[en] An assessment of sixteen void fraction correlations have been carried out using experimental void fraction data compiled from open literature for vertical upward steam-water flow. Nearly 80% of all the data pertained to natural circulation flow. This assessment showed that best prediction is obtained by Chexal et al. (1996) correlation followed by Hughmark (1965) and the Mochizuki and Ishii (1992) correlations. The Mochizuki-Ishii correlation is found to satisfy all the three limiting conditions whereas Chexal et al. (1996) correlation satisfies all the limiting conditions at moderately high mass fluxes (greater than 140 kg/m²s) while Hughmark correlation satisfies only one of the three limiting conditions. The available void fraction data in the open literature for steam-water two-phase flow lies predominantly in the low quality region. This is the reason why correlations like Hughmark which do not satisfy the upper limiting condition (i.e. at x=1, α=1) perform rather well in assessments. Additional work is required for the generation of high quality (greater than 40%) void fraction data. (author)

[en] This report deals with the neutron dosimetry at the accelerator environment with a special emphasis given to the neutron skyshine problem. Theoretical studies have been carried out for the estimation of neutron skyshine dose using experimentally measured neutron spectra as well as those obtained using theoretically simulated nuclear reaction model calculations. A detailed analysis has been performed with respect to several variables. (author)

[en] We present preliminary results for the measurement of [1/σ]dσ/dp_T for the Z boson observed in the e⁺e^- channel for p_T < 50 GeV/c. The data are from a luminosity of ∼ 90 pb^-1 collected with the D0 detector during the 1994-1995 Tevatron run. The differential spectrum is sensitive to non-perturbative predictions of QCD. 3 refs., 1 fig

[en] Over 8500 reactor-years of operating experience have been accumulated with the current nuclear energy systems. New generations of nuclear power plants are being developed, building upon this background of experience. During the last decade, requirements for equipment specifically intended to minimize releases of radioactive material to the environment in the event of a core melt accident have been introduced, and designs for new plants include measures for preventing and mitigating a range of severe accident scenarios. The IAEA Technical Committee Meeting on Impact of Severe Accidents on Plant Design and Layout of Advanced Water Cooled Reactors was jointly organized by the Department of Nuclear Energy and the Department of Nuclear Safety to review measures which are being incorporated into advanced water cooled reactor designs for preventing and mitigating severe accidents, the status of experimental and analytical investigations of severe accident phenomena and challenges which support design decisions and accident management procedures, and to understand the impact of explicitly addressing severe accidents on the cost of nuclear power plants. This publication is intended to provide an objective source of information on this topic. It includes 14 papers presented at the Technical Committee meeting held in Vienna between 21-25 October 1996. It also includes a Summary and Findings of the Working Groups. The papers were grouped in three sections. A separate abstract was prepared for each paper

[en] The US Department of Energy (DOE) is sponsoring a programme in technology development aimed at resolving the technical issues in severe accident management strategies for advanced and evolutionary light water reactors (LWRs). The key objective of this effort is to achieve a robust defense-in-depth at the interface between prevention and mitigation of severe accidents. The approach taken towards this goal is based on the Risk Oriented Accident Analysis Methodology (ROAAM). Applications of ROAAM to the severe accident management strategy for the US AP600 advanced LWR have been effective both in enhancing the design and in achieving acceptance of the conclusions and base technology developed in the course of the work. This paper presents an overview of that effort and its key technical elements

[en] A one-dimensional mathematical model has been developed for a 4250 MW(th) Advanced Pressurized Water Reactor containment analysis following a severe accident. The cooling process of the composite containment-steel shell and concrete shield- is achievable by natural circulation of atmospheric air. However, for purpose of gettering higher degrees of safety margin, the present study undertakes two objectives: (1) Installment of a diesel engine-driven air blower to force air through the annular space between the steel shell and concrete shield. The engine can be remotely operated to be effective in case of station blackout. (ii) Fixing longitudinally plate fins on the circumference of the inside and outside containment steel shell. These fins increase the heat transfer areas and hence the rate of heat removal from the containment atmosphere. In view of its importance - from the safety viewpoint - the long term behaviour of the containment which is a quasi-steady state problem, is formulated through a system of coupled nonlinear algebraic equations which describe the thermal-hydraulic and thermodynamic behaviour of the double shell containment. The calculated results revealed the following: (i) the passively air cooled containment can remove maximum heat load of 11.5 MW without failure, (ii) the effect of finned surface in the air passage tends to decrease the containment pressure by 20 to 30%, depending on the heat load, (iii) the effect of condensing fins is negligible for the proposed fin dimensions and material. However, by reducing the fin width, increasing their thickness, doubling their number, and using a higher conductive metal than the steel, it is expected that the containment pressure can be further reduced by 10% or more, (iv) the fins' dimensions and their number must be optimized via maximizing the difference or the ratio between the heat removed and pressure drop to get maximum heat flow rate

[en] CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

[en] A hot cell facility has been recently commissioned in the Radiochemistry Laboratory, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam with the objective of carrying out post - irradiation studies on fast reactor fuels as well as PHWR fuels. The hot cell facility has several novel features such as a double containment system with high purity argon atmosphere in the containment box, low probabilities of contamination or radiation exposure to operating personnel and economy in both capital and operating costs. This report describes the different concepts used to achieve a high degree of leaktightness in the stainless steel containment boxes as well as in the cells and the development of special gaskets and O-rings like formed gaskets and co-seals. (author)

No abstract available

No abstract available