• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.818-823
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• 1998

Three Mile Island Unit-2 (TMI-2)의 사고 후 OECD-NEA 주관의 연구에 의하면 압력용기 하부의 노즐이 국부열점(hot spot) 영역의 경우 거의 압력용기 바닥까지 용융되었음이 조사되었다. [1]. 이러한 재배치된 용융노심의 열속에 의하여 압력용기의 외부와 통하는 penetration tube weld(노즐 용접부)가 파손된다면 내부의 고압상태로 인해 penetration tube ejection 사고 및 이에 따르는 용융노심의 압력용기 외부로의 유출 가능성까지 배제할 수 없을 것이다. 본 연구의 출발점은 중대사고시 이러한 압력 및 열속에 따르는 노즐 용접부의 파손확률을 결정하는데 있다. 크리프 파출시 기존의 해석에서 쓰인 deterministic approach를 개선하여 probabilistic approach를 개발하였다. 또한 기존의 해석에서 쓰인 단순한 안전 여유도(margin-to-failure)의 개념과 비교하여 용접부에서의 파손확률을 계산하였다.

• Nuclear industry
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• v.36 no.4
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• pp.33-45
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• 2016

'기후 변화 대응을 위한 탈탄소 에너지로의 접근'을 주제로 한 해외 원자력 전문가 좌담회가 4월 20일 부산 BEXCO 컨벤션홀에서 열렸다. 한국원자력문화재단이 주최한 이번 좌담회는 당일 열린 '2016 한국원자력연차대회'에 참석한 해외 원자력 전문가를 초청하여 마련된 것으로, 민계홍 한국원자력산업회의 상근부회장(좌장 사회), 김호성 한국원자력문화재단 이사장, Geoffrey Rothwell OECD/NEA 수석경제연구원, Tomas $K{\aa}berger$ 스웨덴 Chalmers 공대 에너지정책학과 교수, Rashid Sarkar 방글라데시 공대 기계공학부 교수(원자력위원회 위원장 내정자) 등 원자력 전문가들이 참석하여 2시간에 걸쳐 의견을 나누었다. 좌담회 전문을 게재한다.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.981-988
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• 2018

An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with operator recovery actions in a pressurized water reactor. The relief valve of broken SG opened three times after the start of intact SG secondary-side depressurization as the recovery action. Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of high-pressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondary-side pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondary-side pressures through sensitivity analyses with the RELAP5 code.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.920-927
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• 2017

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3801-3807
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• 2023

Outcomes of the project "Comprehensive Verification of the FALCON Code for Calculation of Nuclear Fuel Temperature" relating to calculation of fuel temperature during relatively fast thermal transients are presented. Good prediction capabilities of the FALCON MOD01 code coupled with the GRSW-A code are shown as applied to the data of the TF3 and TF5 tests from the Transient Temperature Experiment IFA-507. The IFA-507 related dataset of the OECD/NEA International Fuel Performance Experiments (IFPE) Database is extended by the reconstructed dynamics of the axial power distribution in the rods during the transient phase of the experiment. Based on the code calculation, the time constant of the thermal fuel response to a power transient is estimated.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.945-952
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• 2009

The Three Mile Island Unit 2 (TMI-2) accident provides unique full scale data, thus providing opportunities to check the capability of codes to model overall plant behavior and to perform a spectrum of sensitivity and uncertainty calculations. As part of the TMI-2 analysis benchmark exercise sponsored by the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD NEA), several member countries are continuing to improve their system analysis codes using the TMI-2 data. The Republic of Korea joined this benchmark exercise in November 2005. Seoul National University has analyzed the TMI-2 accident as well as the currently proposed alternative scenario along with a sensitivity study using the Modular Accident Analysis Program Version 4.03 (MAAP4.03) code in collaboration with the Korea Hydro and Nuclear Power Company. Two input files are required to simulate the TMI-2 accident with MAAP4: the parameter file and an input deck. The user inputs various parameters, such as volumes or masses, for each component. The parameter file contains the information on TMI-2 relevant to the plant geometry, system performance, controls, and initial conditions used to perform these benchmark calculations. The input deck defines the operator actions and boundary conditions during the course of the accident. The TMI-2 accident analysis provided good estimates of the accident output data compared with the OECD TMI-2 standard reference. The alternative scenario has proposed the initial event as a loss of main feed water and a small break on the hot leg. Analysis is in progress along with a sensitivity study concerning the break size and elevation.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4367-4381
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• 2023

Nuclear power plants are increasingly being equipped with digital I&C systems. Although some probabilistic safety assessment (PSA) models for the digital I&C of nuclear power plants have been constructed, there is currently no specific internationally agreed guidance for their modeling. This paper presents an initiative by the OECD Nuclear Energy Agency called "Digital I&C PSA - Comparative application of DIGital I&C Modelling Approaches for PSA (DIGMAP)", which aimed to advance the field towards practical and defendable modeling principles. The task, carried out in 2017-2021, used a simplified description of a plant focusing on the digital I&C systems important to safety, for which the participating organizations independently developed their own PSA models. Through comparison of the PSA models, sensitivity analyses as well as observations throughout the whole activity, both qualitative and quantitative lessons were learned. These include insights on failure behavior of digital I&C systems, experience from models with different levels of abstraction, benefits from benchmarking as well as major contributors to the core damage frequency and those with minor effect. The study also highlighted the challenges with modeling of large common cause component groups and the difficulties associated with estimation of key software and common cause failure parameters.

• Journal of Radiation Protection and Research
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• v.21 no.3
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• pp.183-200
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• 1996

Many information channels have dealt with the radiological consequences of the Chernobyl accident in different voices ever since the time of the accident. Large differences in the data about the amount of released radioactivity, losses of life, environmental effects and economic damage confuse the information receiving group. The intention of this paper is to provide an insight to the consequences of the accident through review of the reports and articles on the given issue and the scientific background. The formal reports reviewed include those from IAEA, EC, OECD/NEA, the governments of the two most-affected countries; Belarus and Ukraine. Much consideration was paid to make the text as plain as possible.

• Mineral and Industry
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• v.21 no.1
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• pp.56-66
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• 2008

이 동향자료는 Geoscience Australia에서 2001년 발간한 "Geoscience Australia Mineral Resources Report No.1"과 역시 동 기관에서 2007년 발행한 "Australia's Identified Mineral Resources 2007" 중 일부 내용을 발췌하여 정리 한 것으로서 요약하면 다음과 같다. OECD/NEA와 IAEA(2000)는 세계적으로 분포하는 우라늄 광상유형을 지질학적인 형성환경에 따라 15개 유형으로 분류하였으며 호주에서는 각력복합형, 부정합형, 사암형, 지표형, 변성교대형, 변성형, 화산형, 관입형 및 맥상형이 보고되어 있다. 유형별 자원량은 각력복합형, 부정합형 및 사암형 3개 유형 광상이 약 93%를 차지하며, 각력복합형광상의 자원량이 63%에 달한다. 현재 개발되는 광상은 각력복합형의 올림픽댐 광산, 부정합형인 레인저 광산 및 사암형인 베벌리 광산이다. 호주는 세계 총 우라늄 자원량의 27%를 보유하고 있어 세계 1위를 차지한다. 올림픽댐광상이 항내채광을 하는 우라늄 광산으로서는 세계에서 가장 큰 광상으로서 US$80 이하에서 회수가능한 RAR(적정확정자원량)이 476,000톤이다. 이 자원량은 세계 총 자원량의 18%를 차지하며, 단일 광산으로서는 세계최대규모이다. 2006년 호주의 우라늄 총생산량은 $U_3O_8$ 8,943톤(7,584톤 U)이며 이는 세 광산에서 생산된 양으로서 캐나다에 이어 두 번째로 많은 양(세계 생산량의 19%)이다. 2006년 우라늄 수출량은 $U_3O_8$ 8,660톤(7,344톤 U)이며 수출가는 호주달러 5억2천9백만불에 달한다. 호주는 우라늄 수출국들과 "원자력협력협약"을 맺어 평화적 목적을 위해서만 공급한다는 단서를 달고 있으며 IAEA에 의해 관리/감독되고 있다. 최근 호주 정부는 지구과학연구소에 많은 예산을 투여하여 육상에너지안전대책을 발의하여 자원개발에 요구되는 탐사자료 확보에 주력하고 있다.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.28 no.5
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• pp.186-194
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• 2016

As a turbulence-enhancing device, a mixing vane installed at a spacer grid of the fuel assembly plays a role in improving the convective heat transfer by generating either swirl flow in the subchannels or cross flow between fuel rod gaps. Therefore, both configuration and arrangement pattern of a mixing vane are important factors that determine the performance of a mixing vane. In this study, in order to examine the flow distribution features inside $5{\times}5$ fuel assembly with swirl-type mixing vanes used in benchmark calculation of OECD/NEA, simulations were conducted with commercial CFD software ANSYS CFX R.14. Predicted results were compared to data measured from MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, the effect of swirl-type mixing vanes on flow pattern inside the fuel assembly was described.