Publications

This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 18019and 21440. Revision 18019 represents the previous official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 21440. Along with the newly updated MELCOR Users’ Guide [2] and Reference Manual [3], users are aware and able to assess the new capabilities for their modeling and analysis applications.

This report represents an assessment of the gaps in Mechanistic Source Term (MST) and consequence assessment modeling for Molten Salt Reactors (MSRs). The current capabilities for MELCOR and the MELCOR Accident Code System (MACCS) are discussed, along with updates needed in order to address specific needs for MSRs. A test plan developed by Argonne National Laboratories is discussed as addressing some of these gaps, while some will require additional attention. Further recommendations are made on addressing these gaps. This report satisfies the DOE NE Milestone M2RD-21SN0601061 to leverage MELCOR and MACCS to identify parameters of importance for source term assessments for salt spill experiments.

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This report documents the progress made under the Terry Turbine Expanded Operating Band (TTEXOB) program's modeling and simulation (MODSIM) initiative at Sandia National Laboratories (SNL ). It describes the US Federal Fiscal Year 2021 (FY21) extended period-of-performance MODSIM work completed since the closure of FY20 with due reference to the Texas A&M University (TAMU) hybrid milestone 5/6 experimental program. This work, which falls under Milestone 7 of the program, provides a counterpart to the various experiments. The overall TTEXOB program and its milestone-based approach are described in the program's Summary Plan. Details of the individual milestone test plans can be found in the corresponding detailed test plan, e.g. the Milestone 3 and 4 Detailed Test Plan. SNL MODISM is conducted alongside experiments performed at TAMU, and SNL technical staff regularly consults with TAMU on the experimental program. In FY21, MELCOR code models and capabilities were exercised in two different contexts: experimental comparisons to the TAMU ZS-1 and GS-2, and stand-alone analyses of a station black-out (SBO) scenario in a generic boiling water reactor (BWR). Code to experiment comparisons met with fair success when turbine losses were well characterized as for the ZS-1 turbine. Both deterministic and Bayesian calibration processes were used to find a recommended turbine torque multiplier for ZS-1 type turbines. This process could be repeated for GS-2 type turbines if GS-2 losses were better understood. Stand-alone generic BWR SBO calculations revealed that three different modes of self-regulating turbopump behavior may be observed depending on certain modeling parameters and choices having to do with turbine nozzles. Aspects of this predicted behavior may have been observed in TAMU GS-2 experiments.

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This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 14959and 18019. Revision 14959 represents the previous official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 18019. Along with the newly updated MELCOR Users Guide and Reference Manual, users are aware and able to assess the new capabilities for their modeling and analysis applications.

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The work presented in this paper presents new techniques for modeling the combined use of the Reactor Core Isolation Cooling (RCIC) System and Accident Tolerant Fuel (ATF) in a Boiling Water Reactor (BWR). With guidance from Sandia National Laboratories' Severe Accident Analysis department, a MELCOR BWR model was developed from open source literature. The demonstration shown herein simulates BWR long-term station blackout (LTSBO) conditions with the Nuclear Regulatory Commission's (NRC) MELCOR severe accident analysis code. By combining state-of-the-art MELCOR modeling practices with new, physics-based RCIC System and ATF MELCOR inputs, this BWR model provides a contemporary platform for BWR severe accident simulations. The authors are investigating the combined use of the RCIC System and ATF as a means of passively enhancing reactor safety. The benefits of this approach were evaluated by performing simulations using traditional fuel designs (i.e. Zircaloy cladding) and ATF with an iron-chromium-aluminum (FeCrAl) clad under BWR LTSBO conditions. ATF performance was evaluated using severe accident metrics, specifically event sequence timings and the hydrogen production rate from cladding oxidation. Preliminary results show delayed core degradation timelines with less hydrogen production for ATF simulations. Although the results are limited in scope, the presented analysis could easily be expanded to a full-scale uncertainty study that considers a range of severe accident boundary conditions. This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the U.S. Department of Energy or the United States Government.

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This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 11932 and 14959. Revision 11932 represents the last official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 14959. Along with the newly updated MELCOR Users' Guide [2] and Reference Manual [3], users will be aware and able to assess the new capabilities for their modeling and analysis applications. Following the official release an addendum section has been added to this report detailing modifications made to the official release which support the accompanying patch release. The addendums address user reported issues and previously known issues within the official code release which extends the original Quick look document to also support the patch release. Furthermore, the addendums section documents the recent changes to input records in the Users' Guide applicable to the patch release and corrects a few issues in the revision 14959 release as well. This page left blank.

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This report provides an overview of technical issues and design features relevant to advanced reactors and reviews MELCOR's current readiness for modeling accidents in such reactor types. This report describes advanced reactor physics models currently available or under development, and gauges the level of effort required to develop new models and capabilities applicable to assessing advanced reactor safety issues. Finally, this report reviews the available database that can be used in verification and validation of new models. Four general advanced reactor types are considered in this report: 1) High Temperature Gas-Cooled Reactor (HTGR) 2) Sodium Fast Reactor (SFR) 3) Molten Salt Reactor (MSR) 4) Fluoride Salt-Cooled High Temperature Reactor (FHR)

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