[en] This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regard to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard Power Plant (KSNPP) reactor. The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective

[en] Recent accident in Fukushima Daiichi plant in Japan has turned down the ambitious stream of nuclear renaissance world-wide and nuclear industry is trying hard to enhance safety of the nuclear plants. Many countries are cutting down their nuclear projects planned or proclaiming to close all nuclear plants. But until we could find other sources of energy in future, nuclear power should still play a major role. Thus, demands for safety being extremely high, we might need a paradigm shift in safety concept. We need to propose higher safety goal in a more systematic way to public. Internationally IAEA proposes a framework from which to develop new safety rules and requirements. This TECDOC recommends that quantitative safety goals stated in probabilistic terms be implemented and proposes new 'Safety Approach (Fig.1)' for new NPPs like:(1) quantitative safety goals, (2) fundamental safety functions, and (3) defence in depth. IAEA-TECDOC- 1366 concludes that tying the levels of defence in depth (DID) concept to safety goals could assure that a NPP design is safe, sound and has a balanced DID. As part of our efforts to establish regulatory framework for safety of High Temperature Gas-cooled Reactor (HTGR), we have recapitulated in this paper the current safety goal and design practice in view of this new trends for safety reflecting lessens from Fukushima

[en] Dust can be generated in VHTR and its impact on VHTR source term has been a long lingering safety issue(1,2). Though the design of tri-structural isotropic(TRISO) fuel is known to prevent large scale failure and fission product release even in accident scenarios, radiologically significant quantities of fission products will be present in the primary system coming from either a small fraction of failed fuel particles or intact fuel because of the diffusion in the fuel(3). The released fission products will be transported in the coolant gas, and plate out onto surfaces in the primary system. But analysis of the fission product distribution, in both normal and accident scenarios, is complicated in VHTR by the likely presence of dust. The FP-laden dust is supposed to be mobilized during the rapid depressurization accident and thus contributes to another source term mechanism. Also in all industrial plants where burnable solid compounds are treated, dust combustion becomes a significant risk and combustion of graphite dust in Chernobyl adds concerns to the dust safety issue. Thus, dust generation, interaction with fission product, transport and combustion constitutes the areas to be resolved for dust safety issue. U.S. NRC is organizing dust working group meeting since 2009 to discuss this dust safety. This paper summarizes the past experiences and current knowledge of dust in section 2 referencing the discussions of the recent 2011 dust working group meeting and suggests a domestic regulatory perspective to resolve the issue in section 3

[en] Recent accident in Fukushima Daiichi plant in Japan makes big impacts on the future of nuclear business. Many countries are changing their nuclear projects and increased safety of nuclear plants is asked for from the public. Without providing safety the society accepts, it might be almost impossible to build new plants further. In this sense high temperature gas-cooled reactor (HTGR) which is under development needs to be licensed reflecting this new expectation regarding safety. It means we should have higher level of safety goal and a systematic regulatory framework to assure the safety. In our previous paper, we evaluated the current safety goal and design practice in view of this new safety expectation after Fukushima accident. It was argued that a top-down approach starting from safety goal is necessary to develop safety requirements or to assure safety. Thus we need to propose an ultimate safety goal public accepts and then establish a systematic regulatory framework. In this paper we are going to provide a conceptual regulatory framework to guarantee the safety of HTGR. Section 2 discusses the recent trend of IAEA safety requirements and then summarize the HTGR design approach. Incorporating these discussions, we propose a conceptual framework of regulation for safety of HTGR

[en] The Next Generation Nuclear Plant (NGNP) Project supports commercialization of the high temperature gas-cooled reactor (HTGR) technology. HTGR can be applied in many industrial applications as a substitute for burning fossil fuels, such as natural gas, in addition to producing electricity, which is the principal application of current light water reactors. Given that the HTGR configuration will be different than the current fleet of licensed reactors, there should be a clear understanding between the HTGR applicant and the regulatory body regarding the demarcation between those systems that are within the nuclear facility and under the regulatory jurisdiction and those that fall outside the scope of the regulation. To communicate the NGNP Project's position regarding this issue, NGNP has recently published a report. This paper evaluates the concept proposed in the report and proposes how to establish the design scope in case HTGR licensing is applied in Korea

[en] In Korea, a small-to-medium sized integral type reactor, called as 'System integrated Modular Advanced ReacTor(SMART)', has been developed since late 1990's by KAERI (Korea Atomic Energy Research Institute). The reactor can be utilized in such areas as seawater desalination and district heating. The designer is targeting to get a Standard Design Approval(SDA) by the end of 2010. SMART aims at achieving enhanced safety and improved economics, the enhancement of safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, component modularization, reduction of construction time, and high plant availability. The design characteristics contributing to the safety of SMART which is a small sized integral type PWR with a rated thermal power of 330 MWt are inherent safety features such as the integral configuration of the reactor coolant system and an improved natural circulation capability. By introducing a passive residual heat removal system and an advanced LOCA mitigation system, significant safety enhancement is achieved

[en] A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.

[en] Understanding of iodine behaviour in the containment is essential for the evaluation of severe accident consequences because iodine is a critical source term for early fatality. Qualified tools for the calculation of the iodine source term are also needed to perform meaningful risk analyses and to make decisions in the field of accident management, mitigation measures and emergency preparedness. The MELCOR1.8.5 code used for consequence analysis in KINS (Korea Institute of Nuclear Safety) has iodine pool chemistry models, and can simulate the formation of molecular iodine gas from containment sump in severe accident condition. But the capability of this model needs to be verified against experimental data and thus to validate the MELCOR pool chemistry model, we have evaluated the ISP-41 and the P11T1 test both of which are iodine formation experiments performed by RTF(Radioiodine Test Facility) of AECL

[en] The concept of passive safety is emphasized in VHTR (Very High Temperature Reactor) to accomplish the enhanced safety goal of future reactors compared to the reactors of previous generation. The Reactor Cavity Cooling System (RCCS) is the passive safety system in VHTR designed to remove the residual and decay heat. Large thermal margin, low power density and core configuration with large heat transfer surface in VHTR make passive cooling effective. In addition, RCCS in VHTR is the only safety-related system which should have a high reliability. Therefore, it is important to study the design characteristics of RCCS and derive safety issues to assure the passive safety of VHTR. In this paper, several safety issues related with RCCS in VHTR were identified and reviewed

[en] The combustible gas control in containment is one of the important items for plant safety, but there are still large uncertainties in analyzing the hydrogen behavior in the containment for the regulatory review of new plants. Moreover since the Shin Kori 1 and 2 and Shin Wolsong 1 and 2 plants named Optimized Power Reactor 1000 (OPR1000) are under licensing process now, so it is important for KINS (Korea Institute of Nuclear Safety) to have a reliable analysis methodology for the assessment of hydrogen risk. An integral severe accident code MELCOR which has been widely used was intensively verified against experiments, but we should always be careful about the capability of a code. As part of our efforts to establish a confident analysis environment we need to simulate experimental data using MELCOR code. This paper describes the first step of our efforts for the code verification. The HM2 test of OECD/THAI (Thermal hydraulics, Aerosol and Iodine) project was selected as a benchmark problem, which focused on the hydrogen mixing phenomenon in containment especially for the erosion process of stratified atmosphere by steam plume. THAI-HM experiment is part of the OECD/THAI Project run in Germany. Brief results of the first step approach are described