[en] The performance and use of PSA:s in Sweden goes back about two decades. During all of this time, the field of PSA has been developing intensively, both internationally and within Sweden. The latest years have been characterised by an increased use of PSA models and results, and by major extensions of existing PSA models. The aim of this document is to describe PSA in Sweden with respect to development, scope and maturity, as well as to the contents of the analyses and the use of results. PSA activities will be described from the point of view of both the authorities and the utilities. The report gives an overview of the development within the area of PSA in Sweden both its history and current trends. The aim has been to include a reasonable amount of detail, both on the methods and results in PSA:s performed and on the numerous supporting research programs dealing with various aspects of PSA. 39 refs 39 refs

[en] This report examines the problems that US and certain foreign reactors have experienced with intergranular and transgranular stress corrosion cracking. Included is a review of the failure modes and mechanisms, various corrective measures, and the techniques available to detect and size the cracks. The information has been organized into four time periods: late 1950s to mid 1960s; mid 1960s to 1975; 1975 to 1985; and 1985 to 1991. The key findings concerning BWRs are: Corrective actions have led to a substantial reduction of IGSCC; Control of carbon levels - through use of ELC or NG grades of austenitic stainless steels - should minimize IGSCC; Control of residual stresses, particularly with IHSI, greatly reduces the incidence of IGSCC; Hydrogen water treatment controls the oxygen and should limit IGSCC; The problem with furnace-sensitized safe ends is well recognized and should not recur; In most cases, severe circumferential SCC should lead to detectable leakage so that leak-before-break can be identified; IGSCC of austenitic stainless steels can occur in all pipe sizes from smallest to largest, especially when stress, sensitization, and oxygen are all present. In the case of PWRs, it is clear that the incidents of primary water stress corrosion cracking appear to be increasing. Cases containing steam generators, austenitic stainless steels, and Inconels have been known for years. Now it is occurring in safe ends and piping at very low oxygen levels. Secondary side water chemistry must be controlled to prevent SCC in PWRs. 18 refs

[en] A simple biosphere model has been designed for use in the SKI Project SITE-94 related to a hypothetical repository for spent nuclear fuel on the island of Aespoe. The model provides results in terms of radiation dose per 1 Bq/year, unless otherwise indicated, and results will thus have to be scaled with actual flux of radionuclides per year entering the primary biosphere recipients. The model does not include radioactive decay as there is assumed no delay in the model system, except for where explicitly mentioned. Specifically, no radioactive transitions resulting in daughter nuclides are considered. Calculated yearly individual and population committed (50 years) radiation doses to man are expressed as mSv/h, under the assumption of a flux of one Bq/year into the primary biosphere recipient. Calculated radiation doses resulting from the present biosphere model are hypothetical, and should under no circumstances be considered as real. Neither should they be used as quantitative information for decision purposes. The biosphere model is of a rough and primitive character and its precision, relative to the real biosphere in the surroundings of Aespoe is envisaged to be several orders of magnitude. 8 refs

[en] In the SKB proposal for a Swedish nuclear waste repository, copper canisters are used for encapsulating the spent fuel. The chemical and physical behavior of Copper in the repository environment will therefore be of critical importance for the repository integrity. The present work concerns a literature review of Copper and selected heavy metal sulfides as they are expected to play an important role in the repository environment. The interest is focused on their properties as described by crystal structure, electrical properties, atom mobility, solubility in water, mechanisms of sulfidation and selected thermodynamical data. 56 refs, 14 figs, 5 tabs

[en] The central scenario includes the following main components: a deterministic description of the most probable climatic state for Sweden (with special ref. to the Aespoe area) for the next c. 120,000 years, a description of the likely nature of the surface and geological environment in the area at each stage of the climate sequence selected, and quantitative information on how these changes might affect the disposal system. The climate models suggest glacial maxima at c. 5, 20, 60 and 100 thousand years from now. The Aespoe region is predicted to be significantly affected by the latter three glacial episodes, with the ice sheet reaching and covering the area during the latter two episodes (by up to c 2200m and 1200m thickness of ice, resp.). Permafrost thicknesses over the next 120,000 years have been calculated. Assumptions, estimates and alternatives to the prescribed climate evolution are discussed. Following definition of a realistic, albeit non-unique, climate sequence, the objective of scenario development is to provide an indicator of the physical, chemical and hydrogeological conditions at the front of and beneath the advancing and retreating ice sheets, with the aim of identifying critical aspects for Performance Assessment modelling. The effect of various factors, such as ice loading, development of permafrost, temperature changes and sea level changes are considered in terms of their impact on hydrogeology, groundwater chemistry, rock stress and surface environments. 183 refs

[en] The purpose of the presented work has been to give a short summary of the origins of many uncertainties arising in the designing and performance assessment of a repository for spent nuclear fuel. Some different methods to treat these uncertainties is also included. The methods and conclusions are in many cases general in the sense that they are applicable to many other disciplines where simulations are used. As a conclusion it may be noted that uncertainties of different origin have been discussed and debated, but one large group, e.g. computer simulations, where the methods to make a more explicit investigation exists, have not been investigated in a satisfying way. 50 refs

[en] The APRI-project started in 1992 with participation of the Swedish Nuclear Power Inspectorate (SKI) and the Swedish utilities. The Finnish utility TVO joined the project in 1993. The aim of the project has been to work with phenomenological questions in severe accidents, concentrating on the risk-dominating issues. The work is reported in separate sub-project reports, the present is the final report of the methodological studies as well as a final report for the total project. The research has led to clarifications of the risk complex, and ameliorated the basis for advanced probabilistic safety analyses, specially for the emission risks (PSA level 2) which are being studied at the Swedish plants. A new method has been tried for analysis of complicated accident courses, giving a possibility for systematic evaluation of the impact of different important phenomena (e.g. melt-through, high pressure melt-through with direct heating of the containment atmosphere, steam explosions). In this method, the phenomena are looked upon as top events of a 'phenomena-tree', illustrating how various conditions must be met before the top-event can happen. This method has been useful, in particular for applying 'expert estimates'. 47 refs

[en] The work presented in this report is a result of reading the SKB program for R,D and D on safe storage of radioactive wastes. Our work, which is focused on the waste canisters, was commissioned by the Swedish Nuclear Power Inspectorate. We find the program very difficult to follow owing to the lack of detail in chapter seven. In our opinion this will make the work difficult to monitor by SKI or SKB. We also feel that the interpretation of information already available is overoptimistic. As a consequence the difficulties ahead are understated and the programme is converging too quickly. We believe that it should be possible to develop a satisfactory canister for disposal of high level nuclear waste according to the general method proposed by SKB and with the proposed capacity within the timescale of the overall programme. We do not believe, however, that all the difficulties have been recognised. As a consequence of this the results to date are interpreted optimistically. We believe that progress should be subjected to more professional review within SKB and that a higher level of metallurgical support is required. We disagree that suitable full size canisters have been created and that production technology is available for both canisters at full size. We also disagree that the long-time durability is ascertained. I.a. it is easy to find corrosion mechanisms for the canister system that have to be demonstrated not to be harmful. We feel there are many areas which need further evaluation, i.a. effects of non uniform loading and creep, effects of departure from circularity, welding, quality control, effects of radiolysis, corrosion properties, etc. We also feel that insufficient emphasis has been placed on the further development on high power electron beam welding, machining, casting of the insert, testing and overall handling. We consider that more information should be provided on the detail and timing of the development plan for the trial fabrication programme of the canister, test programme, determination of quality standards and development of non destructive testing procedures. 11 refs

[en] SITE-94 is a performance assessment of a hypothetical repository at a real site. The main objective of the project is to determine how site specific data should be assimilated into the performance assessment process and to evaluate how uncertainties inherent in site characterization will influence performance assessment results. This report uses scenario definition work as a vehicle to introduce the systems approach to performance assessment which has been developed and tested in SITE-94 and which constitutes one of the main advances made during the project. The results of the application of the methodology are presented separately, in the SITE-94 Summary Report. 40 refs, 21 figs, 12 tabs

[en] This study has revealed a number of joining processes to be used when manufacturing copper-canisters for the final storage of high level nuclear waste. However, the decision on which material and which joining process to be used has to be based on the design criterions. The welding procedure has to be qualified, i.e. it shall be demonstrated whether the procedure is capable of fulfilling specified requirements. 32 refs