• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1881-1895
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• 2020

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

• Journal of Radiation Protection and Research
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• 제4권1호
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• pp.1-4
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• 1979

중성자(中性子) 수송이론(輸送理論)을 써서 기사용(旣使用) 핵연료(核燃料) 저장조(貯藏槽)에 있는 한 개(個)의 PWR용(用) 핵연료집합체(核燃料集合體)에 대(對)한 유효증배계수(有效增倍係數)($k_{eff}$)를 산출(算出)하였다. 이때 중성자(中性子) 수송방정식(輸送方程式)을 Sn-근이법(近以法)이라고 부르는 각분해법(角分害法)(Discrete ordinates method)으로 풀어서 유효증배계수(有效增倍係數)를 구했으며 이것이 핵임계(核臨界) 안전성(安全性) 결정(決定)이 된다. 본(本) 연구(硏究)에서는 각(角)과 에너지를 각각(各各) 4구간(區間)과 16군(郡)으로 분할(分割)하고 공간구간(空間區間)은 27구간(區間)으로 나누되 상이(相異)한 물질(物質)의 경계면근처(境界面近處)에서 세분(細分)하였다. 이와같은 방법(方法)으로 구한 유효증배계수(有效增倍係數)는 0.6145였는 데 이는 타연구자(他硏究者)가 계산(計算)한 반무한배열(半無限配列) 핵연료집합체(核燃料集合體)에 대한 유효증배계수(有效增倍係數)에 비(比)하여 상당히 낮은 값이었다.