• 에너지공학
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• 제20권4호
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• pp.259-266
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• 2011

원자로 정지냉각계통은 원자로 정지 시 핵연료 잔열 제거를 위하여 냉각수가 충분히 공급하고 원자로기기들을 보호할 수 있는 냉각율을 유지할 수 있도록 설계되어야 한다. 경수로 정지냉각계통을 분석하기 위한 KDESCENT코드를 중수로 정지냉각계통에 적용하여 보았으며 기존의 중수로형 해석코드인 SOPHT, SDCS 코드 결과와 비교분석하였다. 정지냉각펌프 모드와 열수송펌프 모드에서 정상냉각 운전상태는 계통의 설계 요건을 만족시켰으며 정지냉각 열교환기를 열제거원으로 사용하였을 때 냉각률은 설계요건에서 규정하고 있는 제한치인 $2.8^{\circ}C/min$ 이하의 값을 얻었다. 전반적인 냉각능력 분석 결과 월성 2, 3, 4호기 정지냉각계통은 핵연료로부터 핵분열 생성물의 방출을 충분히 제한하고 핵연료채널의 건전성을 유지시키기 위한 충분한 냉각을 핵연료에 제공하였다.

• Nuclear Engineering and Technology
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• 제35권3호
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• pp.251-259
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• 2003

The Taguchi method was applied to investigate the effect of design factors on the performance of the shutdown cooling heat exchanger in the SMART-P. This method provided the simulation matrix for the KDESCENT program and an efficient tool for analyzing the simulation results. Levels of the design factors were selected by the effectiveness-NTU method. From 18 runs with the KDESCENT program, it was found that the performance of the system was greatly influenced by the inlet temperature at the shell side and the mass flow rate of the reactor coolant at the tube side. After applying the Taguchi method, we identified the important design factor that should be controlled and designed carefully. This method provides an efficient way to estimate the influence of each design factor on a system performance.

• 한국유체기계학회 논문집
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• 제13권5호
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• pp.5-10
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• 2010

5MW, open pool type research reactor, is commonly used to education and experimental purpose. It is necessary to prepare a standardization of system designs for considering a demand. HANARO has prepared the standardization of 5MW research reactor system designs based on the design, installation, commissioning and operating experiences of HANARO. For maintaining an open pool type reactor safety, a primary cooling system (after below, PCS) should remove the heat generated by the reactor under a reactor normal operation condition and a reactor shutdown condition. For removing the heat generated by the reactor, the PCS should maintain a required coolant flow rate. For a verification of the required flow rate, a flow network analysis of the PCS was carried under a normal operating condition. Based on the flow network analysis result, this paper describes the PCS flow characteristics of a 5MW open pool type research reactor. Through the result, it was confirmed that the PCS met design requirements including design flow rate without cavitation.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2015년도 제49회 하계 정기학술대회 초록집
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• pp.104-104
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• 2015

Due to the demand of the cold neutron flux in the neutron science and beam utilization technology, the cold neutron source (CNS) has been constructed and operating in the nuclear research reactor all over the world. The majority of the heat load removal scheme in the CNS is two-phase thermosiphon using the liquid hydrogen as a moderator. The CNS moderates thermal neutrons through a cryogenic moderator, liquid hydrogen, into cold neutrons with the generation of the nuclear heat load. The liquid hydrogen in a moderator cell is evaporated for the removal of the generated heat load from the neutron moderation and flows upward into a heat exchanger, where the hydrogen gas is liquefied by the cryogenic helium gas supplied from a helium refrigeration system. The liquefied hydrogen flows down to the moderator cell. To keep the required liquid hydrogen stable in the moderator cell, the CNS consists of an in-pool assembly (IPA) connected with the hydrogen system to handle the required hydrogen gas, the vacuum system to create the thermal insulation, and the helium refrigeration system to provide the cooling capacity. If one of systems is running out of order, the operating research reactor shall be tripped because the integrity of the CNS-IPA is not secured under the full power operation of the reactor. To prevent unscheduled reactor shutdown during a long time because the research reactor has been operating with the multi-purposes, the introduction of the standby cooling system (STS) can be a solution. In this presentation, the design considerations are considered how to design the STS satisfied with the following objectives: (a) to keep the moderator cell less than 350 K during the full power operation of the reactor under loss of the vacuum, loss of the cooling power, loss of common electrical power, or loss of instrument air cases; (b) to circulate smoothly helium gas in the STS circulation loop; (c) to re-start-up the reactor within 1 hour after its trip to avoid the Xenon build-up because more than certain concentration of Xenon makes that the reactor cannot start-up again; (d) to minimize the possibility of the hydrogen-oxygen reaction in the hydrogen boundary.

• 한국전산유체공학회지
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• 제17권4호
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• pp.1-8
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• 2012

During a reactor normal operation, two parallel 50% capacity cooling pumps circulate primary coolant to remove the fission reaction heat of the reactor through heat exchangers cold by a cooling tower. When one pump is failure, the other pump shall continuously circulate the coolant to remove the residual heat generated by the fuels loaded in the reactor after reactor shutdown. It is necessary to estimate how much flow rate will be supplied to remove the residual heat. We carried out a flow network analysis for the parallel primary pumps based on the piping network of the primary cooling system in HANARO. As result, it is estimated that the flow rate of one pump increased about 1.33 times the rated flow of one pump and was maintained within the limit of the cavitation critical flow.

• 에너지공학
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• 제8권4호
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• pp.525-532
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• 1999

한국형 차세대 원자로는 ABB-CE사의 System 80+의 설계개념을 근간으로 하여 표준화된 원자로의 계통설계를 추진하고 있다. 본 연구에서는 차세대 원자로 정지냉각계통의 운전시 요구되는 인허가 요건등제반 조건을 충족시킬 수 있는지를 해석하였다. 또한 운전시 필요한 열교환기의 유효면적과 원자로 기기냉각수 유량등 기본적인 설계자료를 산출하여 차후 차세대 원자로 정지냉각계통의 상세설계 업부를 수행하는데 필요한 기초자료를 제시하여 핵증기공급계통 (NSSS)의 기술개발을 이루는데 목적이있다. 차세대 원자로는 울진 3, 4호기 열출력 2.825MWth 에 비해 열출력이 4,000MWth 로 증가되어 정지냉각계통의 관련서례자료를 새로 산출해야하므로 정지냉각계통의 냉각능력을 평가하는 KDESCENT 전산코드를 이용하여 원자로 노심의 잔열과 정지냉각계통의 현열을 제거할 수 있는 최소 유량을 제시하였으며 주요 구성기기인 열교환기, 펌프, 밸브 및 기타 기기의 기능 및 정지냉각계통의 운전절차를 고찰하였다.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.68-79
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• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4335-4349
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• 2023

Nuclear marine propulsion has been emerging as a next generation carbon-free power source, for which proper passive residual heat removal systems (PRHRSs) are needed for long-term safety. In particular, the characteristics of unlimited operation time and compact design are crucial in maritime applications due to the difficulties of safety aids and limited space. Accordingly, a compact and long-term-operable PRHRS has been proposed with the key design concept of using both air cooling and seawater cooling in tandem. To confirm its feasibility, this study conducted system design and a transient analysis in an accident scenario. Design results indicate that seawater cooling can considerably reduce the overall system size, and thus the compact and long-term-operable PRHRS can be realized. Regarding the transient analysis, the Multi-dimensional Analysis of Reactor Safety (MARS-KS) code was used to analyze the system behavior under a station blackout condition. Results show that the proposed design can satisfy the design requirements with a sufficient margin: the coolant temperature reached the safe shutdown condition within 36 h, and the maximum cooling rate did not exceed 40 ℃/h. Lastly, it was assessed that both air cooling and seawater cooling are necessary for achieving long-term operation and compact design.

• 대한기계학회논문집B
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• 제22권8호
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• pp.1083-1090
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• 1998

A kind of three-way check valve, so called hydraulic calve was proposed for the substitute of the density lock of passive reactor such as SPWR (System-Integrated Pressurized Water Reactor). The function of the valve are the separation of the borated water from main coolant loop for normal operation and the insurge of the water into the loop for shutdown and the removal of the decay power when the main coolant flow rate is not enough. To verify the operability and the characteristics of the valve, experimental works were executed with 1/3 scale model calve. Also a diffuser model was proposed for the theoretical analysis of the valve.