[en] An alternative method to reconstruct the source term of a nuclear accident is proposed. The technique discussed here involves the use of in-situ gamma spectrometry. The validation of the applied methodology has been possible through the monitoring of routine releases of Ar-41 originating at a Belgian site from an air cooled graphite research reactor. This technique provides a quick nuclide specific decomposition of the source term and therefore will be have an enormous potential if implemented in nuclear emergency preparedness and radiological assessments of nuclear accidents during the early phase

[en] Short communication

[en] To make a valid contribution to European Fusion objectives, the NET (Next European Torus) must have reactor-like size and operating conditions. This paper proposes for NET; a full size integrated tritium system concept in sufficient detail to permit identification of critical development issues and to serve as a basis for relevant safety and reliability analysis. The proposed options were selected based on the following criteria: (a) minimum risk due to accidents, (b) minimum environmental impact in normal operation, (c) ability to handle a wide range of operating parameters, (d) simplicity and maintainability, (e) required technology development, (f) efficience and cost, (g) relevance to commercial energy production

[en] As nuclear power plants become older, in-vessel maintenance to high radiation components will become more common. This paper describes General Electric's experience in making such repairs on Boiling Water Reactors and how the use of engineering and project management enhances the productivity of special maintenance projects. Design techniques employed to minimize personnel radiation exposure are discussed. Specialized remote, automatic and semi-automatic equipment designs are presented. Included are sample descriptions of special tools developed, problems encountered, and lessons learned

[en] An improved understanding of boiling film dynamics as it relates to energetic steam explosions during degraded core conditions in light water reactors is developed. Several models have been developed and used to predict the characteristics of film boiling when a molten fuel drop suddenly comes into contact with water. An incompressible model and an approximate compressible model, utilizing Gilmore's equation are developed consistent with past works and are determined to have several shortfalls. To improve the treatment of compressibility effects, a model employing Lagrangian equations is developed. This improved model predicts that applying an external pressure pulse can make a stable film go unstable and decreasing water subcooling stabilizes film oscillations; both predictions are consistent with experimental observations. However, the improved model predicts stable film boiling at low melt temperatures that cannot support such boiling. Modeling Taylor surface instability effects at the water/steam interface indicates that the surface area change due to this surface instability can stabilize the film oscillations

[en] The problem of analyzing and quantifying common-cause failures (CCFs) is fundamental in conducting probabilistic safety studies. The use of experience feedback is crucial in addressing this problem, and it is possible to draw in a complementary manner on national experience feedback data bases and ad hoc on-site audits. These two approaches differ in objectives and methods, but are both equally necessary. In this paper, the use of these two approaches and the results obtained are analyzed and are illustrated by two concrete examples: (1) a national collection of >1200 event and failure records concerning the problem of CCFs in EdF 900-MW (electric) units; and (2) an analysis based on audits carried out at an EdF 1300-MW(electric) reactor site

[en] The accident at Three Mile Island Unit 2 (TMI-2) created circumstances that have seldom been encountered in the nuclear industry. These circumstances involved high radiation and contamination conditions in areas to which emergency access was required to maintain reactor control. The radiological conditions in the auxiliary building immediately following the accident, how those conditions varied from one location to another, and how they varied with time during the first 6 to 8 months required to return to some degree of normalcy are reviewed. The controls and methods used to maintain radiological exposures as low as reasonably achievable (ALARA) while the necessary work was completed, as well as the circumstances that led to non-ALARA exposures, are also described

[en] Prior to the accident at Three Mile Island, the radiological control training was provided by the health physics and the chemistry departments. The programs that required the input from the health physics department were the Operations Training and the General Employee Training Programs, which were formally organized and presented radiological controls practices as part of the course objectives. The operations training section periodically requested the health physics department management to provide a radiological portion to their operator qualification or operations requalification class. After the accident in 1979, the General Employee Training Program administration was taken over by the operation training department where it underwent changes with a dedicated effort. The Operations Training Program also underwent a change as well as the Radiological Control Technician Training Program. Each group now has a specialty program designed for their specific needs. These programs have been designed using the Institute of Nuclear Power Operations and regulator guidelines

[en] Planning for the disassembly and defueling of the Three Mile Island Unit 2 (TMI-2) reactor lower core support assembly (LCSA) began early in 1985. Evaluations of methods of defueling were performed based on various assumed LCSA conditions. Tooling was conceptualized and various defueling sequences were evaluated. As defueling of the core region progressed, information and data were obtained that clearly showed that a large amount of fuel was located within the LCSA and an even larger amount was located below the LSCA in the lower head of the vessel. It became apparent that it would be impractical to defuel this area of the reactor without massive cutting of the structural steel internal components. A plan was developed to defuel the LCSA by using two techniques: (a) core-boring LCSA components that had a round configuration (support posts and in-core instrument guide tubes) and (b) plasma arc cutting of components, which required linear cutting (plates and forgings). This paper describes the approach to LCSA disassembly, including the design basis of the equipment and sequence of in-vessel operations

[en] The concept of probabilistic safety analysis (PSA) is defined and its components are given. The differences between probabilistic and deterministic studies are explained. The state of the art in PSA methods is outlined with emphasis on the following aspects: plant analysis, containment analysis and accident sequence development, fission product transport, environmental and economic consequence analysis, and external events. (P.A.) 1 fig., 3 refs