[en] The paper presents fuel management practices at Ignalina Nuclear Power Plant (NPP) with two RBMK-1500 units. Present thermal rating of each unit is 4200 MW. Changes in fuel management practices and reactor operation following the Chernobyl accident are described. Special emphasis is given to the reduction of void reactivity coefficient by increase of the number of control rods and transfer to the integrated burnable absorber (UO₂-0.41% Er₂O₃) fuel with increased (up to 2.4%) enrichment. (author)

[en] Analysis of nuclide composition of Chernobyl NPP releases (inert gases) showed that Ar-41 plays there the main role (84-90%). For its decrease a system for gas tank blow-through by nitrogen of high purification was proposed

[en] This paper briefly reviews the principles involved in the planning and design of the proposed facility to cater for seismic and structural loads. The conceptual layout is discussed, as well as the different load characteristics and scenarios. An outline is given of model used to estimate the seismic loads, whereafter the different analytical models are discussed. (author)

[en] This paper describes present status of initial core fuel fabrication for the HTTR in Japan. The High Temperature Engineering Test Reactor (HTTR) is being constructed satisfactorily, aiming the first criticality in October 1997. Fuel fabrication for the HTTR was started in June 1995 at Tokai works of Nuclear Fuel Industries, Ltd. (NFI). The fabrication technologies have been developed through cooperation works of NFI and Japan Atomic Energy Research Institute. The initial core fuel of the HTTR consists of 150 fuel assemblies. An HTTR fuel assembly is so-called a pin-in-block type of hexagonal graphite block containing 31 or 33 fuel rods. A fuel rod consists of a graphite sleeve and of 14 fuel compacts. In a fuel compact about 13,000 TRISO coated fuel particles are dispersed densely. Fabrication of the fuel is being carried out at NFI's facility, and a half of the initial fuels have been completed. The results of inspection for the size, the density and the sphericity of coated fuel particle, and the as-manufactured particle failure fractions in fuel compacts, which are most important inspection items from the view point of quality assurance, are quite satisfactory. Transportation and assembling of the fuel rods will be carried out at the reactor site from the middle of 1997. (author)

[en] The purpose of this study is to investigate the use of erbium as burnable poison in RBMK reactors. The neutronic code APOLLO2 has been used and a comparison with the Monte-Carlo code TRIPOLI2 has been made. The first chapter briefly presents the RBMK characteristics, the second chapter deals with the neutronic behaviour of a fuel assembly in an infinite lattice which is an important step in the modelling process. The third chapter presents the analysis of the use of erbium in typical elements of the RBMK lattice. A good agreement is obtained between the 2 codes except in the draining situations. Erbium appears to reduce the positive reactivity effect of the draining configuration. (A.C.)

[en] The scope of the seismic response analysis covers the assessment of the Main Building response to the seismic input excitation and the assess the effects of site-specific soil conditions on the dynamic structural response of the building structure. The study included development of 3-D stick model of main building structure; site-specific soil conditions (low-strain soil properties, median estimate, soil data) and seismic excitation time histories (acceleration time histories developed for Kozloduy NPP, scaled to 0.1 g). The Study performed includes: deconvolution of seismic free field surface motion; modeling of structure foundation and generation of impedance and scattering functions; fixed base frequencies and mode shapes extraction; soil-structure interaction analysis using CLASSI chain of computer programs; floor acceleration time histories generation and floor response spectra generation

[en] In the present report, the problem of qualification procedures of electrical equipment with respect to the dynamic excitation subsequent to an aircraft impact (ACC) on a Nuclear Power Plant (NPP) is approached, within the context of IAEA Benchmark on vulnerability of equipment and structures of RBMK-type NPP against the aircraft impact. After a short description of the main objectives of the work and the relevant area of concern (Chapter 1), the safety related equipment more commonly installed in a NPP are grouped in few classes, according to widely accepted classification criteria and the relevant failure modes are described (Chapter 2). Taking as reference a deeply studied RBMK reactor (Ignalina NPP), an overview of its main characteristics and of the equipment ensemble housed in is given in Chapter 3. An overview of the worldwide most used qualification standards for safety related equipment for NPPs is reported in Chapter 4, and a comparison of the practices used in Europe for the qualification of safety related electrical and I and C equipment is described with special attention to seismic and impact qualification (Chapter 5). In the hypothesis that the equipment to qualify against impact excitation has been already qualified against seismic excitation, the problems relevant to the different nature of earthquake and shock phenomena are listed, together with the main criteria to implement a procedure which, based on standardized shock pulses, could be applied for ACC qualification purposes (Chapter 6). Consequently, a possible ACC qualification procedure is outlined (Chapter 7) and the interface data (data coming from numerical analysis and seismic qualification, to be used for ACC qualification purposes) are listed (Chapter 8). Finally, the main conclusions of the work are described (Chapter 9). The main references are listed in Chapter 10. (author)

[en] The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

[en] The Barselina Project was initiated in the summer of 1991. The project is a multilateral co-operation between Lithuania, Russia and Sweden. The long range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. The Swedish BWR Barsebaeck is used as reference plant and the Lithuanian RBMK Ignalina as application plant. During phase 3, from March, 1993 to June, 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 (INPP-2) was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. To increase the realism of the risk model a set of deterministic analyses were performed and plant/RBMK-specific data bases were developed and used. A general concept for analysing this type of reactor was developed. During phase 4, July 1994 to September 1996, the PSA was further developed, taking into account plant changes, improved modeling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The updated model is quantified and new results and conclusions are evaluated

[en] The Modular Helium Reactor (MHR) has been identified as a highly promising advanced nuclear energy system that enables applications which are not possible with other nuclear technologies. Among these, a particularly attractive option is direct utilization of the helium primary coolant as the working fluid in a closed gas-turbine (Brayton) cycle. The resulting Gas-Turbine MHR (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. The recently completed evaluation of the GTMHR has resulted in the recommendation that the direct cycle GT-MHR be identified as the commercial objective of the US Gas-Cooled Reactor Program, replacing the SC-MHR. That recommendation has been accepted by both participants and sponsors. The resultant redirection to the GT-MHR represents a major turning point in the evolution of the gas-cooled reactor. While the development and demonstration of the GT-MHR represents a significant challenge, the inherent and passive safety characteristics, high efficiency and improved economics of the GT-MHR offer an attractive and competitive nuclear generation option for the early 21st Century