capture cross sections. ECN-I-92-O36.
[8] W.J.M. de Kruijf. Reaktiviteit van sphere-pac splijtstof met bijmengsel. NFA-LWR-91-02.
[9] J. Bultman. Burning actinides in the HFR. NFA-Act-91-01.
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[10] B.J. Pijlgroms, J. Oppe, H. Oudshoorn. The PA0C-3 Code System and the UNIPASC Environment. Contribution to "Seminar on SCALE-4 and Related Modular Systems. OECD NEA Data Bank, Saclay, 17-19 September 1991. Preprint ECN-RX—91-078.
[11] B.J. Pijlgroms. Criticality analysis of the HFR fresh-fuel storage facility. Report NFA-HFR-TR-91-07.
[12] Wang Yaoqing, J.L. Kloosterman and H. Gruppelaar. Benchmarking the EJ2-XMAS library by the "Rowlands" benchmark". NFA-LWR-93-04, June 1993. in preparation.
[13] R.C.L. van der Stad and H.L. Oudshoorn. Verification and validation of the SCALE-4.1 system and EJ2-XMAS library". ECN-I--93"..., in preparation.
[14] S. Spoelstra. SBWR Plant Data for MELCOR. ECN-CX—92-095.
[15] J.W.A.M. Kevenaar. High Temperature Material Properties. ECN-CX--92-097.
[16] S. Spoelstra. MELCOR Severe Accident Analyses for the SBWR. ECN-CX—92-106.
[17J J.W.A.M. Kevenaar and M."..C. van Kranenburg. SBWR Core Support Plate Integrity under Severe Accident Conditions. ECN-CX--92-IO9.
[18] J. Hart. SBWR Plant Data for SCDAP/RELAP5. ECN-CX--92-111.
[19] E.J. Velema and S. Spoelstra. MELCOR Sensitivity Analysis on SBWR Core Melt Progression for the SBWR. ECN-CX--93-004.
[20] J. Hart. SCDAP/RELAP5 Severe Accident Analyses for the SBWR. ECN-CX—93~009.
[21] J. Hart and S. Spoelstra. Severe Accident Analyses for the SBWR -Comparison between MAAP, MELCOR and SCDAP/RELAP5. ECN-CX—93-012.
\_2Z\ J. Hart and S. Spoelstra. SBWR Core Melt Progression. ECN-CX--93-013.
[23] M.A.C. van Kranenburg and J.W.A.M. Kevenaar. SBWR Reactor Pressure Vessel Lower Head Integrity under Severe Accident Conditions. ECN-CX—93-014.
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4.7. Dynamics of power reactors (project id no. 1376)
Introduction
Knowledge and experience has been built up in the field of static reactor physics calculations, with emphasis on the range of calculations from basic nuclear data, as are given in e.g. the JEF nuclear data library, down to assembly calculations. Besides, also experience is present in the field of static reactor physics calculations on whole reactor cores, using Monte-Carlo, finite difference neutron diffusion and finite difference neutron transport calculations.
The code systems in use for performing these calculations encompass NJOY (data extraction from JEF/ENDF-libraries, generation of so-called AMPX-Master library), PASC-3 (an extended version of SCALE-3; resonance treatment and cell/assembly calculations), KENO-Va (Monte Carlo), MCNP (Monte Carlo), DOT (2-dimensional neutron transport) and CITATION (3-dimensional neutron diffusion). In 1992 the WIMS suite of codes {LWRWIMS and WIMS-E) was acquired from AEA Technology (Winfx-ith, TJK), together with the associated 69-group data libraries. Work is in progress at ECN to develop a code to generate a WIMS-formatted library from an AMPX-Master nuclear data library in use with the PASC-3 system, which will enable us to use the "same" basic data in PASC-3 and WIMS-E/LWRWIMS calculations.
Also there is some experience in combined reactor physics and thermal hydraulics calculations -static and transient- using the PRORIA code. This code, which combines two-dimensional (R-Z) neutron diffusion with LWR thermal hydraulics, was adapted in order to be able to participate in the NEACRP 3-D LWR core transient benchmark [1].
Problem description
It is desirable to extend the expertise in the field of, both static and transient, 3"dimensional, whole-core, combined reactor physics and thermal hydraulics calculations. Within the Business Unit ECN-Nuclear Energy this will bridge the "knowledge gap", existing between the expertise in the field of "pure" reactor physics, and the expertise in the field of thermal hydraulics and "balance of plant" calculations.
One of the options to acquire such expertise in the field of combined 3-dimensional reactor physics and thermal hydraulics calculations is the introduction of the -commercial- code PANTHER from Nuclear Electric (UK). This code performed very well in the NEACRP 3"D LWR core transient benchmark [1], mentioned above. The codes LWRWIMS and WIMS-E are capable of generating the necessary PANTHER nuclear data base. Together with the capability -mentioned above- to generate a WIMS-formatted library from an AMPX-Master library, this will give us the opportunity to cover the entire range from basic nuclear data, through cell and assembly calculations, down to whole-core, combined reactor physics and thermal hydraulics calculations.
However, the use of the PANTHER code also has some disadvantages. The "black box" nature of a commercially available code limits the possibilities to gain insight in the interior of the code and also the possibilities to modify the code in order to tailor it according to our
70
specific needs. Furthermore, PANTHER is a code in development. At the moment the code is not suited for BWR transient analysis. Also, PANTHER is a 2-group code. In principle it can be extended to multigroup, but this is not considered by the supplier of the code.
In order to alleviate these disadvantages, it is desirable to study other codes as well, like FX2-TH (2-dimensional, time-dependent, multigroup reactor physics and thermal hydraulics; "improved quasistatic method") and PRORIA (2-dimensional, time-dependent, 2-group reactor physics and thermal hydraulics). Both codes can be obtained from the NEA data bank and, consequently, for both codes a source code is available. As mentioned above, within the NFA-group there already exists some experience in the use of the PRORIA code. The FX2-TH code offers the possibility of cooperation with the Interfaculty Reactor Institute of the Delft University of Technology, where this code is in use.
Project definition
The main objective of this project is to extend the expertise of the NFA-group towards static and time-dependent, whole-core, reactor physics and thermal hydraulics calculations for, at first instance, PWR-type reactors. This means the coverage of the entire range of calculations from basic nuclear data downward, including the generation of the necessary nuclear database.
Project description
Experience will be build up in the use of the code PANTHER by performing calculations on ever more complicated (benchmark) problems. An important part of this familiarization phase will be the participation in the sequel of the NEACRP 3-D LWR core transient benchmark, scheduled for 1993/199**-Another part is the study of other codes -PRORIA and FX2-TH-, in order to gain insight in the interior of this type of codes.
After the familiarization phase (beginning of 199*0, it is intended to participate in calculations concerning existing and new PWR designs, like the Borssele PWR.
Project control
Period : 1993-1994 Manpower : 1.6 manyear per year
Note
There exists a close relationship with project "Reactivity effects in advanced reactors" where the same kind of calculations are performed with a different code package.
References
[1] Finneman, H., Bauer, H., Galati, A. & Martinelli, R.t "Results of LWR core transient benchmarks", Proc. M&C+SNA '93> Karlsruhe, Germany, April 19-23, 1993.
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4.8. Creep experiments structural steels (project id no. 1422)
Introduction
ECN is involved in the generation of a materials data base for materials in the primary circuit of European liquid metal fast breeders (MFBR). Mainly the austenitic stainless steels 30^ and 316 with their weldments are investigated. The weldments are of different types such as tungsten inert gas, TIG, plasma weldments and electron beam, EB, welds. The radiation effect determination has a prominent place in the ECN experiments, where also the High Flux Reactor, HFR, contributes. The data and results of studies on fracture mechanisms are used by designers and licensing authorities. In the licensing process of the SNR-300, an LMFBR built in Kalkar by a German/Belgian/Netherlands consortium, these data have been used. In the design of the European Fast Reactor« LFR, the data have been applied too.
Problem description
The primary circuit of a LMFBR with components such as shields reactor vessel, support plate and instrumentation grid is exposed to neutron radiation. In the design the radiation effect on mechanical properties must be taken into account. This effect can be determined by comparing the mechanical properties without radiation and after relevant neutron radiation doses. In the present project the creep properties of DIN 1.4948, a German derivative of type 304 SS, are determined in thermal control condition and after neutron radiation to determine the damage factor on creep and creep rupture properties.
Project definition
The aim of the project is to determine the neutron radiation effect on DIN 1.4948 plate and weldments creep properties. The result of the project is the reduction factors on creep strength, strain rate and ductility brought about by neutron radiation damage levels relevant for primary structures such as reactor vessel and internals. This project does therefore not provide data on core materials such as fuel cladding and core wrappers. The latter materials are exposed to neutron fluences much higher than those on the vessel. These materials are also at regular intervals removed from the core to make room for a fresh fuel inventory. Components of the primary circuit that are not replaceable in an economic way are the subject of this project.
Project description
First a set of heats to be investigated is selected together with the weld type. In this case one reference heat with one reference TIG weld was selected for long term post-irradiation testing. Prior to testing the material was irradiated in the HFR at the test temperature of 825 K, the operating temperature of the circuit. After radiation plates and weldments are tested to rupture times up to 50,000 hrs. The results are compared with identical material tested under thermal control condition and the radiation damage factor is determined.
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Project control
The bulk of the project was compiled before 1990' A few long lasting tests are proceeding. The last test will be switched off in 1993 said the work is finished in 1994. Of the ongoing work with other newly developed type 316 weldments are proceeding in other projects (1.432). On a yearly basis the work amounts to 0.2 manyear.
Reference
B. van der Schaaf, J. Schirra, Long term creep properties, including irradiation effects, of DIN 1.4948 steel from SNR-300 primary components, Fast Breeder Reactors: Experience and Trends, Proc. of a Symposium, IAEA, Wien, I986. Vol. I, pp 423-433-
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5. ENGINE PROGRAMME
5.1. Introduction
ENGINE, acronym of ENergy Generation In the Natural Environment, is a long-term R&D programme of ECN with the objective to contribute to the development of future energy supply systems, which have to be clean, safe and renewable. Thinking is focused to the characteristics of an energy supply system within the Netherlands in the year, say, 2030. It is believed that such a system will consist of a "mix" of renewable (solar, wind, biomass), fossil fuel and nuclear energy. The main guiding themes of ENGINE can be compacted by: inexpensive renewable energy, clean fossil fuel energy and safe nuclear energy.
ENGINE has four components: Renewable energy. Development of photovoltaic cells with higher conversion efficiency and inexpensive manufacture techniques has been selected as the main objective up till now. Fossil fuel. The main objective is to study combustion processes with strongly decreased emission of sulphur, nitrogen and carbon compounds. A possible route is to turn to hydrogen technology in which field a number of experimental projects are executed. Nuclear energy. The goal is to contribute to innovative technological aspects of fission energy, especially directed to safety characteristics and sustainability. Societal aspects. Integrated studies are performed as guideline and support for the R&D in the other ENGINE components. Three research themes have been chosen which play an important role in making operational the strive towards sustainability: life style and energy demands, management of risks and integral management of the material life cycle.
The ENGINE programme has to be regarded as complementary to already running major ECN activities, which, by objective, would fit perfectly within ENGINE but are excluded from the programme for organization reasons (different financing, already embedded in other programmes). Examples are the programmes on wind energy, the development of fuel cells and the fusion programme.
The nuclear component of ENGINE is focused to development aspects of the third generation (innovative) nuclear reactors; the second generation (evolutionary) reactors is studied within the frame of other programmes. The development of innovative reactors is characterized by an increasing number of inherent safety characteristics and by efficient use of fuel and construction materials. ENGINE contributes to this development by performing studies as well as experimental research in the fields of:
advanced fuels and fuel cycles; low-activation construction materials for fission reactors; transmutation of actinides and long lived fission products as a possible contribution to solve the nuclear waste problem.
Besides general analyses and scenario studies on perspectives for each of these fields also experimental research is performed. In this respect it is
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important to mention that ECN has free access to the High Flux Material Testing Roactor (HFR) in Petten and infrastructure is available for chemical analysis and post irradiation examination. The value of the experimental work is increased if the work can be embedded within international collaborations. This has been realized for the transmutation project; serious attempts are being made for collaborations on fuel development and construction materials.
ENGINE-nuclear (in total about 9 manyears in 1993) consists of the following projects:
Materials: - Low-activation construction materials.
Fuel and fuel cycle: - Fuels for innovative reactors. - Advanced fuel cycles and non-proliferation.
Recycling of actinides and fission products (RAS): - Studies on transmutation of nuclear waste. - Reactor physics aspects of PRISM. - Technological research on RAS (experiments)
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5.2. Low-activation construction materials (project id no. 1490)
Problem description The conventional nuclear construction materials become radioactive after exposure of neutrons. This causes limitations for maintenance and repair of the reactor components. Moreover, the remnants of a reactor have to be stored over a long period of time after the decommissioning. By selecting non-conventional materials with tailored chemical compositions, the detrimental radioactivity can be reduced. The development of alternative construction material? with low-activation (LA) characteristics, requires, among others, the verification of the resistance against degradation of material properties by neutron irradiation. Particular in this field ECN has special expertise and facilities to contribute to the development of LA-steels.
Project definition
The objective of the project is to verify the potential benefit of ferritic-martensitic stainless steels, as alternative for the austenitic stainless steel construction materials, which are among the conventional construction steels most prone to radio-activation. The project contributes to the long-term development of LA construction materials with benefit to the design of advanced fission reactors and fusion systems. Development stages such as the production development and the verification of the workability and the fabricability are excluded. The results in the form of engineering data on the mechanical properties will form a reference data base for further development of LA construction materials. The data can he applied in the design and safety analyses of nuclear components fabricated from commercial and new ferritic-martensitic stainless steels.
Project description
The project consists of three successive parts which deal with (a) a general orientation, (b) irradiation and testing of selected commercial ferritic-martensitic stainless steels with reduced-activation (RA) characteristics, and (c) irradiation and testing of a first series of potential LA-steels. Part (a) consists primarily of a literature study on the effects of chemical composition and heat treatment on the microstrueture and mechanical properties of 9-12% Cr steels. Part (b) consists of a series of irradiations at 300*C with selected commercial RA-steels from France, the UK, and the US. Low cycle fatigue specimens, compact tension specimen for fracture toughness tests, and tensile specimens will be irradiated up to a relatively low irradiation damage level of 2.5 dpa ultimately. Part (c) consists of a similar series of irradiations with selected LA-steels from Europe and the US. It is intended also to include a Japanese LA-steel.
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Project control
Part (a) will be completed in 1993* Part (b) is started in 1993. The irradiations will be executed in 1993 and 199*1 and the post-irradiation testing will be performed in 199** and 1995« Part (c) will be started in 199*1. The irradiations are planned for 199*1/1995 and tne post-irradiation testing will be performed in 1995/1996.
Manpower: 2 æanyear per year.
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5.3. Fuels for innovative reactors (project id no. 1491)
Problem description
Development of new fuels and fuel cycles may enhance reactor safety characteristics and may contribute to the solution of the nuclear waste problem by improved reprocessing and partitioning techniques. Such a development is going on in the US. where Argonne National Laboratories (ANL), in the frame of the Advanced Liquid Metal Reactor (ALMR) Program, is developing metallic fuel and pyro-processing for the PRISM reactor.
Project definition
ECN has explored the opportunities of an involvement in the ALMR metallic fuel development. Cooperation between ANL and ECN has been proposed on the following items:
Metallic fuel: phase diagram calculations and experimental work, including fuel irradiations in the HFR, to study the fuel characteristics and fission product behaviour under accident conditions. Electrorefining process: experimental determination of thermodynamic properties of actinide halides in molten salt mixtures. Detachment of a scientist at ANL.
Despite frequent contacts with DOE, ANL and GE such a collaboration could not be established yet due to uncertainties with respect to the future of this project in the USA. As a consequence ECN is presently redefining its effort in this respect.
Project control
Total manpower 0,8 manyear (1993)«
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5.4. Advanced fuel cycles and non-proliferation (project id no. 1492)
Project 1^92 consists of two parts: study of advanced fuel cycles, and proliferation studies of advanced fuel cycles.
Advanced Fuel Cycles
Problem Description
Advanced uranium fuel cycles are envisaged to incorporate recycling of plutonium and minor actinides. In this way, both a high degree of renewability and waste reduction are achieved. With respect to the transmutation process, complementary to partitioning of plutonium and minor actinides, fast reactors offer best performance. Delay of the introduction of fast reactors will lead to further increase of the stockpile of plutonium from spent LWR fuel for yet undefined time. Optimum ways to improve the manageability of stocks of transuranics in the prelude to the fast reactor age are therefore required. Focus is on investigation of the potential of the presently installed LWR's for plutonium consumption. A second major issue to be tackled is the problem of reprocessed uranium (REPU).
Project Definition
The project is embedded in the OECD Working Party on physics of Plutonium Recycling {OECD WPPR). The objectives are to establish the plutonium consumption potential of LWR's, and to contribute to assessment of possible solutions to the problem c-f REPU, both on an international level within the framework of the WPPR. The results of the WPPR will be a report containing results of benchmark calculations of selected cases, summaries of existing results, and studies. The WPPR will focus on the (reactor) physics issues. Interfaces are to be defined with chemistry and MOX fuel fabrication. Due to its limited size, the present ECN project cannot contribute to the following subjects of the WPPR programme: plutonium in inert matrices; plutonium recycling in advanced converters; and plutonium burning in fast reactors.
Project Description
Studies will be made of existing published results on specific objectives, e.g. the role of the plutonium vector, the dependence of safety coefficients on the plutonium enrichment, the role of the moderating ratio, reactor control, minor actinides production. Benchmark calculations will be done in order to establish the required plutonium enrichment for multiple recycled plutonium, and to determine the dependence of the void coefficient on plutonium enrichment.
Project Control
Benchmark calculations and the draft of the qualitative introduction to the report will be completed in 1993« The draft interim report will be completed in 1994. Total commitment for this year is 0.4 manyear.
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Non-Proliferation
Problem Description
The focus is initially on the proliferation aspects of Fast Reactors with recycling of plutonium, uranium and minor actinides. At later stage, also the High Temperature gas cooled (thermal) Reactor (HTR) with recycling of uranium and thorium will be investigated.
Project Definition
The objectives are: 1. To perform analyses of the proliferation aspects of advanced closed
nuclear fuel cycles: fast reactors with partitioning and transmutation of minor actinides, and the HTR with uranium-thorium recycling.
2. Development of Non-Destructive Assay (NDA) methods for plutonium and, if necessary, minor actinides for safeguards purposes.
Results of analyses will be communicated as reports to international fora such as the IAEA. The development of NDA methods is either part of, or communicated through, programmes of the European Safeguards Research and Development Association (ESARDA), such as related to the Plutonium Isotopic Determination Inter-comparison Exercise (PIDIE) project [3] and ESARDA symposia. Development of NDA methods is at present restricted to gamma-NDA of plutonium within the framework of ESARDA. Effort is made to embed the project more firmly in international collaborations, e.g. such as with ANL/DOE on the IFR, or in joint European projects.
Project Description
The analyses will deal with the following four subjects: 1. The proliferation relevant features of advanced nuclear fuel cycles
with respect to present LWR fuel cycles. The major points are the significance of minor actinides, and the technologies for advanced reprocessing. A first survey of the IFR, under development at ANL, has been completed [1].
2. The proliferation resistance required of future nuclear power generation, given recent international findings and political developments.
3. The strengths and weaknesses of the present Non-Proliferation Treaty and accompanying IAEA safeguards system.
4. Possible adaptations of this system (3) to meet the future requirements of (2), given the envisaged technologies of (1).
This analysis has features in common with the International Nuclear Fuel Jycle Evaluation (INFCE), which also focused on the proliferation aspects of recycling and the safeguardability of reprocessing. Reevaluation of IWFCE results is therefore the starting point of the present project [2],
Development of gamma-NDA methods for minor actinides starts with feasibility studies. To this end, pilot measurements have been done of 190 MW.d/kg burnup PFR fuel with initial plutonium content of *30#t which are presently analyzed.
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Project Control
Work on subjects 1 and 3 (section 3.3) will start in 1993 and be completed in 1994; work on subject 2 will start in 199**. The feasibility study of gamma-NDA of minor actinides will be completed in 1993-Total commitment for 1993 is 0.M manyear.
References
[1] V.A. Wichers and R.J. Heijboer. "INFCE REVIEW: Part A. Introduction, Fuel Cycles and Technologies", ECN-I—93-001.
[2] V.A. Wichers, R.J.S. Harry, "Analysis of The Integral Fast Reactor", (draft).
[3] J.K. Aaldijk, P.A.C. de Betué, "Measurements of the Relative Amount of 241Am in PIDIE Samples", NFA-SG-91-09.
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5.5. Studies on transmutation of nuclear waste {project id no. 1115/1*193
Problem description
The present project fits into the much broader ECN 'Integral Research Program 1991~199^ for the study of Recycling Actinides and Fission Products'. This program has the intention to give a contribution to an international effort to evaluate the recycling option of the nuclear waste problem. Besides strategy and scenario studies, reactor physics research, and an effort towards small scale demonstrations of transmutation possibilities, the above mentioned 'Integral Research Program' comprises project 1^93 as its general scientific support, and project 1115, as a main nuclear data project. The nuclear data project is performed on a cost-shared basis together with the CEC.
Project definition
Project 1115 has as an objective the improvement of the nuclear data base needed for inventory calculations for strategy studies on nuclear waste transmutation, whereas project 1493 has general strategy studies on transmutation of actinides as its objective.
In the frame of the IL7 strategy studies, ECN participates with CEA/Cadarache and Siemens (Interatom) in scenario studies in order to investigate the technology and the economy of transmutation of nuclear waste. As far as ECN is concerned, there are two subjects to be considered in this relation: a. Improvement of working libraries of nuclear data, which are relevant
to the waste problem. b. Strategies for long-lived fission products.
Project description
For transmutation of actinides one judges ALMR's as more favourable as LWR's, therefore attention is given to the innovative reactor PRISM. In the frame of PHD-thesis work with title "Burnup of Actinides and Fission Products", a stay at General Electric was sponsored, with *"- e purpose to study the burnup of transuranium elements Ln an ALMR. This ,*ork specifically emphasises differences between metallic and oxidic fuel. On request of the EC a feasibility study is being made on transmutation of fission products.
In the period 1991~1994 a nuclear data base will be prepared, which is in accordance v.ith the most recent European JEF-2 library. This nuclei data base is intended for the verification of sample-burnup calculations. A report is being issued, which contains the description and the contents of a new multi-group library.
For burn-up calculations on transmutation scenarios, several versions of the ORIGEN code are used. In addition to a general assessment of the ORIGEN nuclear data library for transmutation studies the data base, mentioned above, will be inserted in the ORIGEN library, including inspected and updated fission yields and cross sections for relevant traces.
82
A comparison of cross sections of the most important actinides and fission products in the ORIGEN nuclide libraries with cross sections from literature (JEF 1.1, JEF 2.2, JENDL-3) will include the calculation of self-shielded cross sections not only for a PWR, but also for an LMFBR, by means of an update of the ORIGEN nuclide libraries.
In addition to previous global burn-up calculations in-depth calculations will be performed for future heterogeneous transmutation facilities, in which waste elements are contained in inert matrices, to be irradiated in a thermal neutron flux. Validation of the ORIGEN nuclide libraries with burnup calculations on a PWR and a fnst reactor will be included as a validation of the new ORIGEN nuclear data libraries. A contract with CEC includes a study of the transmutation rates of Tc-99» I~129t and Cs-135 in several reactors such as the High Flux Reactor, the Phénix reactor and in a Light Water Reactor. This study will be extended to include other reactor types as well.
A report on 'Investigation of the possibilities of transmutation of long-lived fission products' will be written. Accelerator driven facilities with a high flux density will be included in a comparison with purely reactor based transmutation. An appendix will be presented of nuclear data and nuclear models at an energy up to 1 GeV, which are relevant in designs of accelerator driven machines. For such future facilities calculations will be performed on the irradiation of fission products and inert matrices (including cladding).
Project control
Mo&t. of the work, which is funded by the EC (nuclear data and strategy studies for the transmutation of fission products) will be rounded of in 1994. Reactor physics calculations on large scale transmutation of fission products will be described in a report to be issued to the CEC in 199*1 • In the frame of the contract with the CEC and on request of the European Parliament, a report will be written under the title 'Recycling Fission Products', with chapters on Motivation, Partitioning, Targets and Devices for transmutation. Study of transmutati<\ of actinides in ALMR's and LWR's will lead to a PHD thesis in 1994,
The ECN contribution for 1993 and 1994 will amount to a personnel effort of 6 manyears, including a contribution of the EC of 1 manyear for 1994.
References
[1] Contributions for Global '93: a) J.L, Kloosterman, New Working Libraries for Transmutation Studies. b) A.J. Koning, H. Gruppelaar, and P. Nagel, Nuclear Data Evaluation
for Accelerator-Based Transmutation of Radioactive Waste. c) J. Bultman, C.L. Cockey an T. Wu, Actinide Burning and Breeding in
Metallic and Oxide Fueled ALMR Cores. d) J. Bultman, Reduction of Nuclear Waste by Introducing Advanced
Liquid Metal Reactors. [2] K, Abrahams, EC Progress report under- contract no. F12W-CT91-0104,
March 1993. [3] J. Bultman, Calculation of the Transmutation Rates of Tc-99, 1-129 and
Cs-135 in the High Flux Reactor, the Phénix Reactor and in a Light Water Reactor, ECN-I—92-013.
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5.6. Reactor physics aspects of PRISM (project id no. ltyk)
Problem description
Inherent safety characteristics as well as sustainability are important ENGINE objectives. Although no shortage of presently running and next generation thermal reactors is foreseen, it may still be important for the long-term sustainability to take into consideration reactor systems with high conversion rates (breeders). The "present generation" LMFBR's is not specifically optimized to inherent safety features. Designers of new -innovative - reactors, such as GE-PRISM, claim however a high level of inherent safety. The reactor physics design (control of reactivity) is an essential item in this respect. Fast reactors may play an important role as burner of nuclear waste.
Project definition
The objective of this project is to make an analysis of the reactor physics design of PRISM with respect to safety characteristics.
Project description
Knowledge on fast reactor (PRISM) assessment has to be built up by literature studies and by performing computations in collaboration with General Electric Co. For a period of 8 months a physicists has been assigned at GE for core calculations with emphasis on the specific features of PRISM as an actinide burner. The activities at ECM evolves according to the following lines:
adaption of available multi group cross section libraries for fast reactors benchmark calculations for fast reactors literature review of innovative fast reactorc with metallic fuel (ALMR) related to reactor physics safety analysis of reactivity coefficients (sodium void, Doppler) for the PRISM core
Project control
Most of the activities have been performed. A final report will be made by the end of 1993. The manpower for 1993 amounts to 0.7 manyear.
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5.7- Technological research RAS {project 1*195)
Problem description
Partitioning and transmutation of actinides and fission products may be an option to reduce the long-tern radiological hazard of the high-level nuclear waste. Besides extensive studies on nuclear physics and scenario's for transmutation in various types of nuclear reactors also experimental effort will be required to derive basic information, to explore optimal technological solutions and, at the end, to perform full-scale demonstrations•
Project definition
The objective of this project is to investigate , in internaw Jl frame, 'he technological aspects of transmutation of long-lived waste components Dy irradiation experiments in the High Flux Reactor in Petten. A European collaboration EFTTRA, "Experimental Feasibility of Targets foi T3Ansmutation" is being set up as a shared action for irradiation experiments. The following five partners are participating: CEA Cadarache (France), EdF Paris and Lycn (France), ITU Karlsruhe (European community), KfK Karlsruhe (Germany) and ECN. In this way the available competence and infrastructure (test reactors HFR and PHENIX and the Hot Cell Laboratories of ITU, CEA and ECN) can be deployed.
Presently the programme is restricted to transmutation experiments of the actinide americium and the long-lived fission products technetium and iodine in a high thermal flux; in a later stage also neptunium may be investigated. Only heterogeneous transmutation (irradiation in target pins in stead of having it mixed through the reactor fuel) is being considered. In this way the targets can be optimized for the transmutation process, the irradiation is decoupled from refuelling schedules and multiple partitioning from the fuel waste is avoided. Technetium will be irradiated as pure metal and iodine in the form of the relatively stable cerium or yttrium iodide as well as low-melting lead iodide. For americium the inert matrix concept has been adopted. Key items to investigate are: transmutation rate and self-shieldirg effects and the knowledge of relevant properties and behaviour of technetium, americium and iodine compounds under irradiation.
Project description
A working programme for the EFTTRA collaboration has been defined, split up into three phases: 1. Irradiation of small samples of fission products and bare inert
matrices in HFR and PHENIX. 2. Irradiation of sumples of americium in inert matrices in HFR and
PHENIX 3. Irradiation of full scale pins of fission products in HFR and PHENIX.
A concept collaboration agreement, including a detailed technical annex for the first phase, has been prepared. Project coordination, including full specification of experiments (samples, irradiation conditions), is a common effort of the collaborating partners. ITU, CEA and ECN will be in charge of the experiments: ITU will elaborate and characterize the samples, ECN resp.
85
CEA will perform the irradiations in HFR resp. PHENIX and the post-irradiation examinations will be done by GiA, ECN and ITU.
The first irradiation of fission products and inert matrices will be executed in the HFR [1],[2]. Detailed description is given in ref. [3], During 6 HFR operation cycles (150 days, thermal flux 3 10" n/s/cm2) an insert containing 27 small samples will be irradiated at constant temperature level of 400 *C: 3 metallic technetium samples, 6 iodide samples (Cela and Pbl2) inert matrices (A1203, Ce03, Y203, MgO, MgAl204, Y3Ai2Oi2}.
Presently the technetium samples are being fabricated at ITU. An irradiation facility for the HFR, especially designed for this project in collaboration with JRC establishment in Petten, is under construction at ECN/JRC.
Project control
Irradiation experiments of small samples of fission products and inert matrices will take place in HFR and PHENIX in 1994. Irradiation experiments of americium in inert matrices are planned in 199^/1995. Irradiation experiments of full scale pins of fission products are planned for 1995/1996.
The ECN contribution amounts a personnel effort about '* manyears per year. Additional financial support is supplied for investment in irradiation facility and hot cell equipment. JRC-Petten (the owner of the HFR) is contributing to the design and fabrication of the facility.
References
[1] H. Gruppelaar and W.M.P. Franken, "Recycling long-lived waste
components in high-flux thermal systems". Nuclear Europe Worldscan 5-6, June 1993. P. 39.
[2] R.J.M. Konings, K. Abrahams, W.M.P. Franken, H. Gruppelaar, J.L. Kloosterman, P.J.M/ Thijssen and R. Conrad, "Technological aspects of transmutation of technetium and iodine". Paper presented at GLOBAL 1993.
[3] R.J.M. Konings and W.M.P. Franken, "Irradiation program for fission products and inert matrices in the HFR". ECN Memo ENGINE-93-Q2.
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6. FUSION TECHNOLOGY
6.1. Introduction
In Europe the development towards a future thermonuclear fusion power reactor is a common s "egic issue of the countries of the European Community (EC).
One of the most successful European projects is JET (Joint European Torus), located at Culham, Great Britain. This worlds largest plasma physics experiment is dedicated to demonstrate within a few years from now the scientific feasibility of thermonuclear fusion.
Worldwide, under the auspices of the IAEA, the European Community together with USA, Japan and Russia, are now also designing the next step machine ITER (International Thermonuclear Experimental Reactor). This machine, to be built and tested in the period 2000-2020 has the aim to demonstrate the engineering feasibility of thermonuclear fusion. Thereafter a DEMO fusion power reactor has to demonstrate the economic feasibility of thermonuclear fusion. In Europe there has been a parallel activity on a sin>** ..- macM^e NET (Next European Torus).
To support the design of ITER (or NET) and DEMO an extensive fusion technology R&D programme has been set up. ECN is involved in several items in this fusion technology R&D programme.
The main activities of ECN are directed to material research, i.e. study of the effects of neutron irradiations on the behaviour of structural materials and breeder materials, relevant for ITER and DEMO. In this respect experiments in the HFR are essential. Also reactor physics analyses and safety analyses form part of the ECN contribution in this R&D field.
In the following sections all ECN projects for fusion are discussed. According to the organizational structure and different financing of the European Fusion Technology Programme the ECN fusion technology projects are grouped in the following categories: x. Work in the frame of NET/ITER related Technology. 2. Work in the frame of Breeder Blanket Module Programme. 3. Work in the frame of Long Term (DEMO) Programme. k. Work carried out under JET and NET contract.
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6.2. NET Related Technology Programme
Irradiation and testing stainless steel 3l6L (project id no. 1653)
The final irradiation experiment with TIG weld-deposit material will be completed in 1993« The post-irradiation testing of tensile specimens, low cycle fatigue specimens and compact-tension specimens for fracture mechanics experiments will be continued. Particular emphasis will be put on the low cycle fatigue testing and fatigue crack growth testing of EB-welded specimens, because of preliminary indications of low fracture resistance of the fusion zones adjacent to the weld. Post-irradiation fracture mechanics experiments with TIG weld-deposit material have indicated reduced fracture toughness, even after the low dose irradiation up to 0.5 dpa at 350 K. In this respect it has to be mentioned that the fracture resistance of unirradiated TIG weld-deposit material is much less than that of the 3l6L{N) plate material. Tests with unirradiated TIG-weldments have shown fractures in the heat affected zones adjacent to the weld (HAZ). The fracture behaviour of realistic TIG-weldments has to be characterized in terms of cyclic and monotonic loading crack growth properties of the HAZ. An additional irradiation experiment with TIG-weldments is planned to start in 1993. i*1
order to quantify the effect of irradiation on the fracture resistance of the HAZ.
Pnj.ject control
The manpower for 1993 is 3-3 manyears.
Reweldability of irradiated 316L (project id no. 1668)
The characteristics of joints of irradiated 316L-SPH with non-irradiated material will be investigated. Emphasis will be put on the helium enbrittlement phenomenon, caused by the elevated temperature tensile stresses from the heat input of the joining technique. Specimens will be manufactured in the hot cells from previously irradiated samples. A laser welding equipment is placed in the hot cell laboratory and appropriate welding conditions and parameters will be detenrineo. To determine the fracture properties of the welded joint in som*- qualitative way a number of notched specimen^ will be made for tensile testing, In 199^ a more extensive quantitative investigator, .„ill be made on the mechanical behaviour of laser welded joints. For this purpose additional neutron irradiation experiments will be prepared.
Project control
The manpower for 1993 is 1.0 manyear.
Off-normal high heat loads PFC (project id no. I65I)
For the simulation of the effects of off-normal heat loads on plasma facing materials (associated with plasma disruptions) an upgraded experimental facility is being used. The upgrading includes new and fast diagnostics for laser bear characterization, surface temperature measurement and vapour
88
plume characterization (like composition and expansion behaviour}. Comparative & complementary experiments are being performed in several electron beam facilities, mainly FE200 at Le Creusot (CEA/Framatome) and some with JUDITH (KFA). Materials studied are mainly advanced carbon fiber composites (CFC), including newly developed high conductivity CFC. SiC-CFC and B4C-CFC. More attention will be given to a proper correlation of the observed materials response with thermo-physical properties. The numerical work will concentrate on the supporting work for the analysis cf experiments, like e.g. 2-D effects. Also parametric studj.es will be done on the thrrmal behaviour of potential PFC protective coatings. To study aspects of melt layer stability and movement some metals with low melting point will be investigated like Cu, Al, Si. Preparations will be made to in*, estigate experimentally the performance of beryllium under simulated plasma disruption conditions. As the ITER JCT is focusing on berylliuc for plasma facing components, it is foreseen that dedicated experiments will be carried out in 199** •
Project contr ,1
The manpower for 1993 is 1.2 years.
Tritium retention "'aphx ce (project id nr
Screening w^»cs o.. - loading of graphite samples with tritium have been c, .ried out. Tritium retention measurements on un-irradiated samples are in progress, the results will serve as a reference for the measurements of the irradiated samples. In addition, tests will be performed with a different carrier gas for tritium loading. The retention and the desorption measurements on irradiated graphite samples, as a function of irradiation temperature and neutron dose, will be carried out in 1993 and in the beginning of 199** • The experiments will be continued in 199** with improved (SiC-doped) C-based materials.
Project control
The manpower for 1993 is 0.6 manyear.
EFF file management and user support (project id no. 1655)
The EFF-2 data file is maintained at one central point (ECN-Petten) from which th~ file is distributed, documented and checked. The final goal is to bring EFF-2 to the user and to improve it according to the needs. In this process corrections, small updates, additions and partial re-evaluations are considered as well.
Updates will be implemented and re- uations will be performed according to urgent needs, feedback from users, _.d from benchmark calculations will be taken into account as well. Data resulting from other subtasks will be compiled and introduced in the evaluations, which will be distributed among users. A second intermediate report will be published at the end of 1993-Updates and partial re-evaluations will continue in 199^ and this will be performed according to the highest priorities. In particular major structural materials of stainless steel, resolved resonance range in Pb and
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covariance data will be improved. Final report will be published at the end of 1994.
Project control
The manpower for 1993 is 1.0 manyear.
Safety Studies (project id no. 1656)
The next accidents will be analyzed for the blanket cooling system of the NET/ITER device:
loss-of-coolant accidents outside the vacuum vessel; loss-of-flow accidents; total loss of off-site and on-site electrical power.
The analyses concern the transient thermal-hydraulic cooling system behaviour and tne temperature development in the nuclear components during these accidents. The analyses will be performed using the thermal-hydraulic system analysis code RELAP5/MOD3.
The analyses of the above-mentioned accidents for the divertor cooling system were finished in 1991- For the first wall cooling system, stage I of these analyses were finished in 1992. In 1993 and in 199** the blanket cooling system will be analyzed.
Project control
The manpower for 1993 is 1.0 manyear.
Magnets (project id no. 1662/1663)
To analyze the mechanical behaviour of the ITER-EDA toroidal field coil, a new detailed finite element model has been developed, as well as procedures to define a large range of boundary conditions.
The model and boundary conditions have been discussed in detail with the NET-Team. No changes of the model are necessary. After checking the validity of the model and loads (Lorentz forces at normal operating conditions) the first linear elastic analysis will be made. After extensive evaluation jiore (different) analyses will be made. The model is Also suitable for analyses of faulted conditions. Continuation of the work in 199^ depends on the results of the Technical Magnets Meeting, which will be held in 1993-
Project control
The manpower for 1993 is 0.7 manyear.
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6.3. Solid Breeder Blanket Programme
Development of Ceramic Tritium Breeding Materials (project id no. 1657) The main activities for 1993 will be: Detailed analysis of the measurements obtained from the large number of transients, both temperature and purge gas composition changes, perfcnsed during irradiation of EX0TIO5 and -6 will be continued. This includes not only determination of tritium residence times but also definition of the controlling mechanisms for tritium release at different irradiation conditions I.̂ ke temperature level and purge ?as cojwvr.iition and possibly also the material microstructura. This analysis will be supported by the results trom the post-irradiation measurements, tritium retention and tritium release, on the materials irradiated in EXCfiTC-5 and -6. These measurements will be continued. Tha release experiments include measurements at different heating rates and at different constant temperatures, to determine also activation energies and diffusion/desorption coefficients. For these experiments also a combined Thermobalance-Quadrupole mass spectrometer will be used.
The design of the capsules for the high-bumup irradiation, EXOTIC-7. is completed. The fabrication and the assembly of the capsules will take place in 1993. The samples from KFK-Karlsruhe (Germany), ENEA-Casaccia (Italy), CEA Saclay (France) and ECN Pet ten will be available in the period June to September 1993- The final assembly will be completed in November 1993« The start of the irradiation is scheduled for December 1993* During the irradiation temperature transients will be performed and analyzed for determination of tritium residence times. The irradiation will be terminated at the end of 1994.
Project control
The manpower for 1993 is 2.5 manyear.
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6.4. Long Term Programme
European Activation File (project id no. 1659)
The development of the European Activation Pile (EAF) will continue by preparing data for the new version EAF-4, which will be completed in August 1993* It includes revisions and quality improvements of data for actinides, revisions and additions of radiative capture cross-sections, other available high-quality evaluations (through international exchange), and is provided with a complete uncertainty subfile for all reactions in a three-group structure.
Thereafter, the upc?.te of experimental renormalizations. horizontal revtsvjn cf evaluate..; (new systematica at 14.5 Mev), updating of the uncertainty file . r important reactions, and inclusion of new h\gh-qu&lity evaluations based on the feedback from users will be carried out. This work wiJ*. finally result in the release of the EAF-5 file at the end of 1994.
Project control
The manpower for 1993 is 1.2 manyear.
Development of vanadium alloys (project id no. 1658)
Present work on helium embrittlement of vanadium and vanadium alloys, including recently developed alloys, will be continued during the first half of 1993- Analysis of the data and final reporting on this subject should be completed by September. During the second half of 1993 the emphasis will be on determination of the effects of various alloy additions, as well as interstitial imparities, on mechanical properties of experimental alloys. Alloys to be investigated in 199** will include candidate vanadium alloys developed for fusion elsewhere (e.g. Argonne National Laboratory, US, and Tohoku University, Japan).
Project control
The manpower for 1993 is 1.4 nanyear.
Martensltic steel properties (project id no. 1660)
A series of small irradiation capsules, each with 9 tensile specimens of MANET II material, will be manufactured in the first quarter of 1993- The design of a new rig (CATETO) for fracture toughness specimens will be completed in the same period. The various capsules are planned to be successively irradiated in 1993- The target irradiation dose levels will be about 1 dpa, which is considered to be very close to the saturation dose level for MANET.
Post-irradiation mechanical properties of the MANET material strongly depend on the irradiation temperature. The envisaged application temperatures of the MANET material are in the range of 500 K to 800 K. The lower part of this range is the most damaging in terms of degradation of the fracture toughness by irradiation. Consequently the irradiation
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experiments with MANET II specimens will focus on an irradiation temperature of 575 K. Post-irradiation test temperatures will range from 300 K up to 800 K.
Project control
The manpower for 1993 is 1.2 manyear.
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6.5. JET and NET contracts
Neutron Diagnostics for JET (project id no. 1637)
Detector responses will be calculated with the FURNACE code for the new JET torus topology as it will be in 1993 after the divertor and divertor coils have been installed. The calculations will be performed for DD-operation and DT-operation using new nuclear data from the EFF/JEF library.
Project control
The manpower for 1993 is 0.5 manyear.
B«C coatings for First Wall (project id no. 1666)
Plasma sprayed coatings of boron carbide on 316 type steel will be manufactured by EC industry under subcontract. Coated samples will be tested and analyzed on their potential for First Wall protection in ITER. Experimental campaign«: for thermal shock testing will be performed. In addition mechanical and thermal characterization will be done. The work is performed in close collaboration with CEA & KFA in the framework of coating development and component tests.
Project control
The manpower for 1993 is 0.4 manyear.
Stress analysis ITER coil (project id no. 1670)
Subject of the contract is the structural analysis of the magnet system of ITER-EDA to evaluate the effects of the thermal and electromagnetic loads in normal conditions. The work concentrates on the analysis of the effect of the friction among the contact surface and the stress distribution in cross sections of the TF-coils. The calculation of the global system will be carried out with the finite element model developed by ECN, which was successfully used for preliminary analysis of the ITER magnets. In addition a detailed model of the TF-coil cross section will be used to get the stress distribution in the different parts of the coil.
Project control
The manpower for 1993 is 0.2 manyear.
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7. APPLICATIONS
7.1. Boron Neutron Capture Therapy (project id no. 1099)
Introduction
A selective treatment of malignant tumours by short range charged particles of high energy is offered by Boron Neutron Capture Therapy (BNCT). The basic principle is very simple: Thermal neutrons are with high probability (cross-section) captured by boron-10 nuclei. After neutron capture the resulting boron-11 nucleus decays into an a- and a Li-particle, which are stopped within the range of one cell. As the energy of the emitted particles is sufficiently high the cells stopping the particles will be destroyed. Assuming that i/ selective uptake of boron in tumour cells can be induced and that ii/ the tumour area can be exposed to the proper neutron flux, this simple basic principle can be turned into a selective cancer therapy. If furthermore the parameters of the reaction can be chosen in such a way that no damage is caused to normal tissue, this cancer therapy becomes not only efficient but also safe for healthy tissue.
The availability of tumour specific boron compounds and intense neutron beams from nuclear reactors has currently resulted in several BNCT-projects worldwide. In Japan, treatment of tumours is presently carried out with encouraging results using thermal neutrons. Efforts towards the clinical implementation of BNCT with epithermal neutrons in the United States are exploited around three research institutes (INEL Idaho, BNL Brookhaven and MIT Boston).
Problem description
The basic principle of BNCT is very simple. However, the conversion into a clinical therapy is very complicated.
In Europe the perspectives of treating conventionally incurable brain tumours (glioma) with the healthy tissue sparing modality BNCT has brought forward a European Collaboration embedded in the frame of a Concerted Action. The general objective, conversion of this simple principle into a clinical therapy, is further defined as: 1/ Primary objective, treatment of brain tumours, glioblastoma, with epithermal neutrons from the High Flux Reactor (property of the Commission of the European Communities) and using BSH as a boron carrier. At present this European BNCT project is approaching clinical trials trough a healthy tissue tolerance study on large test animals (dogs). 2/ Secondary objective, treatment of other tumours than glioblastoma with neutrons from other sources than the HFR and using other boron carriers than BSH.
Although epithermal neutrons from the clinical facility of vhe 45 MW materials testing reactor HFR are needed for the European approach to treatment of glioma, thermal neutrons from the much smaller and less expensive low-power research reactor of the Argonaut type, the LFR (property of ECN), have an important function within the further development and application of BNCT. A part of the study of the fundamental radiobiological processes, responsible for the achievement of the therapeutic gain of BNCT, consists of the irradiation of cell cultures and small test animals with neutrons. As the
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cross-section for the 10B(n,a) reaction reaches the high value of 3800 barn for thermal neutrons only, these irradiations must be conducted using thermal neutrons. This is different from a therapy situation treating tumours inside an organ (e.g. brain tumours), where the field of thermal neutrons is resulting from initially epithermal neutrons slowing down in the hydrogenous (brain) tissue.
Project definition
The ECN contribution to the development of a clinical therapy facility at the HFR is of importance for the outcome of the approaching clinical trials. This facility is mainly financed by the JRC. The innovative study is focused on the development of the European BNCT-project at the HFR. This study is not only conducted at the HFR. The LFR is being developed as a thermal neutron source for biological BNCT-studies on the short term and for clinical applications on the long term.
Objective 1. Clinical trials on gliomas at an epithermal facility of the HFR. 2. Preclinical studies at the LFR. 3. Clinical implementation of BNCT at a thermal facility of the LFR.
Results An epithermal facility suitable for clinical application has been developed and characterized at the HFR. This facility includes a liquid argon filter for gamma-suppression. A method for reliable beam characterization has been developed. A healthy tissue tolerance study on a canine model has entered the observation phase. This study implicated irradiation of 47 dogs. For the development of a thermal facility at the LFR, a benchmark experiment has been conducted and a design has been concluded. Phantom measurements are implemented into a (first order) treatment planning model. A proton recoil spectrometer using a small plastic scintillator is being developed in collaboration with the Paul Scherrer Institute in Villigen (CH). A preliminary micro-dosimetric model has been constructed. This model will be further developed in collaboration with Idaho National Engineering Laboratory.
Project descriptii1
Working method The project carries a clear multi-disciplinary character and is coordinated by the Concerted Action of the European Collaboration. As the objective of the primary objective approaches realization (clinical trials expected to start during 1st half of 199*0. the project exhibits increasing ECN-identity through the activities involving the LFR. For the feasibility study of the LFR, the existing network of multidisciplinary contacts within the European Collaboration is extensively used. The project of ECN are solidly framed in this network. The contribution from ECN concerns the following disciplines: nuclear physics, reactor physics, neutron metrology, radiation transport calculations and radiobiology (the radiobiological activities are defined as a separate project within the unit Radiation). The international collaborations for the development of low-powei research reactors will be further extended.
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Activities The main activities of ECN concern the following areas: 1. Development and test of a clinical facility at the HFR (epithermal
neutrons). 2. To conduct the Healthy Tissue Tolerance Study on a canine model in
collaboration with the biology group from the unit Radiation. 3. Construction, installation and characterization of a biological
thermal neutron facility at the LFR. 4. To conduct a feasibility study for a clinical facility at the LFR. 5. Computer simulation of the energy transfer on a cellular scale (aacro-
and micro-dosimetry). This study is conducted at both reactors. 6. On-line measurement through gamma-spectroscopy of boron distribution.
This study is conducted at the LFR. 7. Development of neutron spectroscopy methods sensitive to the energy
range 10 - 100 keV.
Project control
Clinical trials at the HFR are expected to start during the first half of 1991*. The duration of the present LFR-project is 3*5 years.
The manpower is 2.0 manyear per year.