• Nuclear Engineering and Technology
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• 제33권1호
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• pp.12-24
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• 2001

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.328-333
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• 1996

Boric acid concentrations of the refueling water storage tank (RWST) and the accumulators for Ulchin 1&2 (UCN 1&2) are increased to meet the post loss of coolant accident (post-LOCA) shutdown requirement for the extended fuel cycles from 12 months to 18 months. To maintain long term cooling (LTC) capability following a LOCA, the switchover tine is examined using BORON code to prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results show that, at 8 hours after the initiation of LOCA. the emergency core noting system (ECCS) should be manually realigned to the simultaneous recirculation mode from the cold leg recirculation mode.

• Journal of Radiation Protection and Research
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• 제20권1호
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• pp.53-64
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• 1995

월성 원자력발전소 2,3,4 호기의 보조건물 주요 지역에서 냉각재 다량상h7사고 (large LOCA) 후의 방사선장을 평가하였다. 핵분열 생성물의 총량은 ORIGEN2 코드를 사용하여 계산하였고 선원항은 2중고장 시나리오, 즉 LOCA 사고후 비상노심냉각 (ECC) 계통의 고장이 결부된 사고시의 방사능 방출에 근거하였다. 원자로건물, 보조건물 및 ECC 계통의 구조모형을 QAD-CG 모델에 포함하여 계산하였다. 사고시점부터 90일 경과시까지 시간대 별로 선량율과 누적선량을 계산하였다. 결과적으로, 연속출입이 요구되는 중요지역에서의 방사선장은 충분히 낮은 것으로 평가되었다. 그러나, 일부구역에서는 제한적인 출입을 허용할 정도로 상대적으로 높은 방사선장을 나타내었다.

• Nuclear Engineering and Technology
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• 제30권1호
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• pp.47-57
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• 1998

Boric acid concentrations of the refueling water storage tank and the accumulators for Westinghouse 3-loop type plants are increased to meet the post loss of coolant accident shutdown requirement for the extended fuel cycles from 12 months to 18 months. To maintain long term cooling capability following a LOCA, the switchover time is examined using BORON code to prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results show that hot leg recirculation switchover times are shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3&4 and 8 hours from 18 hours for Ulchin 1&2, respectively. The How path in the mode J for Kori 3&4 is recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1&2.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.108-127
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• 2000

Hot leg break LBLOCA(Large Break LOCA) had a potential to be a containment maximum pressure accident in YGN3&4, which was induced from excessive conservatism in the CE analysis methodology of mass and energy release. This study conducted mass and energy release experiment for the hot leg break LBLOCA during post blowdown with an integral test facility, SNUF(Seoul National University Facility). This facility simulated YGN 3&4 with volume ratio of 1/1140 based on Ishii's three level scaling. Experiment showed that SI(Safety Injection) water refilled cold leg first and core later. SI water was vaporized in the core, which resulted in the repressurization of reactor. This increase of pressure drove the water in cold leg to flow up half height of U tubes. However, since the water was drained back soon, the release through the SG side broken section by evaporation was negligibly small. This study also provided experimental assessment of RELAP5 results by KAERI for the release through the SG side broken section.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.779-780
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.163.2-163
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• 1998

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.171-171
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.447-452
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• 1996

Realistic containment analyses for large LOCA using a merged torsion of RELAP5/CONTEMPT4 are conducted. Analyzed are Generic LOCA with respect to the mass and energy releases from the RCS and containment pressure and temperature behaviors. The break locations considered are the double-ended guillotine breaks at the RCP discharge and hot legs for UCN 3&4 plants. For discharge leg break. the predicted containment pressure and temperature reach a peak during blowdown phase, thereafter the pressure and temperature decrease gradually without the second reflood peak. For the hot leg break it is found that the bypass break flow through the broken steam generator-during post-blowdown is negligibly small so that the containment atmosphere is not pressurized after the end of blowdown.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2295-2303
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• 2021

To investigate wall-thinning impact on axial load resistance of Zircaloy-4 cladding rods after a LOCA transient, axial tensile samples have been machined on as-received tubes with reduced thicknesses between 370 and 580 ㎛. After high temperature oxidation under steam at 1200 ℃ with measured ECR ranging from 10 to 18% and water quenching, machined samples were axially loaded until fracture. These tests were modeled using a fracture mechanics approach developed in a previous study. Fracture stresses are rather well predicted. However, the slightly lower fracture stress observed for wall-thinned samples is not anticipated by this modeling approach. The results from this study confirm that characterizing the axial load resistance using semi-integral tests including the creep and burst phases was the best option to obtain accurate axial strengths describing accurately the influence of wall-thinning at burst region.