• Nuclear Engineering and Technology
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• 제18권1호
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• pp.38-47
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• 1986

경수로심의 열 여유도를 분석하기 위하여 반응표면 및 Monte Carlo 방법을 이용하는 통계적 분석 방법이 제시되었다. 통계적인 열 여유도 분석 방법은 입력변수들의 불확실도를 확률론적으로 처리함으로써 열 여유도의 최적 평가를 수행한다. 이 방법은 원자력 1호기 정상상태의 원자로심 분석에 응용되었으며 또한 종래의 결정론적 방법 및 웨스팅하우스의 개선된 열설계 방법과도 비교되었다. 본 연구를 통하여 반응표면 분석 방법은 통계적인 열 여유도 분석에 유용함을 알 수 있었으며, 이 방법을 통한 열 여유도의 증가도 확인되었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.379-384
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• 1996

Thermal-hydraulic performance has been analyzed for an advanced reactor core loaded with hexagonal fuel assemblies. Currently available CHF prediction models and data base for triangular lattice bundles have been thoroughly reviewed, and as a result the KfK-3 CHF correlation with limit CHFR of 1.235 has been determined to be most appropriate. The pressure drop model in COBRA-IV-I code has been modified for the analysis of triangular lattice rod bundles. In view of maximizing the thermal margin, the geometry of a hexagonal fuel assembly, such as rod diameter and rod pitch, has been optimized with a fixed fuel assembly cross sectional area The optimum value of the moderator-to-fuel volume ratio is estimated to lie between 0.65 to 1 with 9.5 mm rod diameter. The thermal margin of these hexagonal fuel assemblies in the AP600 core has been evaluated and compared with that of square lattice fuel assemblies such as VANTAGE-5H and KOFA. The analysis result shows that the performances of hexagonal fuel assemblies are more favorable than the square fuel assemblies in the aspect of steady-state overpower margin.

• Nuclear Engineering and Technology
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• 제16권3호
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• pp.155-160
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• 1984

표준 핵연료집합체나 최적 핵연료집합체로 구성된 원자력 1호기 원자로심의 열적여유도를 기존 열설계 방법과 통계적 열설계 방법을 이용하여 분석하였다. 통계적 열설계 방법은 노심내 운전변수들의 불확실도를 통계적으로 처리함으로써 기존 방법에 비하여 열적여유도를 증가시킨다. 계산을 위하여 정상상태와 과도시 열수력분석 전산코드인 COBRA-IV-i를 사용하였다. 계산결과 통계적 설계방법은 열적여유도를 크게 증가시키며, 표준 핵 연료집 합체는 물론 최적 핵 연료집 합체가 장전된 원자력 1호기의 열설계기준을 만족시키는 것으로 밝혀졌다. 그러나 기존 열설계 방법은 원자력 1호기 노심에 최적 핵연료집합체가 장전된 경우 열설계기준을 만족시키지 못하는 것으로 밝혀졌다.

• Nuclear Engineering and Technology
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• 제27권4호
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• pp.518-531
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• 1995

단일수로 해석 모형을 다수로 해석 모형으로 대체할 경우 얻을 수 있는 열적 여유도 향상에 대한 연구를 수행하였다. 이를 위하여 17$\times$17 국산핵 연료 장전 노심에 적용할 수 있는 새로운 임계열속 상관식을 개발하였으며, 여기에 사용된 부수로 국부 조건은 다수로 해석 코드인 TORC로 계산하였다. 그리고, 고온부구로 DNBR 분석을 위하여 전 노심에 대한 단일단계 해석 모형을 개발하였다. 분석 결과 다수로 해석 모형인 TORC/KRB-1 체제를 사용할 경우 단일수로 해석 모형인 PUMA/ERB-2 체제에 비하여 약 5% 이상의 열적 여유도를 회복할 수 있는 것으로 나타났다. 이러한 열적 여유도의 증가는 두 코드간의 고온부수로 국부조건 예측 성능 차이와 임계열속 상관식의 특성 차이에서 기인한 것이다.

• 전기학회논문지
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• 제58권12호
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• pp.2311-2315
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• 2009

This paper presents a method to apply energy margin for assesment of total transfer capability (TTC). In order to calculate energy margin, two values of the transient energy function have to be computed. The first value is transient energy that is the sum of kinetic and potential energy at the end of fault. The second is critical energy that is potential energy at controlling UEP(Unstable Equilibrium Point). It is seen that TTC level is determined by not only bus voltage magnitudes and line thermal limits but also transient stability. TTC assessment is compared by the repeated power flow(RPF) method and optimization method.

• International Journal of Aerospace System Engineering
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• 제9권2호
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• pp.1-9
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• 2022

The NEA release mechanism is used to provide restraint and release functions with low shock for critical deployment operations on solar arrays after launch. The GK3 solar array consists of 2 wings and 6 hold down points per panel. The NEA SSD9103S1 is a part of the GK3 solar array hold-down and release mechanism. Each NEA unit is equipped with two Z-diodes which provide power to a NEA unit connected in series after actuation of the fuse wire. This paper presents the hot firing test results of a quadrature NEA SSD9103S1 configuration. One output powers a maximum of 4 NEA SSD9103S1 units simultaneously. The necessary actuation pulse duration has been determined to meet margin requirement for thermal energy of minimum 4. Actuation thermal energy difference is about 6.6% between each half of two fired serial NEAs. Thermal energy margin at worst case is minimum 5.9 in case of an actuation pulse duration of 500 ms. Two series Zener impedance depend on current applied has been characterized by an additional actuation after all fuse wires are open circuit. Total number of actuation commands to the GK3 NEA unit reduce drastically from 24 in case of single NEA configuration down to 8 in case of parallel and quadrature NEA configurations. It can be accommodated by the existing HP2U Pyro design without any impact.

• Nuclear Engineering and Technology
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• 제22권1호
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• pp.45-57
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• 1990

아날로그와 디지탈 보호계통의 정상상태 여유도를 비교하였다. 비교된 예는 웨스팅하우스사의 OP Delta T 및 OT Delta T 계통과 CE사의 CPCS 계통이다. 안전해석 방법의 차이에 의한 여유도 영향을 제거하기 위해 Dynamic Offset은 고려하지 않았다. 핵연료봉 중심선의 용융을 방지하는데 있어서 디지탈 보호계통이 아날로그 보호계통보다 약 30% 출력 정도의 운전 여유도를 더 가졌다. DNB를 방지하는데 있어서는 주기말에서는 양 보호계통이 비슷한 여유를 가졌으나 주기초에서는 디지탈 보호계통이 약 10%의 더 많은 운전여유를 가지는 것으로 판단된다.

• Progress in Superconductivity
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• 제8권2호
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• pp.193-201
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• 2007

Since the margins for the minimum quench energy and for the operating current in the superconducting magnet determine the operating regime of the magnet, a thermal stability analysis for the KSTAR superconducting magnet system is performed using 1-D Gandalf code. The result shows that the minimum quench energy is about 500 mJ/cc and the operating current margin is about 70 %. These values are larger than those of the KSTAR design criteria and the KSTAR superconducting magnet system can be operated stably under various experimental environments.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.374-379
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• 1997

The thermal margin of CANDU-6 reactor is estimated by the CCP, which is dependent on fuel channel hydraulics and the CHF of fuel bundle. This paper intents to describe the characteristics of CCP behavior for the CANDU-6 channel in which CANFLEX-NU fuel bundles are assumed to be loaded. Also, it includes the thermal margin evaluation of the CANDU-6 channel loaded with a mixed CANFLEX-NU and 37-element fuel bundles as a simulation of the partial loading of CANFLEX-NU fuel bundle in the CANDU-6 reactor. For the mixed fuel channels, the effects of axial flux distribution(AFD) on CCP were investigated by using the AFD tilted in the downstream. The CCP of CANFLEX-NU fuel bundle was found to be improved by the CHF enhancement, despite of the slight flow decrease, in case of both full and partial loading, compared with those of a standard 37-element fuel bundle.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.684-701
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• 2016

The Gaussian process model (GPM) is a flexible surrogate model that can be used for nonparametric regression for multivariate problems. A unique feature of the GPM is that a prediction variance is automatically provided with the regression function. In this paper, we estimate the safety margin of a nuclear power plant by performing regression on the output of best-estimate simulations of a large-break loss-of-coolant accident with sampling of safety system configuration, sequence timing, technical specifications, and thermal hydraulic parameter uncertainties. The key aspect of our approach is that the GPM regression is only performed on the dominant input variables, the safety injection flow rate and the delay time for AC powered pumps to start representing sequence timing uncertainty, providing a predictive model for the peak clad temperature during a reflood phase. Other uncertainties are interpreted as contributors to the measurement noise of the code output and are implicitly treated in the GPM in the noise variance term, providing local uncertainty bounds for the peak clad temperature. We discuss the applicability of the foregoing method to reduce the use of conservative assumptions in best estimate plus uncertainty (BEPU) and Level 1 probabilistic safety assessment (PSA) success criteria definitions while dealing with a large number of uncertainties.