• 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.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.343-352
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• 2014

Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II. For this, the GRS sampling based tool XSUSA is employed together with the TRITON depletion sequences from the SCALE 6.1 code system. Uncertainties for multiplication factors and nuclide inventories are determined, as well as the main contributors to these result uncertainties by calculating importance indicators. The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy.

• 한국환경과학회지
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• 제23권6호
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• pp.1067-1074
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• 2014

Life Cycle Assessment(LCA) has been carried out to evaluate the environmental impacts of glass bottle recycle. The LCA consists of four stages such as Goal and Scope Definition, Life Cycle Inventory(LCI) Analysis, Life Cycle Impact Assessment(LCIA), and Interpretation. The LCI analysis showed that the major input materials were water, materials, sand, and crude oil, whereas the major output ones were wastewater, $CO_2$, and non-hazardous wastes. The LCIA was conducted for the six impact categories including 'Abiotic Resource Depletion', 'Acidification', 'Eutrophication', 'Global Warming', 'Ozone Depletion', and 'Photochemical Oxidant Creation'. As for Abiotic Resource Depletion, Acidification, and Photochemical Oxidant Creation, Bunker fuel oil C and LNG were major effects. As for Eutrophication, electricity and Bunker fuel oil C were major effects. As for Global Warming, electricity and LNG were major effects. As for Ozone Depletion, plate glasses were major effects. Among the six categories, the biggest impact potential was found to be Global Warming as 97% of total, but the rest could be negligible.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1195-1208
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• 2019

This paper introduces a new two-step procedure for PWR depletion analyses. This procedure adopts the albedo-corrected parameterized equivalence constants (APEC) method to correct the lattice-based raw cross sections (XSs) and discontinuity factors (DFs) by accounting for neutron leakage. The intrinsic limitations of the conventional two-step methods are discussed by analyzing a 2-dimensional SMR with the commercial DeCART2D/MASTER code system. For a full-scope development of the APEC correction, the MASTER nodal code was modified so that the group constants can be corrected in the middle of a microscopic core depletion. The basic APEC methodology is described and color-set problems are defined to determine the APEC functions for burnup-dependent XS and DF corrections. Then the new two-step method was applied to depletion analyses of the SMR without thermal feedback, and its validity was evaluated in terms of being able to predict accurately the reactor eigenvalue and nodal power profile. In addition, four variants of the original SMR core were also analyzed for a further evaluation of the APEC-assisted depletion. In this work, several combinations of the burnup-dependent and -independent XS and DF corrections were also considered. The results show that the APEC method could enhance the nodal equivalence significantly with inexpensive additional costs.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4722-4730
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• 2022

This paper introduces a new point depletion-based source term calculation code named BESNA (Bateman Equation Solver for Nuclear Applications), which is aimed to estimate nuclide inventories and source terms from spent nuclear fuels. The BESNA code employs a new modified CE/CM (Constant Extrapolation - Constant Midpoint) predictor-corrector scheme in depletion calculations for improving computational efficiency. In this modified CE/CM scheme, the decay components leading to the large norm of the depletion matrix are excluded in the corrector, and hence the corrector calculation involves only the reaction components, which can be efficiently solved with the Talyor Expansion Method (TEM). The numerical test shows that the new scheme substantially reduces computing time without loss of accuracy in comparison with the conventional scheme using CRAM (Chebyshev Rational Approximation Method), especially when the substep calculations are applied. The depletion calculation and source term estimation capability of BESNA are verified and validated through several problems, where results from BESNA are compared with those calculated by other codes as well as measured data. The analysis results show the computational efficiency of the new modified scheme and the reliability of BESNA in both isotopic predictions and source term estimations.

• Nuclear Engineering and Technology
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• 제44권2호
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• pp.207-224
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• 2012

This article discusses selects of some outstanding problems in nuclear reactor analysis, with proposed approaches thereto and numerical test results, as follows: i) multi-group approximation in the transport equation, ii) homogenization based on isolated single-assembly calculation, and iii) critical spectrum in Monte Carlo depletion.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• 대한전자공학회논문지SD
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• 제39권1호
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• pp.1-11
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• 2002

본 논문에서는 나노영역의 고해상도 도핑 농도 측정 장비 개발을 위해 공핍 근사 조건하 복잡한 계산 영역에서 공핍 영역을 간단히 계산할 수 있는 방법을 개발하였다. 개발된 공핍영역 계산 방법은 유한요소법을 이용한 적응분할 포아송 방정식 해석기를 사용하여 대전된 영역의 경계에서 전위가 0인 등고선과 일치하도록 하여 계산하는 방법이다. 이 방법의 타당성을 검증하기 위해 계산된 대전영역 및 전위분포가 공핍영역의 정의에 맞는지 확인하였으며, pn 접합에서의 공핍영역 깊이 및 MOS 구조에서 정전용량을 계산하여 비교해 본 결과 이론치와 정확히 일치함을 알 수 있었다. 이러한 Pn 접합 및 MOS 에서 공핍영역 계산 검증을 바탕으로 나노영역의 탐침을 장착한 SCM에서 전압에 따른 실리콘 내의 공핍영역 모양과 전위를 분석하여, 정전용랑 모델링을 하였으며, 이로부터 CV 곡선과 SCM의 출력인 dC/dV곡선을 계산하였다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

• 한국수자원학회논문집
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• 제45권12호
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• pp.1243-1258
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• 2012

본 연구에서는 지하수 양수에 따른 하천수 감소량을 예측할 수 있는 상관관계식을 지표수-지하수 통합모의 결과를 이용하여 유도하였다. 신둔천과 죽산천 두 개의 시험유역에 대해 지표수-지하수 통합모형 SWAT-MODFLOW을 적용하여 다양한 양수조건에 따른 하천수 감소량 자료를 모의 생성하였으며, 생성된 자료를 바탕으로 다중회귀분석을 실시하여 지하수 양수량, 하천과 양수정 이격거리, 대수층 및 하천바닥의 수리특성, 강수량 등의 함수인 하천수 감소량 산정 식을 유도, 제시하였다. 개발된 상관관계식은 하천인근 지하수 양수에 따른 하천수 영향을 평가하는데 간편하게 적용될 수 있으며, 지하수 양수량 중 하천수에 해당하는 부분인 하천수 기여도를 산정하는데 활용될 수 있을 것으로 기대된다.