• Nuclear Engineering and Technology
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• 제45권4호
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• pp.459-468
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• 2013

During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3665-3676
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• 2023

The use of ion exchange resins to remove ionic impurities from solution is prevalent in industrial process systems, including in the primary heat transport system (PHTS) purification circuit of nuclear power plants. Despite its extensive use in the nuclear industry, our general understanding of ion exchange cannot fully explain the complex chemistry in ion exchange beds, particularly when operated at or near their saturation limit. This work investigates the behaviour of mixed-bed ion exchange resin, saturated with species representative of corrosion products in a CANDU (Canadian Deuterium Uranium) reactor PHTS, particularly with respect to iron chemistry in the resin bed and the removal of lithium ions from solution. Experiments were performed under deaerated conditions, analogous to normal PHTS operation. The results show interesting iron chemistry, suggesting the hydrolysis of cation resin bound ferrous species and the subsequent formation of either a solid hydrolysis product or the soluble, anionic Fe(OH)₃^-.

• 대한기계학회논문집A
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• 제33권2호
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.257-264
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• 2013

In Canada Deuterium Uranium (CANDU)-type nuclear power plants, the reactor is composed of 380 fuel channels and refueling is performed on one or two channels per day. At the time of refueling, the fluid force of the cooling water inside the channel is exploited. New fuel added upstream of the fuel channel is moved downstream by the fluid force of the cooling water, and the used fuel is pushed out. Through this process, refueling is completed. Among the 380 fuel channels, outer rows 1 and 2 (called the FARE channel) make the process of using only the internal fluid force impossible because of the low flow rate of the channel cooling water. Therefore, a Flow Assist Ram Extension (FARE) tool, a refueling aid, is used to refuel these channels in order to compensate for the insufficient fluid force. The FARE tool causes flow resistance, thus allowing the fuel to be moved down with the flow of cooling water. Although the existing FARE tool can perform refueling in Korean plants, the coolant flow rate is reduced to below 80% of the normal flow for some time during refueling. A Flow rate below 80% of the normal flow cause low flow rate alarm signal in the plant operation. A flow rate below 80% of the normal flow may cause difficulties in the plant operation because of the increase in the coolant temperature of the channel. A new and improved FARE tool is needed to address the limitations of the existing FARE tool. In this study, we identified the cause of the low flow phenomena of the existing FARE tool. A new and improved FARE tool has been designed and manufactured. The improved FARE tool has been tested many times using laboratory test apparatus and was redesigned until satisfactory results were obtained. In order to confirm the performance of the improved FARE tool in a real plant, the final design FARE tool was tested at Wolsong Nuclear Power Plant Unit 2. The test was carried out successfully and the low flow rate alarm signal was eliminated during refueling. Several additional improved FARE tools have been manufactured. These improved FARE tools are currently being used for Korean CANDU plant refueling.

• Nuclear Engineering and Technology
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• 제27권2호
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• pp.226-233
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• 1995

CANDU 원자로에서 심각하게 대두되는 압력관 파손을 방지하기 위해 압력관의 두께를 증가시키는 방안이 연구되었다. 본 연구에서는 압력관 두께변화가 Zr-2.5Nb 압력관의 응력, 수소농도 및 수소지연파괴에 미치는 영향에 대해 연구를 수행하였다. 압력관 두께가 현재의 4.2 mm에서 5.2 mm로 증가할 경우에 압력관이 받는 응력과 발전소 가동중에 누적되는 중수소 흡수량은 19% 줄어드는 것으로 나타났으며, 압력관에 균열이 발생할 경우 발전소 냉각동안에 일어나는 균열 성장은 상당히 감소한다. 수소지연파괴는 압력관이 받는 응력과 누적되는 수소량에 비해 지배되는데 이와 같은 결과로부터 두꺼운 압력관은 수소지연파괴 관점에서 상당한 이점이 있는 것으로 평가되었다. 그러나 압력관 두께 증가는 수소지연파괴의 성장속도를 가속할수도 있으므로 앞으로 연구할 사항이다.

• 방사성폐기물학회지
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• 제4권4호
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• pp.373-384
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• 2006

중수로 원전내 여러 계통으로 부터 발생된 폐수지내에는 $^{14}C$ 핵종이 다량 함유되어 있으며, Class A 및 C 폐기물로 분류되는 폐수지의 적정 처리 기술 개발을 위한 기초연구를 수행하였다. IRN-150 혼상 이온교환수지를 이용하여 비방사성 $HCO_3$ 이온과 양이온의 흡착 특성 및 탈차용액을 이용한 $HCO_3$ 이온의 제거 특성을 고찰하였다. IRN-150 수지의 $HCO_3$ 이온의 흡착능은 이론값에 근접한 11 mg-C/g-IRN-150을 나타내었고, $CS^+,\;CO_2^+,\;Na^+,\;NH_4^+$ 양이온의 흡착 친화도를 단일성분 및 복합성분 시스템을 이용하여 분석하였다. 여러 가지 탈착용액을 이용한 폐수지로부터 $HCO_3$ 이온의 제거 특성을 평가한 결과, $^{14}C$ 핵종을 전량 효과적으로 제거하기 위해서는 $NaNO_3,\;Na_3PO_3$ 보다도 $NH_4H_2PO_4$ 용액이 유리한 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제11권4호
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• pp.263-274
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• 1979

국내도입이 예상되는 900MWe급 가압경수로형 (PWR) 원자력 발전소와 캐나다형가압중수로형 (PHWR-CANDU) 원자력발전소에 대하여 throwaway 핵연료주기를 가상하여 두 노형의 상대적인 경제성을 비교 검토 하였다. 계산을 목적으로 발전단가를 발전소 투자비, 운전보수비, 운전자본비 및 핵연료비로 구분했으며 건설단가는 보완된 ORCOST 전산코드를 그리고 발전단가는 보완된 POWERCO-50 전산코드를 사용하여 구하였다. 계산에 요구되는 각종의 경제인자에 대하여는 단일의 수치값을 갖는 상수보다는 어떤 범위의 수치대를 이루는 통계적인 변수로 처리하였으며 ORCOST 및 POWERCO-50을 통한 무작위 추출법을 통하여 발전소 건설비 및 발전단가의 화율돗수 분포도를 얻었다. 계산결과 두노형간의 발전단가 분포도는 서로 겹치고 있으며 발전 단가의 기대치는 1986년도 미화로 PHWR의 발전단가가 PWR의 발전단가, 39.41mills/kwh보다 약 0.4mill/kwh만큼 적지만 PHWR의 건설기간이 PWR 보다 1년정도 더 걸리게되는 경우 차이가 없음을 알았다. 따라서 두 노형간의 경제성은 거의 우열을 가릴 수 없으며 한국에서 원자력발전소 노형을 선정할 때 기술전수, 국산화 등 경제외적 인자도 경제적 인자로 수량화하여 검토하는 것이 필요하다고 결론을 내렸다.

• 에너지공학
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• 제24권2호
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• pp.40-44
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• 2015

중수로 일차냉각재계통 액체방출밸브의 개방압력에 대한 안전여유 및 시간지연을 반영하여 열수력코드로 경년열화가 반영된 노심에 대해 민감도를 평가하였다. 과거에는 안전해석을 수행할 때 안전여유와 시간지연을 반영하여 평가하지 않았으나, 월성1호기 안전해석 인허가 심사과정중 반영 평가하였다. 중수로 안전해석에서 압력경계는 일차냉각재계통 액체방출밸브이다. 따라서 액체방출밸브 응동이 안전해석에 직접적인 영향을 주므로 안전여유와 시간지연 부가가 안전해석 결과에 미치는 영향을 파악하고 해석에 반영하기 위해 일차냉각재계통 과압이 걸리는 사고들에 대해 평가하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.