• Nuclear Engineering and Technology
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• 제10권4호
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• pp.203-213
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• 1978

전산 코-드인 LATREP, HWRAXAV 및 CITATION을 이용하여 CANDU-PHWR인 월성 1호기의 핵설계 특성 해석을 시도하였다. 계산된 주요 핵 특성은 CANDU 핵 연료봉 집합체에 대한 격자상수와 로심내의 출력 분포이며 그 계산 결과는 월성 1호기의 예비 안전성 보고시와 비교되었다. 계산치와 예비안전성 보고서에 제시된 설계치 사이의 차이점에 관해서는 예비안전성 보고서의 로심 기술에 대한 불완전한 자료와 계산 방법이 서로 다르다는 관점에서 검토되었다.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1993년도 추계학술발표회 초록집
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• pp.103-108
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• 1993

In nuclear power plants using pressurized water as the main coolant, it is necessary to maintain system pressure within operational range. During transients, the coolant shrinks and expands causing insurge and outsurge of coolant in the pressurizer. In CANDU system, the pressure is controlled mainly by the pressurizer/degasser-condenser system. In CANDU system, the control of heat transport system pressure is achieved by giving heat to the pressurizer by activating the heaters to compensate a diminution in pressure or by removing heat from the pressurizer by bleeding steam to the degasser-condenser to compensate an increase in pressure. This study aims at developing a theoretical model capable to simulate various operational transients in the CANDU primary heat transport system (PHTS), applicable to CANDU engineering simulator on real time basis.

• Nuclear Engineering and Technology
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• 제12권2호
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• pp.111-120
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• 1980

LATREP의 새로운 version인 PHWCELL을 사용하여 월성 CANDU원자로의 핵물리상수를 계산하였다. 이 코드는 주로 중수원자로에 대한 격자상수를 계산하며, 이 코드를 사용하여 중수원 자로의 격자계산의 model 방안을 개발하였다. 본 연구에서 고려된 원자로 운전조건은 Cold Zero Power (CZP)와 Hot Full Power (HFP)로서 독작용이 평형인 상태에서 고려한 것이다. 격자상수는 핵연료의 연소에 대한 것도 고려하였으며, 계산된 결과들은 월성 원자로의 예비안전보고서에 주어진 값과 이전의 연구결과와 비교하였다.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1995년도 춘계학술발표회 초록집
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• pp.101-112
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• 1995

In order to improve the recovery efficiency of heavy water vapour from the atmosphere inside a reactor building, and to recover and upgrade the heavy water which escape, special treatments, such as reducing the ingress of light water vapour, are studied in the design of the CANDU reactor systems. This is considered in controlled method of the humidity over drawing fresh air through a desiccant dehumidifier which dries the air by absorption. Comparing with the moisture loads between summer and winter operations, the moisture removal rates are calculated. Those are proportional to the difference between the controlled space and the surrounding environment Installation of a new dehumidifier will be able to reduce the moisture loads from the cooling systems, improving overall system efficiency and saving operating costs.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.651-656
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• 1997

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been peformed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of the total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle.

• Nuclear Engineering and Technology
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• 제9권1호
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• pp.15-31
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• 1977

중수형 원자력발전소의 가동중에 연료를 재장전하는 특성을 고려하여 새로운 핵연료 batch와 주기의 개념을 서정하고, 연속적인 에너지 계산방법으로 개발하여 핵주기비 계산관계식을 유도하였으며, 이러한 관계식들로서 중수형 원자로에 사용될 수 있는 전자계산기 코드 HWRCOST를 개발하였다. 이 코드로서 현재 우리나라에 건설중인 CANDU-PHWR의 전수명에 걸친 핵연료 주기비를 계산하였고 아울러 우라늄 원광비, 성형 가공비, 사용핵연료 보관처리비 및 발전소 가동율의 변화에 대한 핵연료 주기비의 감응도를 분석하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.291-295
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• 1996

In CANDU reactor, each pressure tubes contain twelve fuel bundles and provide the inlet and outlet for the primary coolant. If a leak develops in the pressure tube, it is detected by Annulus Gas System which contains circulating dry $CO_2$ gas. Since the leaks caused by the flaws are resulted in pressure tube break, establishment of flaw assessment method is very significant in view of the fracture mechanics. In this paper, various criteria for assessing the flaws are presented to prevent the tube rupture and ensure the integrity of reactor operating.

• 대한기계학회논문집A
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• 제26권11호
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• pp.2305-2311
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• 2002

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate fur the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• 방사성폐기물학회지
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• 제20권3호
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.