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The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). SDTP consists of nearly 60 organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, and the United States. Innovative Systems Software (ISS) is the administrator for the program. RELAP/SCDAPSIM is used by program members and licensed users to support a variety of activities. The paper provides a brief review of some of the more important activities including the analysis of research reactors and Nuclear Power Plants, design and analysis of experiments, and training.

This report was prepared as an account of work sponsored by an agency of the United States Government Neither the United States Government nor any agency thereof, nor any of their cmpioycu, makes any warranty, express or implied. or ujumts any legal liability or rrsponsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its w would not infringe privately owned rights. Reference herein to any specific commercial product, process, or Icrvicc by trade name, trademark, manufac-Cum, or othemke doej not necessarily constitute or imply its endorsement, ncommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors c x p d hemn do not nrmcarily state or reflect thosc of the United States Government or.any agency thereof.. . Contents Department of Energy (DOE) contracted the University of California-Santa Barbara (UCSB) to produce the peer-reviewed report, DOE/ID-10460.2 To assist in the Nuclear Regulatory Commission's (NRC's) evaluation of IVR of core melt by exvessel flooding of the AP600, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform the following: ERVC effectiveness. method to identify areas where technical concerns weren't addressed. margins to failure and conclusions presented in the UCSB study. An in-depth critical review of the UCSB study and the model that UCSB used to assess An in-depth review of the UCSB study peer review comments and of UCSB's resolution An independent analysis effort to investigate the impact of residual concerns on the This Technical Evaluation Report (TER) summarizes results from these tasks. As discussed in Sections 1.1 and 1.2, INEEL'S review of the UCSB study and peer reviewer comments suggested that additional analysis was needed to assess: (1) the integral impact of peer reviewer-suggested changes to input assumptions and uncertainties and (2) the challenge present by other credible debris configurations. NRC tasked INEEL to perform an independent assessment to address this need. Section 1.3 summarizes the corresponding analysis approach developed by INEEL. The remainder of this report provides more detailed descriptions of INEEL's analysis methodology, input assumptions, and results. 1.1 Review of UCSB Study and Model As noted in Reference 2, the objective of the UCSB study was to demonstrate the effectiveness of ERVC for an AP600-like design and to provide a readily adaptable path for demonstrating ERVC for other reactor designs. Figure 1-1 illustrates the UCSB approach for demonstrating AP600 vessel integrity for cases with complete RCS depressurization and ERVC. As indicated in this figure, the UCSB study attempts to demonstrate vessel integrity by proving two assertions: Assertion 1: For all heat fluxes at or below the critical heat flux (0, the corresponding Assertion 2: Heat fluxes from relocated melt to the lower head always remain below the CHF. minimum vessel wall thicknesses are sufficient that the...

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Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.