• Nuclear Engineering and Technology
• /
• v.44 no.2
• /
• pp.177-198
• /
• 2012

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.22-22
• /
• 1999

The HELlOS-MASTER code system is verified through the benchmark of the critical experiments that were performed by RRC "Kurchatov Institute" with water-moderated hexagonally pitched lattices of highly enriched Uranium fuel rods (8Ow/o). We also used the same input by using the MCNP code that was described in the evaluation report, and compared our results with those of the evaluation report. HELlOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves neutronics model with the AFEN (Analytic Function Expansion Nodal) method for hexagonal geometry. The results show that the HELIOSMASTER code system is fast and accurate enough to be used as nuclear core analysis tool for hexagonal geometry.ometry.

• Proceedings of the Korea Society for Simulation Conference
• /
• 2000.04a
• /
• pp.168-173
• /
• 2000

지금까지 국내에서 설치되어 있는 원전 시뮬레이터용 노심 (Neutronics) 모델 프로그램은 주로 전산기 성능이 오늘날 비해 낮은 환경에서 실시간으로 노물리 계산을 위해 중성자 확산(Diffusion)현상을 미리 반영한 곡선을 사용하는 등 빠른 계산을 위해 많은 가정과 간략화가 있었다. 본 논문에서는 중성자 물리 계산을 2 Group 3-D로 계산이 가능한 최신의 노심코드(REMARK)를 이용하여, WH사가 공급한 900Mw의 3 Loop PWR인 영광 1호기 12주기를 기준으로 한 시뮬레이터의 노심모델 개발하기 위한 핵설계 전산체계인 APA(ALPHA-PHOENIX-ANC) 시스템의 출력으로부터 자동으로 REMAR 입력데이타를 생성하기 위한 GUI툴 개발과 개발된 노심모델의 자체 검증 및 원자력발전소 사고해석에 쓰이는 최적평가코드(RETRAN)를 기반으로 하는 최신 실시간 열수력 시뷸레이션(ARTS) 모델과 결합(Integration)되어 원자로 냉각재 펌프 1대 정지 및 터빈정지 시험등 과도시험한 결과를 기술하였으며 개발된 노심 모델은 원자력 교육원 2호기 시뮤레이터에 적용될 예정이다.

• Proceedings of the Korea Society for Simulation Conference
• /
• 2005.11a
• /
• pp.153-158
• /
• 2005

영광 원자력발전소 1호기가 16주기로 운전됨에 따라 훈련용 시뮬레이터의 입력자료 또한 16주기가 반영되어야 한다. 시뮬레이터의 여러 모델 중 노심모델(REMARK)에 필요한 입력자료는 Westinghouse의 핵 설계 코드체계인 APA 시스템의 Output에서 얻을 수 있으나 그 양이 방대하기 때문에 수작업을 통한 입력자료 생산은 큰 어려움을 갖는다. 따라서 이러한 작업을 수행할 프로그램 개발이 필수적이며 개발된 프로그램을 매 교체주기마다 적용하여 노심모델에 대한 원활한 입력상수 생산을 가능하게 할 수 있다.

• Proceedings of the Korea Society for Simulation Conference
• /
• 2005.11a
• /
• pp.165-170
• /
• 2005

원전 시뮬레이터의 노심모델(REMARK)은 원자로에서 발생하는 1차원 열원을 정확히 모사하고, 정상상태 및 과도상태의 반응도 변화에 따른 중성자속의 거동을 실제 원자로와 유사하게 모사할 수 있어야 한다. 모사의 정확성을 높이기 위해 Tuning 작업이 필수적이나 그 작업 단계가 매우 복잡하여 많은 시간과 노력이 필요하기 때문에 이를 간소화하면서 정확성을 높일 수 있는 Tuning 지원 프로그램을 개발하였다. 개발된 프로그램의 사용결과 신속하고 정확한 Tuning이 가능하였다.

• Nuclear Engineering and Technology
• /
• v.39 no.3
• /
• pp.187-192
• /
• 2007

The basic research on fast reactor technology was started in the mid-1960's in China. The emphasis was put on fast reactor neutronics, thermohydraulics, sodium technology, materials, fuels, safety, sodium devices and instrumentation. In 1987, the research turned to applied basic research with the conceptual design of a 60 MW experimental fast reactor as a target. The Project of the China Experimental Fast Reactor(CEFR) with a thermal power 65 MW was launched in 1993. The R&D of fast reactor technology then carried out to serve a design demonstration connected with the different phases of the conceptual, preliminary and detailed design of the CEFR. Recently, three directions of fast rector technology R&D activities have been considered, and some research programs have been developed. They are: (1) R&D related to the CEFR, i.e. experiments to be conducted on the CEFR for its safe operation, (2) R&D related to the projects of a prototype and the demonstration of fast reactors, and(3) advanced SFR technology within the framework of the international cooperation of INPRO and GIF.

• Nuclear Engineering and Technology
• /
• v.49 no.6
• /
• pp.1181-1188
• /
• 2017

This paper proposes the adoption of Monte Carlo perturbation theory to approximate the Jacobian matrix of coupled neutronics/thermal-hydraulics problems. The projected Jacobian is obtained from the eigenvalue decomposition of the fission matrix, and it is adopted to solve the coupled problem via the Newton method. This avoids numerical differentiations commonly adopted in Jacobian-free Newton-Krylov methods that tend to become expensive and inaccurate in the presence of Monte Carlo statistical errors in the residual. The proposed approach is presented and preliminarily demonstrated for a simple two-dimensional pressurized water reactor case study.

• Nuclear Engineering and Technology
• /
• v.49 no.6
• /
• pp.1326-1338
• /
• 2017

The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of $0.99994{\pm}6.8E-6$ (95% confidence interval). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.

• Nuclear Engineering and Technology
• /
• v.49 no.6
• /
• pp.1339-1345
• /
• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.21-27
• /
• 1996

KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.