• Nuclear Engineering and Technology
• /
• 제41권7호
• /
• pp.945-952
• /
• 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
• /
• 제15권4호
• /
• pp.219-228
• /
• 1983

릴리프 밸브의 차단까지 TMI-2 사고의 blowdown history를 검토하고 TMI-2 사고와 같은 소규모의 냉각제 상실 사고 동안 노심 파괴를 막기 위해 더 가산해야할 측정 기구에 대하여 논의하였다. 가산된 기구를 이용하여 어떻게 노심의 uncovered level과 operator time margin을 계산하는 가를 검토하였으며, TMI-2 사고에 대해 uncovered level과 operator time margin을 결정하기 위한 샘플 계산을 수행하였다. 이 방법을 이용해서 측정되는 변수들의 함수로써 uncovered level과 operator time margin을 보여주는 도표를 작성하였다.

• Nuclear Engineering and Technology
• /
• 제38권8호
• /
• pp.697-732
• /
• 2006

In this paper, a historical review of the developments in the safety of LWR power plants is presented. The paper reviews the developments prior to the TMI-2 accident, i.e. the concept of the defense in depth, the design basis, the large LOCA technical controversies and the LWR safety research programs. The TMI-2 accident, which became a turning point in the history of the development of nuclear power is described briefly. The Chernobyl accident, which terrified the world and almost completely curtailed the development of nuclear power is also described briefly. The great international effort of research in the LWR design-base and severe accidents, which was, respectively, conducted prior to and following the TMI-2 and Chernobyl accidents is described next. We conclude that with the knowledge gained and the improvements in plant organisation/management and in the training of the staff at the presently-installed nuclear power stations, the LWR plants have achieved very high standards of safety and performance. The Generation 3+LWR power plants, next to be installed, may claim to have reached the goal of assuring the safety of the public to a very large extent. This review is based on the historical developments in LWR safety that occurred primarily in USA, however, they are valid for the rest of the Western World. This review can not do justice to the many fine contributions that have been made over the last fifty years to the cause of LWR safety. We apologize if we have not mentioned them. We also apologize for not providing references to many of the fine investigations, which have contributed towards LWR safety earning the conclusions that we describe just above.

• Nuclear Engineering and Technology
• /
• 제44권7호
• /
• pp.807-816
• /
• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 추계학술발표회논문집(1)
• /
• pp.713-718
• /
• 1997

In the TMI-2 accident, approximately twenty(20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However, one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100$^{\circ}C$ for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant(KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options.

• 대한인간공학회지
• /
• 제20권2호
• /
• pp.1-28
• /
• 2001

This paper deals with the selection of the important Influencing Factors (IFs) under accident management situations in nuclear power plants for use in the assessment of human errors. In order to achieve this goal, we collected two types of IF taxonomies, one is the full set IF list mainly developed for human error analysis. and the other is the IFs for human reliability analysis (HRA) in probabilistic safety assessment (PSA). Five sets of IF taxonomy among the full set IF list and ten sets of IF taxonomy among HRA methodologies were collected in the study. From the review and analysis of BRA IFs, we could obtain some insights for the selection of HRA IFs. By considering the situational characteristics of the accident management domain, candidate IFs are chosen. Finally, those IFs are structured hierarchically to be appropriate for the use in the assessment of human error under accident management situation. Three nuclear accidents such as TMI. Chernobyl and JCO were analysed to validate the proposed taxonomy.

• Nuclear Engineering and Technology
• /
• 제51권8호
• /
• pp.1916-1938
• /
• 2019

The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor.

• Nuclear Engineering and Technology
• /
• 제36권5호
• /
• pp.430-449
• /
• 2004

A quantitative model for a control system that integrates human operators, systems, and their interactions is developed based on discrete functions. After identifying the major entities and the key factors that are important to each entity in the control system, a quantitative analysis to estimate the recovery failure probability from an abnormal state is performed. A numerical analysis based on assumed values of related variables shows that this model produces reasonable results. The concept of 'relative sensitivity' is introduced to identify the major factors affecting the reliability of the control system. The analysis shows that the hardware factor and the design factor of the instrumentation system have the highest relative sensitivities in this model. T도 probability of human operators performing incorrect actions, along with factors related to human operators, are also found to have high relative sensitivities. This model is applied to an analysis of the TMI-2 nuclear power plant accident and systematically explains how the accident took place.

• Journal of Radiation Protection and Research
• /
• 제12권1호
• /
• pp.40-47
• /
• 1987

우리나라에 현재 건설중인 원자력발전소 9/10호기와 동일형인 프랑스형 900 MWe PWR 에 대해 프랑스에서 TMI 사고이후 선원항을 보수적으로 설정한 RFS V.1.a의 가정에 따라 LOCA시의 핵분열생성물질방출분석과 그에 대한 파급효과를 평가 해석하였다. 방사능 환경방출에 의한 영향평가결과 주거제한구역경계 (500 m)에서 전신외부피폭선량은 사고발생후 2시간 경과시 0.66 rem이며 방사성 옥소의 방출에 의한 갑상선 피폭선량도 동일한 시간에서 유기성 옥소의 누출율이 l0%일때 13.5 rem 으로 사고시 피폭선량 제한치이하임이 나타났다. 그러나 격납용기외부로 누출되는 방사성 옥소중 유기성 옥소의 누출율이 갑상선의 방사선피폭에서 중요한 역활을 하고있음이 나타났으며 그 누출율이 10%이상이 될 경우 주거제한구역경계에서 사고시 갑상선 피폭선량제한치를 초과할 수도 있다는 가능성을 보여주고 있다.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 1993년도 하계학술대회 논문집 A
• /
• pp.525-528
• /
• 1993

KEPCO Research Center is carrying out a simulator(full scope replica type) development project for two nuclear power plants(Kori-2, Younggwang-3,4) and one fossil power plant(Poryong-3,4). In this project, we aim not only the installation of high performance simulators at the power plant sites but also the realization of self reliance of power plant simulation technology in Korea. In the course of preparing procurement specification for the 3 simulators, the present status of power plant simulation technology has been surveyed and is presented in this paper. The fidelity of simulation and the automation of simulation model production has been greatly improved due to the ever increasing computing power of today's workstations. The need and importance of the application of high fidelity simulators to the operator training is refocused since the accident at TMI Nuclear Power Plant, U.S.A.