[en] The report is structured as follows: (i) Description of the problem (phenomena occurring inside the grain and on the grain boundary, modelling of bulk gas diffusion through the lattice of an equivalent shape grain); (ii) classical mathematico-physical models of fission product release (general view upon the 'classical' models; mechanistic models; links between the intragranular and intergranular models; effect of grain growth); (iii) Fission gas release models in practical uses (Lassmann's model; Massih's model; Strizhov-Valach-Zymak model; 'standard' calculation of proportional fission product release from oxide fuel); (iv) Critical analysis of the fission gas bubble evolution theory; (v) Release modelling by the finite element method (Matthews-Wood model; Ito-Iwasaki-Iwan model); (vi) Summary of drawbacks of modelling and differences between the theoretical description and some experimental observations. (P.A.)

[en] An overview is presented of inspection practices and safety requirements during the operation of nuclear power plants with emphasis on the refuelling process. The report includes: (i) requirements for nuclear power plant quality and licensing; (ii) regulatory body requirements during the power plant operation and the refuelling process; (iii) procedures for inspections and inspection programmes. Experience from the following countries was gained: Belgium, Finland, France, Hungary, Italy, Japan, Mexico, The Netherlands, Spain, Sweden, UK and USA. (author)

[en] The data and assumptions underlying the first draft guidelines for refining safety criteria applicable to PWR fuel in conditions of a maximum credible accident are given. A survey of safety criteria applied abroad is presented, and applicable regulations valid in Czechoslovakia are referred to. The draft guidelines include basic terms, underlying assumptions and data, binding regulations and fuel element failure criteria. Permissible values of quantities to be used in the criteria are proposed, based on available technical knowledge. The most important physical models applicable to nuclear fuel and needed for the evaluation of the safety criteria are defined. Also included is a description of the required documentation procedures. (author). 37 refs

[en] High temperature gas cooled reactors (HTGR's) are discussed as prospective nuclear sources capable of replacing fossil fuels. Research programs are described of HTGRms development in some countries, such as the UK, the USA, the FRG, Japan, France. Variants such as very high temperature reactors, the development and improvement of nuclear fuel and the fuel cycle for HTGR's, safety of HTGR's including design-basis accidents, and HTGR economics are also discussed. (H.S.)

[en] The results are presented of a blind calculation of CORA-W2 Experiment by modified ICARE-V2mod1 performed at the Nuclear Research Institute, Rez in the frame of the OECD-CSNI International Standard Problem (ISP) No. 36. The objective of the ISP-36 was to analyze the heatup and meltdown phase of the CORA-W2 Experiment (containing a WWER-1000 test bundle) performed at KfK Karlsruhe. The thermohydraulic, heat transfer, and chemical interaction models are briefly described used in the ICARE2-V2mod1. Code modifications, nodalization of the CORA-W2 bundle, input parameters and other modeling assumptions are also described. The main code modification consists in the implementation of two recently developed chemical interaction models. Calculated results are presented in the form as required by ISP-36 Specifications. 52 figs., 19 refs

[en] Following up the analytical approach to mathematico-physical modelling of mechanisms contributing to gaseous fission product release from high-burnup fuel pellets into the void volume of the fuel pin during operation in a PWR (including VVER), the report presents a brief description of an innovated model of the transport of gaseous fission products. Validation and subsequent practical application of this method to safety analyses is envisaged

[en] Within the partial project aimed at testing the predictive power of the TRANSURANUS code modified for VVER fuel, calculations of the maximum temperature in the centre of the fuel and of the maximum pressure inside the most stressed fuel pins (RSAC parameters) during the planned operation of the Dukovany-2 reactor unit were performed by using the TRANSURANUS v1m2j99 code. The results were compared with those obtained by means of the PIN99W and RODQ2D integral codes. (author)

[en] The main changes performed within the innovation of the PIN-micro code to the PIN99w version are highlighted. The most recent version of the code describes steady-state LWR fuel behaviour up to a very high degree of burnup

[en] The results of FUMEX-II 'must' cases obtained by two codes, viz. FEMAXI-V and PIN2K are summarized. The calculated temperature and FGR from the 2 codes are compared with observed values (where available). (P.A.)

[en] The applicability is tested of program PIN (derived from program GAPCON-THERMAL-2) in the thermomechanical modelling of light water reactor fuel elements. Four variants of fuel elements designated A through to D were calculated using the PIN. Variant B consists of two fuel elements with a different irradiation history. The elements were irradiated in different reactors. The experimental data (temperature in the middle of the fuel, the amount of fission products, axial and radial can deformations, fuel-can gap) are shown in tables and graphs. (H.S.)