• Nuclear Engineering and Technology
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• 제48권2호
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• pp.318-329
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• 2016

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed "Burnable absorber-Integrated Guide Thimble" (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.152-161
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• 2022

We analyzed mitigation strategies for steam generator tube rupture (SGTR) accidents using MARS code under both full-power and low-power and shutdown (LPSD) conditions. In general, there are two approaches to mitigating SGTR accidents: supplementing the reactor coolant inventory using safety injection systems and depressurizing the reactor coolant system (RCS) by cooling it down using the intact steam generator. These mitigation strategies were compared from the viewpoint of break flow from the ruptured steam generator tube, the core integrity, and the possibility of the main steam safety valves opening, which is associated with the potential release of radiation. The "cooldown strategy" is recommended for break flow control, whereas the "RCS make-up strategy" is better for RCS inventory control. Under full power, neither mitigation strategy made a significant difference except for on the break flow while, in LPSD modes, the RCS cooldown strategy resulted in lower break and discharge flows, and thus less radiation release. As a result, using the cooldown strategy for an SGTR under LPSD conditions is recommended. These results can be used as a fundamental guide for mitigation strategies for SGTR accidents according to the operational mode.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1892-1900
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• 2023

A protective oxide layer forms on the material surfaces of a Nuclear Power Plant during operation due to high temperature. These oxides can host radionuclides, the activated corrosion products of fission products, resulting in decommissioning workers' exposure. These deposited oxides are iron oxides such as Fe₃O₄, Fe₂O₃ and mixed ferrites such as nickel ferrites, chromium ferrites, and cobalt ferrites. Developing a new chemical decontamination technology for domestic CANDU-type reactors is challenging due to variations in oxide compositions from different structural materials in a Pressurized Water Reactor (PWR) system. The Korea Atomic Energy Research Institute (KAERI) has already developed a chemical decontamination process for PWRs called 'HyBRID' (Hydrazine-Based Reductive metal Ion Decontamination) that does not use organic acids or organic chelating agents at all. As the first step to developing a new chemical decontamination technology for the Pressurized Heavy Water Reactor (PHWR) system, we investigated magnetite dissolution behaviors in various HyBRID inorganic acidic solutions to assess their applicability to the PHWR reactor system, which forms a thicker oxide film.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• 대한금속재료학회지
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• 제50권8호
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• pp.583-589
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• 2012

This work is concerned with a statistical approach to the life prediction on irradiation-assisted stress corrosion cracking (IASCC) of stainless steel (SS) for core internals of a pressurized water reactor (PWR). The previous results of the time-to-failure of IASCC measured on neutron-irradiated stainless steel components were statistically analyzed in terms of stress and irradiation. The accelerating life testing model of IASCC of cold worked Type 316 SS was established based on an inverse power model with two stress-variables, the applied stress and irradiation dose. Considering the variation of the yield strength and applied stress with the irradiation dose in the model, the remaining life of the baffle former bolt was statistically predicted during operation under complex environments of stress and irradiation.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• 방사성폐기물학회지
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• 제16권2호
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• pp.195-202
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• 2018

원칙적으로, 지층 처분은 고준위 방사성 폐기물의 최종 처분을 위한 안전한 방법으로 간주된다. 그러나 사용후핵연료에 함유된 $^{99}Tc$ 및 $^{129}I$와 같은 일부 장수명 핵분열 생성물은 지하 환경에서 흡수성이 적은 음이온 핵종으로 이동성이 매우 크며 수백 keV 범위의 베타선 방출로 생태계에 피폭선량을 야기시킬 수 있다. 따라서 이 두 핵종을 효율적으로 분리하여 방사능으로 유해하지 않은 핵종으로 전환할 수 있다면 처분 안정성에 긍정적인 영향을 줄 수 있다. 이를 위한 하나의 방법은 이 두 가지 핵종을 원자로에서 수명이 짧은 핵종 또는 안정적인 핵종으로 변환하는 것이다. 이를 위해 두 핵종을 태우는 데 어느 원자로 유형이 더 효율적인지 평가하는 것이 필요하다. 본 연구에서는 경수로(PWR), 중수로(CANDU) 및 고속로(SFR, MET-1000)의 $^{99}Tc$ 및 $^{129}I$의 핵 변환 시뮬레이션 결과를 비교하고 고찰하였다.

• 한국전산구조공학회논문집
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• 제23권3호
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• pp.289-301
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• 2010

가압경수로(PWR)에서 배출되는 고준위폐기물을 지하 500m의 화강암 암반의 처분장에 장기간(약 10,000년 동안) 처분하기 위하여 여러 구조적 안전성 평가 수행을 통하여 처분용기모델이 개발되었다. 기존에 설계 개발된 가압경수로용 처분용기 모델은 구조적으로 처분용기 내부에 정사각형 단면의 네 개의 고준위폐기물 다발이 처분용기 단면의 중심에 대칭되게 나란히 배열된 형태를 취하고 있다. 그러나 이와 같은 배열 형태가 최선의 구조인지는 아직 결정할 수 없다. 왜냐하면 나란한 배열구조의 처분용기는 정사각형 다발단면의 외곽모서리와 외곽 쉘과의 거리가 가장 짧아 경량화를 위한 단면 직경 축소에 한계가 있기 때문이다. 따라서 처분용기 단면 중심에 대하여 대칭형이면서 나란하게 배열된 네 개의 고준위폐기물 다발 각각을 각 다발의 중심에 대하여 일정 각도 회전하여 처분용기 단면 중심 면에 대하여 대칭성을 유지하면서 고준위폐기물 다발이 배열된 처분용기구조에 대한 구조안전성 평가가 매우 필요하다. 비록 지금까지의 연구에 이러한 회전된 다발의 배열단면을 갖는 처분용기는 발견되지 않지만 처분용기모델들의 구조적 안전성 비교 연구를 위해서 고준위폐기물 다발이 회전된 배열단면 변화에 따른 처분용기에 대한 구조해석이 요구된다. 따라서 본 연구에서는 네 개의 고준위폐기물 다발이 각각 다발의 중심에 대하여 일정각도 회전하여 처분용기 중심 면에 대하여 대칭적으로 배열된 단면의 가압경수로용 처분용기에 대하여 구조해석을 수행하였다. 구조해석을 수행한 결과 기존의 설계 개발된 처분용기 단면의 중심에 대칭되게 나란히 고준위폐기물 다발이 배열된 단면의 처분용기보다 다발의 중심에 대하여 일정각도(30~35도) 회전하여 처분용기 중심 면에 대하여 고준위폐기물 다발이 대칭적으로 배열된 단면의 처분용기가 구조적으로 좀 더 안정성이 있음이 밝혀졌다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.745-750
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• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

• 대한인간공학회:학술대회논문집
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• 대한인간공학회 1993년도 추계학술대회논문집
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• pp.21-30
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• 1993

This paper describes the development of a simulation model of nuclear power plant operators including cognitive aspects by using a network modeling soft ware, Micro-SAINT (System Analysis of Integrated Networks of Tasks) for the analysis of operator performance. Network model description based on Micro-SAINT includes tasks, resources, precedence relations among tasks, flow of information and PSFs (Performance Shaping Factors) on task performance. We have tried to evaluate the performance with several performance measures such as the number of tasks allocated, relative time presure among operators within a shift, for the selected test accident scenarior; small-break LOCA (Loss of Coolant Accident) in a PWR (Pressurized Water Reactor) type nuclear power plant.