• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1847-1861
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• 2020

A design study was performed to improve the limit aseismic performance (LSP) of a primary heat transport system (PHTS) pump. This pump is part of the primary equipment of a prototype generation IV sodium-cooled fast reactor (PGSFR). The LSP is the maximum allowable seismic load that still ensures structural integrity. To calculate the LSP of the PHTS pump, a structural analysis model of the pump was developed and its dynamic characteristics were obtained by modal analysis. The floor response spectrum (FRS) initiated from a safety shutdown earthquake (SSE), 0.3 g, was applied to the support points of the PHTS pump, and then the seismic induced stresses were calculated. The structural integrity was evaluated according to the ASME code, and the LSP of the PHTS pump was calculated from the evaluation results. Based on the results of the modal analysis and LSP of the PHTS pump, design parameters affecting the LSP were selected. Then, ways to improve the LSP were proposed from sensitivity analysis of the selected design variables.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3928-3942
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• 2022

Thermal striping is a complex thermal-hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

• 한국전산유체공학회지
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• 제21권1호
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

• 한국전산유체공학회지
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• 제19권4호
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• pp.29-36
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• 2014

A computational study of a thermal striping in the upper plenum of PGSFR(Prototype Generation-IV Sodium-cooled Fast Reactor) being developed at the KAERI(Korea Atomic Energy Research Institute) is presented. The LES(Large Eddy Simulation) approach is employed for the simulation of thermal striping in the upper plenum of the PGSFR. The LES is performed using the WALE (Wall-Adapting Local Eddy-viscosity) model. More than 19.7 million unstructured elements are generated in upper plenum region of the PGSFR using the CFX-Mesh commercial code. The time-averaged velocity components and temperature field in the complicated upper plenum of the PGSFR are presented. The time history of temperature fluctuation at the eight locations of solid walls of UIS(Upper Internal Structure) and IHX(Intermediate Heat eXchanger) are additionally stored. It has been confirmed that the most vulnerable regions to thermal striping are the first plate of UIS. From the temporal variation of temperature at the solid walls, it was possible to find the locations where the thermal stress is large and need to assess whether the solid structures can endure the thermal stress during the reactor life time.

• 대한기계학회논문집A
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• 제37권5호
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• pp.665-671
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• 2013

제 4 세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX 와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX와 FHX는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다.