• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.106-111
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• 1998

ANL의 액체금속로 노심 해석 코드 DIF3D(1)를 OECD/NEA를 통하여 도입하여, DIF3D의 경계면 중성자류 노달 방법의 수학적 수반해를 정확하게 계산하는 직접 해법을 DIF3D 코드에 구현하고 검증 계산을 수행하였다. 이 직접 해법은 각 노드의 수반 부분 중성자류의 선형 조합을 이용하여 수학적 수반해를 정확히 계산한다. 미 방법에서는 수반 부분 중성자류의 선형 조합을 통하여 수학적 수반 방정식이 본래의 노달 방정식과 매우 유사한 형태로 변형되며, 그 결과 본래의 노달 방정식 해법이 최소한의 수정을 통해 수반 방정식 해결에 적용된다

• Fire Science and Engineering
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• v.27 no.2
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• pp.80-88
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• 2013

The validation of Fire Dynamics Simulator (FDS) was conducted for the under-ventilated fire in well-confined multi-compartments representative of nuclear power plant. Numerical results were compared with experimental data obtained by the OECD/NEA PRISME project. The effects of the numerical boundary conditions (B.C.) in ventilated system and the flame suppression model applied within FDS on the thermal and chemical environments inside the compartment were discussed in details. It was found that numerical B.C. on the vent flow resulting from over-pressure at ignition and under-pressure at extinction should be considered carefully in order to predict accurately the species concentrations rather than temperatures and heat fluxes inside the multi-compartment. The default information of suppression model applied within FDS resulted in artificial phenomena such as flame extinction and re-ignition, and thus the FDS results on the under-ventilated fire showed good agreement with the experimental results as the modified suppression criteria of the fuel used was adopted.

• Fire Science and Engineering
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• v.27 no.2
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• pp.31-39
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• 2013

The Fire Dynamics Simulator (FDS) has been applied to simulate a full-scale pool fire in well-confined and mechanically ventilated multi-compartments representative of nuclear power plant. The predictive performance of FDS was evaluated through a comparison of the numerical data with experimental data obtained by the OECD/NEA PRISME project. To identify clearly the FDS results regarding to the user-dependence in the process of FDS implementation except for the intrinsic limitation of FDS such as simple combustion model, only the over-ventilated fire condition was chosen. In particular, the importance of accurate boundary conditions (B.C.) in mechanically ventilated system were discussed in details. It was known from FDS results that the B.C. on inlet and outlet vents did significantly affect the thermal and chemical characteristics inside the compartments. Finally, it was confirmed that the FDS imposed an accurate ventilation B.C. provided qualitatively good agreement with temperatures, heat fluxes and concentrations measured inside the nuclear-type multi-compartments.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.5
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• pp.329-337
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• 2016

As a turbulence-enhancing device, a mixing vane, which is installed at a spacer grid of the fuel assembly, plays an important role in improving convective heat transfer by generating either swirl flow in the subchannels or cross flow between the fuel rod gaps. Therefore, both the geometric configuration and the arrangement pattern of a mixing vane are important factors in determining the performance of a mixing vane. In this study, in order to examine the flow-distribution features inside a $5{\times}5$ fuel assembly with split-type mixing vanes, which was used in the benchmark calculation of the OECD/NEA, we conduct simulations using the commercial computational fluid dynamics software, ANSYS CFX R.14. We compare the predicted results with measured data obtained from the MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, we discuss the effect of the split-type mixing vanes on the flow pattern inside the fuel assembly.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.765-774
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• 2009

For over 30 years, investigations of the thermohydraulic behavior of pressurized-water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany. The PKL facility models the entire primary side and significant parts of the secondary side of a of pressurized water reactor at a height scale of 1:1. Volumes, power ratings and mass flows are scaled with a ratio of 1:145. The experimental facility consists of four primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermohydraulic phenomena. The PKL tests began in the mid 1970s with the support of the German Research Ministry. Since the mid 1980s, the project has also been significantly supported by the German PWR operators. Since 2001, 25 partner organizations from 15 countries have taken part in the PKL investigations with the support and mediation of the OECD/ NEA (Nuclear Energy Agency). After an overview of PKL history and a short description of the facility, this paper focuses on the investigations carried out since the beginning of the international cooperation, and shows, by means of some examples, what insights can be derived from the tests.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2008.11a
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• pp.245-246
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• 2008

한국원자력연구원은 1997년부터 10여년 동안 정부 출연 연구기관으로서 원자력 발전의 부산물인 사용후핵연료를 포함한 다양한 방사성폐기물의 안전한 처분을 확인하기 위한 연구 개발에 주력해 왔다. 처분 안전성 확인(Safety Case)이란 개념은 지난 10여년 전부터 OECD/NEA와 SKB 등 서구 지역 국가들을 중심으로 발전된 처분 안전성에 대한 인허가 및 신뢰성 증진을 위한 통합적 접근 방안으로 미국, 영국 등 법률적 규정에 의거 안전성을 규제 기관과 법원 등이 개업된 청문회 등을 통해 다루는 체제와 상반되는 개념이다. 이러한 처분 안전성 확인은 처분 개념과 안전성 확인 개념화부터 인허가에 관련된 각종 보고서 작성과 이해성 증진을 위한 다양한 활동을 총합하는 개념으로 방사성폐기물 사업자가 아닌 객관적이고 전문성 있는 기관들이 추진하는 것이 바람직 할 것이다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.427-434
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• 1996

OECD-NEA 주관으로 수행된 TMI-2의 압력용기 변형연구의 결과, 하반부의 creep해석에 많은 문제점이 제기되어 있다. 본 논문은 TMI-2 노심용융 사고에 대한 기존 구조해석에서 creep 상관식의 형태, 적용방법 및 FEM 해석절차상의 상이점을 밝혀내고 이에 따라 압력용기 하반부의 파손확률이 크게 다르게 결정됨을 보였다. 기존의 TMI-2 구조해석에서 주 오차의 요인으로서 시간의 변화에 따른 국부열점 및 이를 포함한 재배치된 용융노심의 열경계조건의 불확실도와 압력용기강의 creep strain을 시간 및 온도에 대하여 불충분하게 묘사한 점을 밝혔다. 또한 creep-rupture 예측에 사용된 Larson-Miller Parameter도 해석을 지나치게 보수적인 결과로 유도하였다. 중대사고시 압력용기 하반부 천공방어를 위한 방안인 용기하부 외벽 냉각방식을 적용하였을 때 TMI-2 사고를 재해석한 결과, 압력용기의 건전성이 충분한 보수성을 가지고 유지됨을 보였다.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.719-726
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• 2012

The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.448-455
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• 2020

The H.B. Robinson Unit 2 (HBR-2) pressure vessel dosimetry benchmark is an in- and ex-Reactor Pressure Vessel (RPV) neutron dosimetry benchmark based on experimental data from the HBR-2 reactor, a 2300-MW PWR designed by Westinghouse and put in operation in March 1971, openly available through the SINBAD Database at OECD/NEA data Bank. The goals of the present work were to carry out three-dimensional (3D) fixed source transport calculations in both Cartesian (X,Y,Z) and cylindrical (R,θ,Z) geometries by using the TORT-3.2 discrete ordinates code on very detailed 3D HBR-2 geometrical models and to test the latest broad-group coupled (47 neutron groups + 20 photon groups) working cross section libraries in FIDO-ANISN format with same structure as BUGLE-96, such as BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7. The results obtained with all the cited libraries were satisfactory and are here reported and compared.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.635-640
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• 1998

Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of Thermal stratification is investigated using Theofanous's empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing.