三哩島事故後，核能管制委員會（U.S. Nuclear Regulatory Commission , USNRC）於1982年要求電力公司必須依據NEDO-22215建立冷卻水及圍阻體內放射性核種的濃度，評估爐心損毀程度的能力，此種事故中冷卻水及圍阻體內放射性物質的濃度測量須透過「事故後取樣系統（Post Accident Sampling System, 以下簡稱PASS ）」進行。但PASS取樣分析耗費相當長的時間，不足以滿足爐心損毀程度評估的及時性要求；依據PASS提供的資料所做的分析，也無法提供正確的資訊決定爐心損毀程度。
NEDO-22215所提出之爐心損毀程度評估方法是根據當時對爐心損毀以及放射性物質外釋過程的認知。經過多年的研究發展，核能界對核能電廠爐心損毀及放射性物質外釋過程已經有了新的瞭解。1980年代中期所發展爐心損毀程度評估方法可能已經不適用。
本研究的目的即是依照BWROG新的爐心損毀程度評估指引NEDC-33045P(已經為美國核管會所接受)建立適用於核一廠的爐心損毀程度評估導則。新的指引僅依靠核電廠已有之儀控系統所提供的資料，如壓力槽水位、圍阻體輻射強度及圍阻體氫氣濃度，即可以決定爐心損毀程度，使電廠人員能在事故正進行時，決定爐心熔損程度，及時決定應採取之適當的緩和及保護行動。
本研究中亦利用MAAP4.04程式，分析核能一廠具代表性之事故序列例如破管、暫態等高、低壓事故，瞭解嚴重事故中各主要參數；例如分裂產物外釋情形及其分佈、燃料棒護套氧化程度、氫氣產生量及其分佈情形與爐心損毀程度的關係，做為運轉人員及技術支援中心 (Technical Support Center, TSC) 成員的參考。

US Nuclear Regulatory Commission(USNRC) and Nuclear Industry had spent an extensive effort to investigate and study the Three-Mile Island Accident since it happened in 1979。Through this effort，USNRC has set many mandatory requirements to keep the core damage accident as that of Three-Mile Island from happening again in nuclear power plants 。One of them is to request utilities establishing，based on NEDO-22215，the capability of obtainig the concentration of radioisotopes of reactor coolant and containment to evaluate the extent of core damage during and after severe accident。
The capability is achieved by setting up Post Accident Sampling System (PASS)，which can obtain the concentration of radioisotopes of reactor coolant and containment。But after more than 20 years of PASS sampling experience，it has been found that PASS sampling is a time consuming job， which could not give plant personnel the current information about the core conditions during the accident to take correct mitigation actions to cope with the accident。
The purpose of this study is to develop the plant specific core damage assessment guideline of Chinshan Nuclear Power Plant based on the generic guideline NEDC-33045P of US BWR Owners Group。In the new guideline the on-line plant instrumentation such as reactor water level，radiation level and hydrogen concentration level of primary containment are used to assess the extent of core damage。The new guideline can help plant personnel make a quick assessment of the extent of core damage and take correct actions to mitigate the accident and protect the health and safety of public in a timely manner。
This study also utilizes MAPP4.04 program to analyze the representative sequences series of high/low pressure accidents such as loss of coolant accident and transients。The purposes of analysis are to understand the primary parameters as the release fraction of fission products、the extent of oxidation of fuel rod cladding、and hydrogen production and their relations to the extent of core damage。These information can be used by Technical Support Center personnel and reactor operation personnel during and after an un-antipated severe accident。

[1] U.S. Nuclear Regulatory Commission, ”Post Accident Sampling Guide for Preparation of a Procedure to Estimate Core Damage,” Supplement to NUREG-0737, March 1982.
[2] U.S. Nuclear Regulatory Commission, “ Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,” NUREG-0654, Rev.1, October 1980.
[3] “ Methodology for Development of Emergency Action Levels,” NUMARC/ /NESP-007, April 1990.
[4] “美國 NUMARC/NESP-007 核電廠緊急事故類別研判與應用,” 全委會作業執行室, 中華民國九十年十一月.
[5] General Electric, “Procedures For The Determination Of The Extent Of Core Damage Under Accident Conditions” NEDO-22215, August 1982.
[6] U.S. Nuclear Regulatory Commission, ” Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors” Regulatory Guide 1.3, Rev.2, June 1974.
[7] U.S. Nuclear Regulatory Commission, “Accident Source Terms for Light-Water Nuclear Power Plant,” NUREG-1465, February 1995.NUREG-1465
[8] NEDC-33045P，”Methods of Estimating Core Damage in BWRs” General Electric, July 2001.
[9] U.S. Nuclear Regulatory Commission, “Clarification of TMI Action Plan Requirements, ” NUREG-0737, November 1980.
[10] “Estimation of Fission Product and Core material Characteristics,” IDCOR Technical Reports 11.1, 11.4 and 11.5, October 1983.
[11] U.S. Nuclear Regulatory Commission, “Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors” Regulatory Guide 1.4, Rev.2, June 1974.
[12] J.J. DiNunno et al., “Calculation of Distance Factors for Power and Test Reactor Sites,” Technical Information Documents (TID)-14844, U.S. Atomic Energy Commission 1962.
[13] U.S. Nuclear Regulatory Commission, “Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants,” WASH-1400 (NUREG-75/014), December 1975.
[14] J.A. Gieseke, et.al.,” Source Term Code Package (STCP)：A User’s Guide(MOD1),” Battelle Columbus Laboratories, NUREG/CR-4587, July 1986.
[15] R.M. Summers, et.al., “MELCOR 1.80：A Computer Code for Nuclear Severe Accident Source Term and Risk Assessment Analysis,” Sandia National Laboratories, NUREG/CR-5531, 1991.
[16] “ Modular Accident Analysis Program for LWR Power Plants : MAAP4,“ Fauske & Associates Inc., May 1994.
[17] “Severe Accident Risks：An Assessment for Five U.S Nuclear Power Plants, Final Summary Report,” U.S. Nuclear Regulatory Commission, NUREG-1150, December, 1990.
[18] 核一廠程序書1413 “事故後取樣作業程序”
[19] U.S. Nuclear Regulatory Commission,10CFR50.47 “Emergency plans.”
[20] U.S. Nuclear Regulatory Commission, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident” Regulatory Guide 1.97, Rev.3, May 1983.
[21] “Severe Accident Management Technical Basis Report,“ Electric Power Research Institute, TR-101969, 1994
[22] “ RTM-96, Response Technical Manual ”, U.S. Nuclear Regulatory Commission, NUREG/BR-0150, Vol.1, Revision 4, March 1996.
[23] “ Introduction to Nuclear Engineering ”,John. R. Lamarsh, 2nd Edition, 1985.