• Nuclear Engineering and Technology
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• 제45권7호
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• pp.895-910
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• 2013

The reflood test data from the rod bundle heat transfer (RBHT) test facility showed that the grids in the upper portion of the rod bundle could become wet well before the arrival of the quench front and that the sizes of liquid droplets downstream of a wet grid could not be predicted by the droplet breakup models for a dry grid. To investigate the water droplet generation from a wet grid spacer, a viscous linear temporal instability model of the water sheet issuing from the trailing edge of the grid with the surrounding steam up-flow is developed in this study. The Orr-Sommerfeld equations along with appropriate boundary conditions for the flow are solved using Chebyshev series expansions and the Tau-Galerkin projection method. The effects of several physical parameters on the water sheet oscillation are studied by determining the variation of the temporal growth rate with the wavenumber. It is found that a larger relative steam velocity to water velocity has a tendency to destabilize the water sheet with increased dynamic pressure. On the other hand, a larger ratio of steam boundary layer to the half water sheet thickness has a stabilizing effect on the water sheet oscillation. Droplet diameters downstream of the spacer grid predicted by the present model are found to compare reasonably well with the data obtained at the RBHT test facility as well as with other data recently reported in the literature.

• Nuclear Engineering and Technology
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• 제12권3호
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• pp.153-162
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• 1980

냉각재상실사고의 재관수 단계중 연료봉 피복재의 온도거동 및 열전달 기구를 파악하는 것은 비상노심냉각계통 및 원자로의 안전성해석에 중요하다. 냉각재유동채널의 방위가 rewetting과정에 미치는 영향을 연구하기 위하여 수직 및 수경 유동채널을 이용한 실험을 수행하였으며, 노심이 수평압력관으로 구성되어 있는 CANDU원자로에 관한 실험을 중점적으로 수행하여 그 결과를 수직채널의 결과와 비교 하였다. 또한 rewetting현상을 육안관찰가기 위해 환상형 테스트부 및 외부에서 가열되는 석영관을 사용하였다. 실험결과로써 수평채널에서의 rewetting 속도는 유동의 층상 현상에 크게 영향을 받으나 그 평균값은 수직채널리 경우와 큰차이없음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.690-703
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• 2010

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between $5^{th}$ and $95^{th}$ percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1532-1539
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• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.429-434
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• 1996

K-REM［1］, which is under development as a realistic evaluation model of large break LOCA, is applied to the analysis of cold leg guillotine break of Ulchin 3&4. Fuel parameters on which statistical analysis of their effects on the peak cladding temperature (PCT) are made and system parameters on which the concept of limiting value approach (LVA) are applied, are determined from the single parameter sensitivity study. 3 parameters of fuel gap conductance, fuel thermal conductivity and power peaking factor are selected as fuel related ones and 4 parameters of axial power shape, reactor power, decay heat and the gas pressure of safety injection tank (SIT) are selected as plant system related ones. Response surface of PCT is generated from the plant calculation results and on which Monte Carlo sampling is made to get plant application uncertainty which is statistically combined with code uncertainty to produce the 95th percentile PCT. From the break spectrum analysis, blowdown PCT of 1350.23 K and reflood PCT of 1195.56 K are obtained for break discharge coefficients of 0.8 and 0.5, respectively.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.355-366
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• 1994

미국원자력규제위원회에서는 최근 안전해석에 최적전산코드의 사용을 허용하는 개정된 비상노심냉각계통 평가 규정을 제시하였다. 당 규정에서는 계통해석에 최적전산코드를 사용할 경우 불확실성 평가를 수행할 것을 요구하고 있다. 본 논문에서는 이러한 비상노심냉각계통의 규제요건을 만족하는 실제적인 최적평가방법론을 개발하여 대형냉각재상실사고에 적용하였다. 최적평가전산코드로는 RELAP5/MOD3.1을 개선한 RELAP5/MOD3/KAERI를 사용하였으며, 코드의 불확실성은 수개의 분리효과 및 총체효과 실험에 대한 평가를 수행함으로써 정량화 하였다. 적용대상 발전소로는 고리 3 & 4호기를 선정하였다. 민감도 분석을 통하여 응답방정식을 구성하였으며 각 응답방정식에 대하여 무작위 추출방식, Monte Carlo 방식으로 확률밀도함수를 구하였다. 최종 불확실성은 95%의 신뢰도로 정량화 하였으며 대형냉각재 상실사고시의 안전여유도에 대하여 논의하였다.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.727-734
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• 2012

A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.

• 에너지공학
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• 제14권1호
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• pp.54-61
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• 2005

한국형 신형원자로1400(APR1400)은 3983MWt급의 2×4 루프 개량형 가압경수로(PWR)로서 대형 냉각재상실사고 발생시 안전주입수의 원자로용기 직접주입(DVI) 방식을 채택하고 있으며, 안전주입수탱크(SIT) 내부에 유량조절기(Fluidic Device, FD)를 장착하고 있다. 본 연구에서는 신형원자로 1400의 안전주입계통에 새로이 적용된 주요 특징 중 하나인 유량조절기에 대하여 최적안전해석코드인 TRAC-M/F90, 3.782버전을 이용한 성능평가 및 민감도 분석을 수행하였다. 연구결과 유량조절기가 안전주입수의 원자로 유입을 적절하게 조절하고 있음을 확인하였으며, 안전주입수탱크 내부의 압축질소체적 감소가 안전 주입수체적 감소에 비하여 노심의 급냉 완료 시점을 빠르게 하였다. 또한 안전주입계통의 전체 저항계수(K factor)가 최소 또는 최대일 때 노심의 급냉 완료 시점은 평균값인 경우보다 다소 늦어졌으나, 피복재 최고온도(PCT)는 상대적으로 큰 차이가 발생하지 않았다.

• Nuclear Engineering and Technology
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• 제23권1호
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• pp.56-65
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• 1991

RELAP5/MOD2 Cycle 36.04코드를 이용하여 LOFT대형냉각재 상실사고 모사실험 L2-3를 계산함으로써 코드의 대형냉각재상실사고에 관련된 열수력현상 예측능력을 평가하였다. 기본계산에서 원자로 압력용기는 이중노심유로와 분리강수관 모델로 모사되었다. 기본계산의 결과 계통의 전반적인 수력학적 거동과 감압기간동안 노심 고출력 부위에서의 열적 거동은 비교적 타당하게 예측되었다. 한편 과냉각-이상유동의 천이 기간동안 임계유량모델, 고질량유속에서의 임계열유속 상관식, 감압기간중의 재접수(Blowdown Rewet)의 판정기준등 코드의 모델/상관식의 부분적 결함이 발견되었다. 이 결함들에 의해 냉각재 재고량이 과대 평가되어 재환수기간의 노심의 열적거동 예측의 정확도가 감소되었다. RELAP5/MOD2 Cycle 36.04로 부터 개선된 코드를 사용한 계산 결과 재접수 현상의 예측 정확도를 개선할 수 있었다.

• Nuclear Engineering and Technology
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• 제21권4호
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• pp.259-266
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• 1989

LOFT L2-5 대형 냉각재상실사고 모사실험에 대해 RELAP5/MOD2/Cycle 36.04로부터 개선된 코드를 이용, 코드평가를 수행하였다. 강수관(Downcomer)모델 및 노심유로모델에 따른 코드 민감도분석을 위해 보충계산을 수행하였다. 계산결과는 1차계통의 압력, 파단부를 통과하는 질량유량, 노심 고온부의 피복재 온도등에 대해서 실험결과와 비교 분석되었다. 분석결과 RELAP5/MOD2 계산에 의해 1차계통의 수력학적 거동은 잘 묘사될 수 있으며, 단일 노심유로 모델을 이용한 계산에서는 사고발생이후 감압기간동안 노심이 과대 냉각되는 현상이 발견되었다. 노심의 고온유로에서의 수력학적 거동을 잘 묘사할 수 있는 이중 노심유로 모델계산을 이용하여 이 현상을 극복하고 실험치에 근사하는 결과를 얻을 수 있음을 알 수 있었다.