Results 1 - 10 of 14
Results 1 - 10 of 14. Search took: 0.018 seconds
|Sort by: date | relevance|
[en] Some nuclear power plants in Fukushima Daiichi site had lost their emergency reactor cooling function for long-time so the fuels inside the reactors were molten, and the integrity of containment was damaged. Therefore, large amount of radioactive material was released to environment. Because the social and economic effects of severe accidents are enormous, Korean Government already issued 'Severe Accident Policy' in 2001 which requires nuclear power plant operators to set up 'Quantitative Safety Goal', to do 'Probabilistic Safety Analysis', to install 'Severe Accident Countermeasures' and to make 'Severe Accident Management Plan'. After the Fukushima disaster, a Special Safety Inspection was performed for all operating nuclear power plants of Korea. The inspection team from industry, academia, and research institutes assessed Korean NPPs capabilities to cope with or respond to severe accidents and emergency situation caused by natural disasters such as a large earthquake or tsunami. As a result of the special inspection, about 50 action items were identified to increase the capability to cope with natural disaster and severe accidents. Nuclear Safety Act has been amended to require NPP operators to submit Accident Management Plant as part of operating license application. The KINS Severe Accident Regulation Preparation TFT had first investigated oversea severe accident regulation trend before and after the Fukushima accident. Then, the TFT has developed regulation draft for severe accidents such as Severe accident Management Plans, the required design features for new NPPs to prevent severe accident against multiple failures and beyond-design external events, countermeasures to mitigate severe accident and to keep the integrity of containment, and assessment methodology on safety assessment plan and probabilistic safety assessment
[en] As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code
[en] The two popular transverse integrated nodal methods (TINMs), the nodal expansion method (NEM) and analytical nodal method (ANM), and the analytic function expansion nodal (AFEN) method are integrated into a single unified nodal formulation for the space-time kinetics calculations in rectangular core geometry. In particular, the nodal coupling equations of the conventional ANM and AFEN method are reformulated by the matrix function theory based on the unified nodal method (UNM) principle for the solution to the transient two-group neutronics benchmark problems. The difference between the two transient AFEN formulations by the UNM and the conventional AFEN principles is pointed out. The performance of the UNM formulation is examined in terms of the solutions to the transient light water reactor benchmark problems such as the Nuclear Energy Agency Committee on Reactor Physics pressurized water reactor rod ejection kinetics benchmark problems. Through comparison of several nodal computational options by the UNM formulation, it is shown that one node-per-fuel assembly (N/A) calculations by the AFEN method are superior to those by the NEM and the ANM, but that 4 N/A calculations by the AFEN method are not better than those by ANM, in prediction accuracy at the sacrifice of the computational time. The advantages of the transient UNM formulation over the conventional TINM and AFEN method formulations are discussed
[en] This paper provides technical backgrounds for the regulatory requirements for an AMP and discusses the anticipated issues for its implementation. To legislate for the regulatory control of accident management including severe accident management, the Nuclear Safety Act (NSA) was amended in June 2015. As the effective date of the amendment of the NSA was set to be the 23rd of June 2016, the subsequent rulemaking for the implementation of the amendment of the NSA was completed by June 2016 and the regulatory framework on accident management including the management of severe accident is currently in effect. As required by the amended NSA, the applicant for operating license of a nuclear power plant (NPP) shall submit an accident management program (AMP) and its contents shall meet the pertinent regulatory requirements which were set forth by recently completed rulemaking efforts. Major contents of recently completed legislation of accident management are introduced and their technical backgrounds are discussed. The new regulations for the AMP requires various evaluations and assessments to assure that the severe accident is appropriately prevented and, should it occurs, its consequence is appropriately mitigated enough to protect people and the environment. For development of an AMP, a few anticipated issues are introduced, which are mainly related to the absence of specific safety review guidelines. These guidelines are to be developed by the end of 2017 and that can enhance the consistency of the regulation of AMPs.
[en] Hydrogen generation, distribution and combustion in containment during accident conditions are very complex and highly plant- and scenario-specific phenomena. Moreover, hydrogen combustion can take place in a variety of forms likely mild deflagration, fast or accelerated flames, deflagration to detonation transition (DDT) and detonation. The distribution of the hydrogen released within the containment determines local and global hydrogen concentrations, which are decisive for the evaluation of the various combustion modes, such as diffusion flames, deflagration and detonation, depending on geometrical effects and concentrations. In terms of implementing effective hydrogen management measures, an understanding of all these phenomena is crucial. The purpose of this paper is focused to review the current safety perspectives for hydrogen control and management in the countries aimed at enhancing severe accident management strategies. This study provides a summary of the status of knowledge on hydrogen control and management strategies and measures implemented by the major countries. It is observed that all countries have taken significant steps to improve the safety of their plants with various degrees of practical implementation, in particular for hydrogen control and management, high priority must be given to installing means of hydrogen mitigation designed for severe accidents so as to practically eliminate containment failure due to hydrogen combustion.
[en] Interface control activities during the nuclear power plant (NPP) construction and operation have been reviewed for enhancing the safety of NPP. The primary focus of the study is given on analysis of lessons learned from the recent significant events of Korean Standard Nuclear Power plant (KSNP), such as a series of break-off of thermal sleeves at YGN 5 and 6 and radioactivity leak at YGN 5, in respect of interface control. Based on the results of the analysis, this study recommends measures for the improvement of interface control among utility and technical supporting organizations (TSO), and suggests new regulatory systems, such as reporting of safety significant non-conformances, to effectively verify the adequacy of interface control activities during construction and operation of NPPs
[en] As an efficiency enhancement numerical scheme of transient nonlinear nodal calculations, a three-grid correction scheme (3GCS) using a modified W cycle based on three grid structures of three-dimensional (3-D) four-node-per-assembly (4N/A), 3-D 1N/A, and two-dimensional (2-D) 1N/A is developed. Its computational efficiency is compared with a single-grid biconjugate gradient stabilized (BICGSTAB) iteration scheme in popular use in terms of 3-D 4N/A nonlinear analytical nodal method solutions to Nuclear Energy Agency Committee on Reactor Physics pressurized water reactor rod ejection benchmark problems. It is shown that in computational efficiency, the 3GCS excels the BICGSTAB iteration method using preconditioners such as Jacobi, incomplete lower and upper (ILU), and 3-D block incomplete lower and upper (BILU3D) preconditioners. It is also shown that coarse-grid residual equations based on the 3-D 1N/A grid structure can predict temporal truncation errors as accurately as the 3-D 4N/A fine-grid residual equation but with considerably less overhead computing time for variable time-step size control calculations by a step doubling method. In addition, incorporation of preconditioners into the 3GCS is shown to enhance further efficiency of the nonpreconditioned 3GCS. From these results, it is concluded that the temporal adaptive 3GCS employing coarse-grid residual equations for temporal step-size control as well as the preconditioner like the BILU3D can provide a very efficient iterative solution scheme for transient nonlinear nodal calculations
[en] The purposes of the tests are to assure that (i) leakage through systems and components penetrating primary containment shall not exceed allowable leakage rate as specified in the technical specifications or associated bases; and (ii) periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment, and systems and components penetrating reactor containment. Integrate Leakage Rate Test (ILRT) is performed at the design pressure of the containment and leakage rates are determined by calculating dry air mass in the containment by measuring pressure, temperature and relative humidity with applying the ideal gas law. This study reviewed the effect of random and system errors must be considered in the data analysis. The purpose of the containment leakage rate test is to assure the leakage tightness of the reactor containment for accident conditions. Leakage rate test is performed at the design pressure of the containment and leakage flowrates are determined by measuring the dry air pressure in the containment by applying the ideal gas law. This study reviewed the effect of random and system errors introduced by a difference between the measured parameter and the actual value of the parameter, produced by either predictable or identifiable (system error), or unpredictable or unidentifiable (random error).
[en] In order to optimize of nuclear safety regulation in the rapidly changing nuclear safety environment, Korea government determined that the existing safety standards needed to be reviewed from Integrated Regulatory Review Service(IRRS) team of International Atomic Energy Agency(IAEA). For optimizations of nuclear safety regulation, the reviews were performed by IAEA IRRS team from July 10-22, 2011. In the results of 2011 IRRS mission, 12 suggestion and 10 recommendation were found. To confirm follow-up measures, IRRS follow-up mission would be also performed by IRRS team 18-24 months later after the mission was over. In order to prepare the IRRS follow-up mission, the establishment of MS of Korea Institute of Nuclear Safety(KINS) had been initiated by reflecting the 4 found supplement items in module 4 and IAEA GS-R-3 requirements. As a result of the initiation, MS of KINS was established. To introduce the MS of KINS and gather another suggestions for its enhancement, the MS was considered as a theme
[en] The capability of the Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes in the area of containment thermal hydraulics and atmospheric gas/steam distribution has been assessed through various benchmarks including the international standard problems. TOSQAN, MISTRA and THAI experiments which were carried out as parts of the ISP-47 exercise showed that thermal stratification was established after injection of hot helium, steam or air. The authors suggested ‘a common international activity to perform generic plant application exercise to study nodalisation effects, impact of steam and light gas injection etc. based on a generic (probably simplified) PWR containment.’ An integrated leakage rate test (ILRT) of the containment, performed to validate the assumptions taken for the calculations of the offsite consequence following a reactor accident, could provide useful information that may partly meet the purposes suggested as above. In this study, we examined the thermal stratification phenomena during an ILRT performed at a CANDU 6 plant in Korea, and then simulated them using the MELCOR 1.8.6 code and the plant input model developed by the Korea Institute of Nuclear Safety (KINS).