• Nuclear Engineering and Technology
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• 제17권4호
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• pp.256-266
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• 1985

냉각재 유량상실 사고가 가압경수형 원자로인 고리 1호기에 대하여 해석되었다. 냉각재 유량 상실 사고는 그 심각도에 따라 다음과 같이 3가지로 분류된다. 즉, 일부 유량 상실사고, 완전 유량 상실 사고, 그리고 펌프 축 고착 사고이다. 사고 해석은 계통 과도 현상 및 평균 노심분석, DNBR 계산, 그리고 고온점 분석의 3단계로 수행된다. 원자로 계통과도 현상 코드인 KTRAN이 본 사고를 빠른 시간에 모사할 수 있도록 개발되었다. DNBR계산을 위해서는 열수력학 코드인 SCAN및 COBRA IV-I가 채택되었으며, 고온점 분석을 위해서는 연료봉 과도 현상 코드인 LTRAN이 쓰였다. 이러한 전산코드 시스템은 과도 현상 해석에 빨리 응답하여야 한다. 왜냐하면 사고가 발생한 후 수 초안에 심각한 상태에 이르기 때문이다. 불행히도 KTRAN코드에 의하여 이러한 목적은 충족되지 않았다. 그러나 다른 계통 해석 코드에 비하여 잔은 계산 시간에도 불구하고 KTRAN에 의한 계산 결과는 FSAR의 결과와 전반적으로 잘 일치함으로써 KTRAN코드가 사고 해석에 유용함이 밝혀졌다.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2104-2125
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• 2021

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3909-3918
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• 2022

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.

• 에너지공학
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• 제23권4호
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• pp.256-262
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• 2014

국내 원자력발전소의 주기길이는 전력회사의 전력수급계획에 따라 결정된다. 주기길이는 노심에 장전할 신연료 다발수와 핵연료 농축도를 조정하여 결정할 수 있다. 전력회사에서는 특정 주기길이를 만족시키기 위한 방법으로 신연료 다발수를 정한 후 핵연료 농축도를 결정하는 방법을 적용하고 있다. 그러나 이 방법의 경우 같은 주기길이를 갖는 다른 신연료 다발수와 핵연료 농축도의 조합들 보다 핵연료 주기비 측면에서 가장 경제적인지 판단할 수가 없다. 따라서 본 분석에서는 상용 노심설계 코드인 CASMO/MASTER 코드를 사용하여 OPR1000(Optimized Power Reactor 1000) 발전소를 대상으로 신연료 다발수와 핵연료 농축도 조합에 대한 노심 연소계산을 수행하여 동일한 주기길이를 갖는 최적의 신연료 다발수와 핵연료 농축도 조합은 무엇인지 분석하였다. 천이노심계산에서 발생할 수 있는 불확실도를 최소화하기 위해 노심 특성인자들이 변하지 않는 평형노심(equilibrium cycle)까지 계산을 수행하여 이때의 계산결과를 핵연료 주기비 계산에 사용하였다. 또한 평준화 핵연료 주기비(levelized fuel cycle cost) 계산에 있어 중요한 인자인 할인율(discount rate)에 대해서 국내뿐만 아니라 다른 나라의 실정에도 적용 가능하도록 민감도 분석을 수행하였다. 평준화 핵연료 주기비(levelized fuel cycle cost) 평가 결과 할인율(discount rate)이 낮은 경우 신연료 다발수는 줄이고 대신 핵연료 농축도를 높이는 조합을 통해 특정 주기길이를 만족시키는 방법이 경제적인 것으로 나타났다. 반면 할인율(discount rate)이 높은 경우는 핵연료 농축도는 낮추고 신연료 다발수를 늘리는 조합을 통해 특정 주기길이를 만족시키는 방법이 경제적인 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1626-1637
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• 2020

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence -which data is obtained from a real PWR test- is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1964-1969
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• 2019

The safety and efficiency of nuclear reactors largely depend on the monitoring and control of nuclear radiation. Due to the unique nuclear-physical characteristics, Hf is one of the most promising materials for the manufacturing of the control rods and the emitters of neutron detectors. It is proposed to use the Compton neutron detector with the emitter made of Hf in the In-core Instrumentation System (ICIS) for monitoring the neutron field. The main advantages of such a detector in comparison the conventional β-emission sensors are the possibility of reaching of a higher cumulative radiation dose and the absence of signal delays. The response time of the detection is extremely important when a nuclear reactor is operating near its critical operational parameters. Taking Hf as an example, the general principles for calculating the chains of materials transformation under neutron irradiation are reported. The influence of ^179m1Hf on the Hf composition changing dynamics and the process of transmutants' (Ta, W) generation were determined. The effect of these processes on the absorbing properties of Hf, which inevitably predetermine the lifetime of the detector and its ability to generate a signal, is estimated.

• Nuclear Engineering and Technology
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• 제15권4호
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• pp.267-279
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• 1983

우라늄 및 플루토늄을 핵연료로 이용하고 경수를 감속재로 쓰는 원자로에 대한 해석적 체제를 수립하였다. 이 체제는 두개의 주요 전산코드로 구성되어 있는 바, 하나는 단위격자 세포코드인 KICC로서 이는 GAM및 THERMOS의 이론적인 기초에 여러가지 현상을 적절하게 포현할 수 있는 근사식을 결합한 것이다. 다른 하나는 다차원 확산-연소 방정식코드인 KIDD이다. 이 체제는 다양한 종류의 임계실험로에 대하여 철저한 검증계산을 수행하므로써 그 신뢰성을 입증하였다. 즉 서로 다른 노심구조를 가진 19가지의 비균질 임계실실험로에 대하여 유효증배계수를 계산한 결과 이의 평균치 1.0006, 표준편차 0.0039로서 잘 일치하였다. 또한 우라늄과 플루토늄을 핵 연료로 사용하는 여러종류의 임계로에 대하여 출력분포를 계산하여 측정치와 비교하였으며 계산치는 최대 오차 $\pm$5%이내에서 측정치에 일치하였다. 이러한 사실은 KICC/KIDD체제가 경수로의 해석에 아주 유용한 도구가 됨을 밝혀 준다.

• Nuclear Engineering and Technology
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• 제20권2호
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• pp.112-119
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• 1988

이 연구의 목적은 가혹한 사고후 손상된 원자로심을 모의한 열이 발생하는 성충 데브리층에서의 강제대류 드라이아웃 열유속 자료를 실험 적으로 얻고자 한 것이다. 여 기서는 일정한 층의 깊이와 냉각재 유입온도 조건하에서 선정된 몇 가지 크기의 입자로 성층을 형성한 데브리층에서 주로 냉각재질량유속이 드라이아웃 열유속에 미치는 영향을 관찰하였다. 이 실험적 연구로부터 얻은 주요 결과는 다음과 같다 (1) 성층 데브리층에서의 드라이아웃 열유속은 질량유속의 증가와 더불어 증가하며 그 증가의 경향은 크기가 균일한 입자층에 대한 것과 유사하다. (2) 이론치와 실험치와의 비교에서 입자직경으로는 표면적 평균 직경을 사용하는 것이 적합하다.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.