• Journal of Radiation Protection and Research
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• 제40권2호
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• pp.92-100
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• 2015

경수로 사용후핵연료 건식 중간저장시설의 방사선영향평가 효율성 개선을 목적으로 '선원항 지정방법에 따른 민감도 평가', '2-Step 계산'기법 개발 및 '냉각기간 이득효과' 적용에 따른 방사선 영향평가를 수행하였다. 본 연구에서는 저장건물의 용기배열 순서에 따라 순차적으로 선원항을 지정하여 직접선량에 미치는 민감도를 평가하였으며, 차폐건물 외벽에서의 방사선량은 내벽과 인접한 최근접 2개 열에 의한 영향이 지배적임을 확인하였다. 또한, 저장시설에 차폐 건물이 도입될 경우, 막대한 전산해석 시간을 감소시키기 위해 '2-Step 계산'기법을 수립하여 평가한 결과는 절반가량의 해석시간으로 직접(1-Step) 계산결과와 유사한 결과를 도출하였다. 마지막으로, 저장시설에 순차적으로 저장되는 저장용기의 보관기간을 사용후핵연료의 실제 냉각기간을 적용하면 건물 외벽에서의 방사선량이 냉각기간을 모두 동일하게 설정한 계산값에 비해 40% 정도 낮게 평가됨을 확인하였다. 본 연구는 중간저장시설의 방사선 영향평가를 위한 몬테칼로 차폐해석 방법의 효율성을 향상시키고자 수행되었으며, 좀 더 다양한 사례에 대한 평가를 통하여 신뢰성을 향상시킨다면 저장시설의 설계 및 부지경계 기준설정에 활용할 수 있을 것이다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 학술대회 논문집 정보 및 제어부문
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• pp.104-106
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• 2005

In this work, a PWR reactor core dynamics is identified online by a recursive least squares method. Based on this identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to design an automatic controller for thermal power control in PWRs. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired one, and the variation of the control rod positions. Also, the objectives are subject to maximum and minimum control rod positions and maximum control rod speed. Therefore, the genetic algorithm that is appropriate to accomplish multiple objectives is used to optimize the model predictive controller. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), is used to verify the proposed controller for a nuclear reactor. From results of numerical simulation to check the performance of the proposed controller at the 5%/min ramp increase or decrease of a desired load and its 10% step increase or decrease which are design requirements, it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• 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.

• 방사성폐기물학회지
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• 제14권4호
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• pp.435-443
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• 2016

본 논문은 국내 원전의 습식저장조에 저장 중인 경수로형 사용후핵연료를 금속겸용용기를 이용해 건식으로 운영하기 위한 운영공정을 개발하는 것이다. 국내 경수로형 원전의 사용후핵연료는 1990년대 초부터 습식으로 소내에서 운반을 한 경험은 많으나 건식으로 운전한 경험은 전혀 없는 실정이다. 이에 따라 금속겸용용기를 운영할 수 있는 세부 운영공정을 개발하였으며 주요 운영공정에서 금속겸용용기의 주요 구성품 및 사용후핵연료의 안전성이 유지됨을 확인하였다. 단기운영공정은 총 21시간 내에 이루어지도록 절차를 수립하였고 단계별로 허용운전 시간(15시간 습식공정, 3시간 배수공정, 그리고 3시간 진공공정)도 제시하였다.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.345-355
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• 2019

Use of the reconstructed fuel assemblies from partially burnt nuclear fuel pins is analyzed. This reutilization option is a potential candidate technique to make better use of the nuclear resources. Standard two step method is used to calculate node i.e. fuel assembly average burnup and then pin by pin ${\eta}$ values are reconstructed to ascertain the residual reactivity in the used fuel pins. Fuel pins with ${\eta}$ > 1:0 are used to reconstruct to-be-reused fuel assemblies. These reconstructed fuel assemblies are burnt during the cycle 3, 4, 5 and 6 of a 1000 MW PWR core by replacing fresh, once burnt and twice burnt fuel assemblies of the reference core configurations. It is concluded that using reconstructed fuel assemblies for the fresh fuel affect dearly on the cycle length (>50 EFPD) when more than 16 fresh fuel assemblies are replaced. However, this loss is less than 20 days if the number of fresh fuel assemblies is less than eight. For the case of replacing twice burned fuel, cycle length could be increased slightly (10 days or so) provided burnt fuel pins from other reactors were also available. Reactor safety parameters, like axial off set (< ${\pm}10%$), Doppler temperature coefficient (<0), moderator temperature coefficient at HFP (<0) are always satisfied. Though, 2D and 3D pin peaking factors are satisfied (<1:55) and (<2:52) respectively, for the cases using eight or less reconstructed fuel assemblies only.

• Nuclear Engineering and Technology
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• 제27권6호
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• pp.859-869
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• 1995

본 연구는 가압경수로의 부분충수 운전중 잔열제거기능 상실사고 해석시 신뢰성을 확보하기 위해 RELAP5/MOD3.1 코드로 관련 대형 실험을 모의 계산하여, 사고시 예상되는 주요 물리적 현상의 파악과 코드의 예측능력을 평가하는 것이다. 대상 실험으로 선택된 BETHSY Test 6.9a는 이 사고중 증기발생기가 작동하지 않고, 가압기 Manway를 개방한 상태 (Configuration)를 모의한 실험이다. 이 연구 결과는 실제 원전 사고시 예상되는 중요 현상 뿐 아니라, 이에 영향을 미치는 민감한 인자를 파악하여 사고 해석결과의 유효성을 판단하는 데 상당히 기여할 것으로 기대한다. 연구결과 RELAP5/MOD3.1 코드는 대체적으로 계통의 과도기 거동은 타당하게 예측하고 있지만, 모의계산에서 Time-Step이 아주 짧아 막대한 시간이 소요된다는 문제점이 발견되었다. 그 외에도 노심팽창수위 (swollen level)를 과대평가하여 가압기의 수위 및 계통의 압력을 높게 계산하였다. 이로 인해 가압기를 통한 방출량도 과대계산하여 노심노출을 약 400초 빨리 예측하였다. 실험과 코드 예측결과를 종합할 때 가압기 Manway 만의 개방으로는 계통압력이 상승하고, 중력주입냉각수로는 노심수위 회복에 불충분하며, 결국 강제주입에 의해서 노심수위가 회복될 수 있음이 입증되었다.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1209-1217
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• 2019

Due to the limitation of the computers, the conventional homogenization method is based on many assumptions and approximations, and some tough problems such as energy spectrum and boundary condition are faced. To deal with those problems, the Monte Carlo global homogenization is adopted. The Reactor Monte Carlo code RMC is used to study the global homogenization method, and the one-step global homogenization method is proposed. The superimposed mesh geometry is also used to divide the physical models, leading to better geometric flexibility. A set of multigroup homogenization cross sections is online generated for each mesh under the real neutron energy spectrum and boundary condition, the cross sections are adjusted by the superhomogenization method, and no leakage correction is required. During the process of superhomogenization, the author-developed reactor core program NLSP3 is used for global calculation, so the global flux distribution and equivalent homogenization cross sections could be solved simultaneously. Meanwhile, the calculated homogenization cross section could accurately reconstruct the non-homogenization flux distribution and could also be used for fine calculation. This one-step global homogenization method was tested by a PWR assembly and a small reactor model, and the results show the validity.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.336-343
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• 1995

적응제어의 한 방식인 자기동조제어(STR) 방식이 비선형 노심 모델의 출력 조정에 적용된다. 적응제어는 비선형, 시변 및 확률(Stochastic) 시스템을 위한 준최적 제어기를 설계하기 위한 적절한 제어 방식이다. 제어계통은 미지의 시변 파라메타를 갖는 3차 선형 모델에 기초한다. 파라메타는 가변 망각계수를 도입한 늑장 최소자승법에 의하여 매시간(Time Step) 순환적으로 평가된다. 평가된 파라메타를 이용하여 한 스텝 먼저 냉자재 평균온도가 예측되고 이 예측된 값과 Setpoint 값과의 차이를 최소화함은 물론, 제어봉의 움직임을 막고자 가중(Weighted) One-step-ahead 제어기가 설계된다. 또한 적분동작이 첨가되어 정상상태 에러가 제거된다. 넓은 운전영역을 포괄하는 비선형 PWR 모델이 원자로 출력 조정을 위한 본 제어기를 시뮬레이션하는데 이용되었다. 시뮬레이션 결과로부터 본 제어기의 성능이 우수한 것으로 판명되었다.

• Nuclear Engineering and Technology
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• 제35권5호
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• pp.399-411
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• 2003

A model predictive control method is applied to design an automatic controller for thermal power control in a reactor core. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the second optimal control input is not implemented and the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize the difference between the output and the desired output and the variation of the control rod position. The nonlinear PWR plant model (a nonlinear point kinetics equation with six delayed neutron groups and the lumped thermal-hydraulic balance equations) is used to verify the proposed controller of reactor power. And a controller design model used for designing the model predictive controller is obtained by applying a parameter estimation algorithm at an initial stage. From results of numerical simulation to check the controllability of the proposed controller at the $5\%/min$ ramp increase or decrease of a desired load and its $10\%$ step increase or decrease which are design requirements, the performances of this controller are proved to be excellent.

• Nuclear Engineering and Technology
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• 제32권6호
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• pp.566-594
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• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.