[en] Multidimensional thermal-hydraulic code calculations are necessary to predict complex thermal-hydraulic phenomena within the downcomer during LBLOCA calculation. The main objective of present study is to evaluate the TRACE code capability with DVI related Upper Plenum Test Facility Data, which is on the developing by US-NRC. Another objective is to develop the APR-1400 plant standard input deck preparing the LBLOCA calculation using TRACE code. TRACE input decks for UPTF Test 6,7,21 was developed and TRACE code was evaluated with theses input decks. The result of comparison on lower plenum penetration rate with data showed that TRACE code predict reasonable lower plenum penetration rate. On the calculation of Test 21(Run 272) and Test 5(Run 062), high subcooled water is injected, TRACE code overprediced the bypass rate. Analysis showed that the condensation model in TRACE code overpredict the condensation of steam within upper downcomer region. It is identified that the condensation model in TRACE code needs improvement. TRACE code input deck for APR-1400 was developed and preliminary LBLOCA case was calculated. The result was very sensitive to its input deck change in terms of peak cladding temperature. it showed that the quality assurance during input deck development is crucial especially in the best estimate thermal-hydraulic calculation

[en] Thermal-hydraulic response in the discharge piping at the upstream of water pool following the opening of the safety relief valve is analyzed. To predict the basic pressure wave propagation and interaction with reflection wave, the RELAP5/MOD3 code is used. Pressure wave propagation behavior in a simple geometry is calculated and the effect of the important parameters including the loss factor, the pipe configuration, the water slug inflow, the valve opening time, and subdivision of sparger are investigated. And the affecting factors influencing the pressure wave propagation and their mechanisms are discussed

[en] ATWS (Anticipated Transients Without Scram) are anticipated operational occurrences accompanied by the failure of the reactor trip portion of the reactor trip system. ATWS accidents are an cause of concern because under certain postulated conditions they could lead to significant core damage including core melt and to the large release of radioactivity to the environment. In this study, considerations on reduction of risk from ATWS were discussed with examination of the technical background of 10CFR 50.62. Considering the recent trends of the extended core cycle and the power uprating, it is recognized that the moderator temperature coefficient can become less negative than to suppress the RCS overpressure followed by ATWS. Because the negative reactivity feedback is one inherent level of multiple defenses, the effect against the RCS overpressure needs to be assessed in detail

[en] The primary function of the reactor containment is to serve as an essentially leaktight barrier that will trap any radioactive materials that may leak from the primary system and prevent their release to the atmosphere. To accomplish this task, the containment must be capable of withstanding the pressurization of a hypothesized accident with a quite small leak rate. In Korea, the containment leakage tests are periodically conducted in compliance with the technical specification of each nuclear power plant. Now, the development of domestic testing requirements is in progress to improve the relevant regulations by adopting performance-oriented approaches. This paper summarizes the development status and the regulatory assessments focused on applicability of the performance based containment leakage test requirement developed by US NRC

[en] This study provides the working items and licensing issues required for the applications of multi-dimensional reactor kinetics methodology to accident analysis. The multi-dimensional reactor kinetics model takes further more uncertain parameters and kinetics parameters compared with the point kinetics model, and thus more various and systematic uncertainty analysis and sensitivity analysis can be required. It needs that input parameters used in a accident code be simplified and validated by quantifying their effects through theses analyses

[en] The applicability of a 3-dimensional reactor kinetics model to the rod ejection accidents was examined in the view of the enthalpy rise in the fuel rod. PARCS code was used for the 3-dimensional reactor kinetics mode. As a result of the various parametric analysis, the values of 'maximum enthalpy rise' for 'ρ_rod(ejected rod worth)- β(delayed neutron fraction)' were obtained, and it could be expressed as a linear curve for the complicated and various reactor design and operation conditions. If the theoretical basis of this linear curve can be verified for all the loading patterns and the operation conditions, it will be an index in the regulatory evaluation for the validation of the 3-dimensional reactor kinetics analysis for the rod ejection accidents

[en] Development of the specific technical standards of the evaluation method for Emergency Core Cooling System (ECCS) performance in PWR has been requested since issuance of the MOST Notice 2001-39. Korea Institute of Nuclear Safety (KINS) developed a regulatory guide on the conservative ECCS evaluation method and one on the best-estimate ECCS evaluation method based on the standards and evaluation method in USA. The issue of Direct Vessel Injection of Advanced Pressurized Reactor (APR) in Korea and the review experience including KEPRI Realistic Evaluation Method (KREM) have been fed back into the development. Comments and opinions from the experts in the industry and research were solicited and implemented into the affected regulatory guides as appropriate. Those guides will be finalized through the review of KINS committee on regulatory rule and is expected to apply to the related safety review

[en] In PWR the rod cluster control assembly (RCCA) for shutdown is released upon the action of the control drive mechanism and falls down through the guide thimble by its weight. Drop time and impact velocity of the RCCA are two key parameters with respect to reactivity insertion time and the mechanical integrity of fuel assembly. Therefore, the precise control of the drop time and impact velocity is prerequisite to modifying the existing design features of the RCCA and guide thimble or newly designing them. During its falling down into the core, the RCCA is retarded by various forces acting on it such as flow resistance and friction caused by the RCCA movement, buoyancy mechanical friction caused by contacting inner surface of the guide thimble, etc. However, complicated coupling of the various forces makes it difficult to derive an analytical dynamic equation for the drop time and impact velocity. This paper deals with the development of a computer program containing an analytical dynamic equation applicable to the Korean Fuel Assembly (KOFA) loaded in the Korean nuclear power plants. The computer program is benchmarked with an available single control rod drop tests. Since the predicted values are in good agreements with the test results, the computer program developed in this paper can be employed to modify the existing design features of the RCCA and guide thimble and to develop their new design features for advanced nuclear reactors. (author)

[en] Male rats were exposed to a whole body with a single dose of 4.0Gy. The animals were sacrificed 68 hours following exposure. The specific amine (ο-dianixidine) oxidase activity of ceruloplasmin purified from the γ-irradiated rat serum was not different from that of normal ceruloplasmin. On the other hand, the specific ferroxidase activity of ceruloplasmin from the γ-irradiated rat serum was 2.5 fold as high as compared to that of normal ceruloplamsin. The copper content in ceruloplasmin was also changed from 5.9 copper/molecule to 4.5 copper/ molecule. It may be that irradiation induces the structural change of ceruloplasmin in the biosynthetic process. The ferroxidase activity of ceruloplasmin prevents free Fe²⁺, which can be released by γ-irradiation, from producing oxygen free radicals in the presence of H₂O₂

[en] Performance evaluation was executed for each component and its assembly for the deduced Top Nozzles to develop the new Top Nozzle for LWR. This new Top Nozzle is composed of the optimum components among the derived Top Nozzles that have been evaluated in the viewpoint of structural integrity, simpleness of dismantle and assembly, manufacturability etc. In this study, the developed Top Nozzle satisfied all the related design criteria. In special, it makes fuel repair time reduced by assembling and disassembling itself as one body, and improves Fuel Assembly holddown ability by revising the design parameters of its spring and the structural integrity through the betterment of its geometrical shpae of Flange and Holddown Plate as compared with the existing LWR Top Nozzles