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[en] The hypothetical accident analysis for the internal subchannel blockages and the external inlet blockage were carried out in an assembly of KALIMER-600 using MATRA-LMR code. The 6 subchannel blockages which are the design basis event for the internal blockage in the KALIMER-600 design were assumed that it would be occured at three different locations. For the sensitivity study of blockage size, the accidents of the 24 subchannels and 54 subchannels blockages were also analysed. The external inlet blockages were analysed from 0 % to 100 % reduced inlet flow rate with the interval of 10 % of the reduce rate. The peak temperature is 620 .deg. C and the core exit temperature is beyond the temperature limit of 600 .deg. C of KALIMER-600 for the 6 subchannel blockages. The temperature is within the temperature limit of the cladding design, however, is violated at the temperature limit of the infrequent events. Thus the detection measure of the accidents should be devised. The results of 24- and 54 subchannel blockages have been shown the more severe consequences than those of 6 subchannel blockages. As a results of the external inlet blockage, the clad integrity is secured upto the reduce rate of 40 % of inlet flow rate. The decrease of more 50 % of inlet flow rate threatens the integrity of the clad and the decrease of more 70 % of the flow cause of sodium boiling
[en] The flow blockage accident in an assembly of LMR brings the various flow fields such as Stoke's flow, laminar flow, and turbulent flow. Hybrid numerical scheme which both of donor cell method and central difference method used together has been newly applied into MATRA-LMR to analyze reasonably the accident. The code capability has been investigated using the experimental data observed in FFM(Fuel Failure Mock-up)-2A and 5B for the two typical flow conditions in a blocked channel. The predicted results by MATRA-LMR applied hybrid numerical scheme with the distributed resistance model have agreed well with the experimental data
[en] Summary: • KAERI is concentrating on the development of key technologies for the implementation of safety design concept: - Validation of Passive function of DHRS; - MA bearing metallic fuel. • The licensing of demonstration SFR will be pursued in the current regulatory framework on the basis of deterministic approach. The safety design will be supported by probabilistic approach. • A system analysis code for SFR system has been developed. This code will be validated further with available test data worldwide. • An evaluation methodology for DBEs is under development which aims to perform the safety analysis of demonstration SFR.
[en] As a preliminary work to validate the operability of a letdown stop valve of chemical and volume control system (CVCS) in a pressurized water reactor type nuclear power plant, the differential pressure through the valve is evaluated with RELAP5/MOD 3.2.2 γ PC-version. This kind of evaluation is important when the thrust force predicted with simple conservative method is too high to allow the operation of actuator. The excessive conservatism is removed and more realistic differential pressure during the valve closure is calculated in the present study. The calculated differential pressure is about 1231 psid at the moment of valve closure, which is much less than the value obtained in the conventional method. The results show that the operability of a motor-operated valve can be guaranteed for more complicated and dynamic situation when we use two-phase thermal-hydrualic code such as RELAP5/MOD3
[en] Summary: • Korea is preparing licensing infrastructure for future reactor systems, especially for the demonstration of SFR and VHTR technologies. • The safety requirements for SFR will be developed based on those for current LWR. • Technology-neutral safety criteria based on RI-PB approach would not be applied for the time being: - Technology-neutral safety criteria is under development stage; - Reliability data in PSA is not fully established due to the lack of operational experience of future reactor systems. • Integrated and systematic review method would be investigated: - Essential items for safety requirements development will be identified using OPT.
[en] Summary: • Long-term Plan for SFR Development was approved by the KAEC in December 2008: - Standard design approval by 2020; - Construction of demonstration SFR by 2028. • Activities for the development of Advanced SFR Concept: - Advanced concept design studies; - Development of advanced technologies; - Development of basic technologies. • Experimental studies for the validation of PDRC concept and performance test are undergoing. • Feasibility Study on Ultrasonic Waveguide Sensor Technology has been performed. Technology for under sodium viewing is being refined with design improvement of Ultrasonic Waveguide Sensor Module.
[en] The distributed resistance model has been newly implemented into MATRA-LMR to improve its prediction capability over the wire-wrap model for the flow blockage analysis in LMR. The code capability has been investigated using the experimental data observed in FFM(Fuel Failure Mock-up)-2A and 5B for the two typical flow conditions in a blocked channel. The predicted results by MATRA-LMR with the distributed resistance model have agreed well with the experimental data for the wire-wrapped subchannels. However, it has been suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for the reasonable prediction of the temperature field under a low flow condition
[en] There are some phenomenological uncertainties for partial or whole core melting of Sodium-cooled fast reactor with the metal fuel. Especially, the major phenomenology to be experimentally resolved is the potential for freezing and plugging of molten metallic fuel in above-core and below-core structures and possibly in inter-subassembly spaces. For suspect to these phenomenological uncertainties, we are required to understand characteristic behavior of molten fuel. Also, proof of the capacity of the debris bed cooling that is one of the technical issue for safety of Gen IV SFR is an essential condition to solve the problem of in-vessel retention of the core debris. For these sequences, experimental data for liquidus/solidus temperature and mobilization temperature of metal fuel are required essentially. In this study, review of the technical status on the liquidus/solidus temperature of metal fuel was carried out to investigate the behavior of liquidus/solidus/relocation of metal fuel
[en] The reactivity feedback effect of metallic fuel is determined by the fuel burnup characteristics, the configuration of core and fuel assembly, and the complicated interaction between the fuel assembly and core internal structures. Currently, a quite simple evaluation model is frequently applied for the calculation of reactivity feedback. The simple model usually induces some over-conservatism to compensate the simplification, which is an obstacle to take advantage of the positive characteristics of metallic fuel over the oxide fuel. Therefore, to develop a detailed reactivity feedback model and to remove the over-conservatism in the existing simple model would be the foundation to strengthen the economic and operational competitiveness of a liquid metal-cooled fast reactor. In the present study, the plan for the development of the detailed reactivity feedback model and the methodology to combine the spatial kinetics code with the thermal-hydraulic code have been set up, which are two prerequisites for the evaluation of the detailed reactivity feedback effect. The proposed detailed model is expected to be developed in short-term, thus, easily implemented in the SSC-K code. The development of the spatial kinetics code and the merging it to the detailed thermal-hydraulics code would be achieved in long-term, but finally minimize the uncertainty in the reactivity feedback evaluation by including the detailed thermal-hydraulic information in the reactivity calculation
[en] A flow blockage accident is one of the most important accidents threatening the safety of a liquid metal-cooled reactor (LMR) because the configuration of fuel rod in a subassembly is very compact. The analyses for the internal subchannel blockages were carried out for the subassembly design of KALIMER-600 using MATRA-LMR-FB code. The blockage of 6 subchannels constitutes the design basis in KALIMER-600. The design basis flow blocakges at various locations were analyzed with the conservative assumption for subassembly power, flow rate, and inlet temperature. The analyses for the beyond-design basis accidents, in which 24 or 54 subchannels were blocked, were performed at nominal design conditions. The conservative design basis flow blockage of the KALIMER-600 resulted in a enough margin of more than 370 .deg. C to the sodium boiling temperature which is the safety limit for coolant. However, the most conservative analysis resulted in the average subassembly exit temperature of 657 .deg. C, which violates the safety limit of 650 .deg. C for XE category accidents. Therefore, the integrity of the structure has to be evaluated for the local over-temperatue considering the allowance time of over-limit. Based on the evaluation result, it could be required to monitor the subassembly exit temperature due to a blockage and provide a measure by operator or automatic reactor trip. If a more large scale of blockage is classified into the accident of residual risk, a beyond design basis flow blockage is estimated to satisfy the safety goal because it is not developed to a severe core damage accident. However, it is required to equip a highly reliable core exit temperature detector or other kinds of detector which monotors the occurrence of a large scale blockage. The effective detection of blockage enables a timely reactor trip and quarantees the limited failure of fuel cladding