[en] The technique of simulation of heat-physics, strength and safety characteristics of reactor RBMK and WWER rods under steady-state, transient and accident conditions is presented. That technique is used in mechanic and heat physics codes PULSAR-2 and STALACTITE. Simulation in both full scale and the most stress-loading part of cladding statement under accident conditions are considered. In this zone local swelling and cladding failure are possible. The accident simulation is based on the mechanical creep-plasticity problem solution in three-dimensional approach. The local cladding swelling is initiated with determining of little hot spot on the clad with several degrees temperature departure from average value. Mechanical problem is solved by finite elements method. Interaction of Zr with steam is taken in to account. Fuel and cladding melting, shortness and dispersion formation processes are simulated under subsequent rods warming up. (author). 2 refs., 6 figs

[en] A program of analysis which was developed to evaluate the structural integrity of PWR fuel assemblies is presented. The static and dynamic structural models implemented in computer programs are discussed and the experimental program is briefly presented. (author). 12 refs

[en] Indian Fast Breeder Test Reactor uses 70% Pu C-30% UC as fuel. Analytical estimates have been made to find the attainable burnup for this fuel. Data on fuel and clad material are very crucial in the analysis. Analysis has been performed for those subassemblies which are representative of entire core, using the data available and the information found in literature, for nominal and hot spot temperature conditions. It is found that the burnup potential of the fuel is only about 70% of what is achievable under nominal temperature conditions and the burnup attainable at fuel middle column is the limiting one compared to fuel top column. (author). 8 refs., 3 figs., 1 tab

[en] The main phenomena which influence an LMFBR core seismic response are the fluid structure interaction and the impacts between subassemblies. To study the core behaviour seismic tests and calculations have been performed on the core mock-up RAPSODIE in air or in water and for different excitation levels. (author). 2 refs., 6 figs

[en] Ultrasonic testing method for the nuclear grade graphite was established. The effectiveness of this method was demonstrated through the application for the graphite component of the HTTR. The characteristics of noise echo were also investigated by relating with mechanical properties. (author). 2 refs., 6 figs., 3 tabs

[en] This paper presents the current status of the national program on radioactive waste management in the Republic of Korea especially in the field of site selection, characteristics of radioactive wastes, conceptual design of the proposed repository and the future project plan. (author). 5 refs., 5 figs., 1 tab

[en] The application of the Leak Before Break (LBB) concept to Liquid Metal Reactor (LMR) components is sometimes delicate due to the low level of primary loads (pressure, gravity) compared to thermal gradients. This is unfavorable for the flow rate. When applying procedures based on simplified analytical methods, the calculated flow rate for a critical through-wall defect is often very low and the conclusion is that the component does not verify the LBB concept. FRAMATOME-NOVATOME has undertaken several studies in order to take into account more accurately the effects of blending stresses due to an axial thermal gradient and residual welding stresses in the evaluation of the leak are for a through-wall circumferential defect in a cylindrical structure, typical of LMR's vessel. The main results issued from three-dimensional finite element calculations are the following: for a welding process giving tension on the skins of the vessel and compression in the inner part, the effect of residual stresses on the leak are is negligible; for a long defect, the leak area given by simplified analytical methods is pessimistic. All these results allow the verification of the LBB concept for the studied case. They will contribute to the development of LBB methodology for LMR components. (author). 5 refs., 4 figs

[en] Modal properties namely, natural frequency and damping factor of the thermal baffles of the 500 MW(e) Prototype Fast Breeder Reactor were measured on a 1/16 scale model. Impulse excitation method was used and experiments were carried out with different levels of water in the annulus between the two baffles. Value of each modal frequency was found to be nearly independent of water level whereas the damping ratio showed a decreasing trend with lower level in the annulus. (author). 2 refs., 5 figs

[en] A general approach to the dynamic time-history analysis of the coupled internals and core in the vertical direction is presented as a part of the fuel assembly qualification program. A fuel assembly of a pressurized water reactor is not restrained mechanically in the vertical direction and it necessary to check an uplift for a large dynamic motion. Dynamic analyses are performed for the excitations induced from earthquake and pipe break. The peak responses such as fuel assembly axial forces and lift-off characteristics are investigated. (author). 6 refs., 6 figs., 2 tabs

[en] The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs