Search NASA⌕ Search

SEARCH · Search NASA

Results for “Mechanistic reactor source term”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Planning Near-Term Tasks for the Development of Advanced Reactor Mechanistic Source Term Modeling and Simulation Tools

To assist both the advanced reactor industry and U.S. Nuclear Regulatory Commission (NRC) in the pursuit of reactor design and licensing, the U.S. Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has established a mechanistic source term (MST) research project under the auspices of the Application Drivers technical area. The MST research project is a collaboration between Argonne National Laboratory (Argonne) and Sandia National Laboratories (SNL) with a high-level objective to coordinate the development of comprehensive advanced reactor MST mod/sim capabilities to support risk-informed design and licensing decisions. As part of this effort, an MST mod/sim development pathway was developed in FY21, which outlines the high-level objectives and near-term tasks necessary to achieve the project objectives. Since publication of the development pathway, Argonne and SNL have been addressing the “near-term tasks” outlined in the report. The current provides a status update of the progress achieved in the fourth quarter of FY21. The report structure follows that of ref and is divided by advanced reactor type: High temperature gas reactor (HTGR), molten salt reactor (MSR) including fluoride salt-cooled high temperature reactor (FHR), and sodium fast reactor (SFR). Each section reviews the near-term tasks associated with the reactor type, current progress, and future plans. The near-term tasks associated with consequence modeling are reviewed in an upcoming work by SNL.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Progress on Near-Term Tasks for the Development of Advanced Reactor Mechanistic Source Term Modeling and Simulation Tools

To assist both the advanced reactor industry and U.S. Nuclear Regulatory Commission (NRC) in the pursuit of reactor design and licensing, the U.S. Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has established a mechanistic source term (MST) research project under the auspices of the Multiphysics Applications technical area. MST analysis is central to the reactor licensing process and a focus of the NRC regarding their mission to provide reasonable assurance of adequate protection of public health and safety and environment. The MST research project is a collaboration between Argonne National Laboratory (Argonne) and Sandia National Laboratories (SNL) with a high-level objective to coordinate the development of comprehensive advanced reactor MST mod/sim capabilities to support risk-informed design and licensing decisions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Progress on Long-Term Objectives for the Development of Advanced Reactor Mechanistic Source Term Modeling and Simulation Tools

To assist both the advanced reactor industry and U.S. Nuclear Regulatory Commission (NRC) in the pursuit of reactor design and licensing, the U.S. Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has established a mechanistic source term (MST) research project under the auspices of the Multiphysics Applications technical area. MST analysis is central to the reactor licensing process and a focus of the NRC regarding their mission to provide reasonable assurance of adequate protection of public health and safety and environment. The MST research project is a collaboration between Argonne National Laboratory (Argonne) and Sandia National Laboratories (SNL) with a high-level objective to coordinate the development of comprehensive advanced reactor MST mod/sim capabilities to support risk-informed design and licensing decisions. As part of this effort, an MST mod/sim development pathway was developed in FY21, which outlines the high-level objectives and near-term tasks necessary to achieve the project objectives. Since publication of the development pathway, Argonne and SNL have focused initial efforts on addressing the “near-term tasks” outlined in the report, and the current work provides a status update of the progress achieved in FY22. This report is a progress update on 6 tasks supporting molten salt reactor (MSR) and sodium fast reactor (SFR) MST analyses which were initiated in FY22 as recommended high priority near-term tasks. These tasks were continued into FY23 as they also support previously recommended long-term development objectives to advance MST analyses for these advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

A vital part of the licensing process for advanced (non-LWR) nuclear reactor developers in the United States is the assessment of the reactor’s source term, i.e., the potential release of radionuclides from the reactor system to the environment during normal operations and accident sequences. In comparison to source term assessments which follow a bounding approach with conservative assumptions, a mechanistic approach to modeling radionuclide transport, which realistically accounts for transport and retention phenomena, is expected to be used for advanced reactor systems. As the designs of advanced reactors increase in maturity and progress towards licensing, there is a need to advance modeling and simulation capabilities in analyzing the mechanistic source term (MST) of a prospective reactor concept. In the present work, a survey is provided of existing computational capabilities for the modeling of advanced reactors MSTs. The following reactors are considered: high temperature gas reactors (HTGR); molten salt reactors (MSR) which include salt-fueled reactors and fluoride salt-cooled high temperature reactors (FHR); and sodium- and lead-cooled fast reactors (SFR, LFR). A review of relevant codes which may be useful in providing information to MST analyses is also completed, including codes that have been used for source term analyses of LWRs, as well as those being developed for other aspects of advanced reactor system modeling such as reactor physics, thermal hydraulics, and chemistry. A discussion of MST modeling capabilities for each reactor type is provided with additional focus on important phenomena and functional requirements. Additionally, a comprehensive survey is provided of tools for consequence modeling such as atmospheric transport and dispersion (ATD).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A Pathway for the Development of Advanced Reactor Mechanistic Source Term Modeling and Simulation Capabilities

Source term analysis, or the estimation of the potential radionuclide release to the environment during reactor events, is a central focus of the reactor licensing process and a vital part of risk-informed reactor design approaches. A mechanistic source term (MST) analysis is designed to realistically model the release and transport of radionuclides from the source to the environment for specific scenarios, while accounting for retention or transmutation phenomena and associated uncertainties. The objective of MST analyses, in comparison to bounding or conservative source term assessments, is to provide a non-biased representation of reactor risk and improve the information available for siting, emergency planning, and reactor design decisions. In support of the advanced reactor community in its pursuit of MST analysis capabilities, this work outlines a recommended research pathway for the development of MST modeling and simulation (mod/sim) tools.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of a Thermochemical Database for Sodium Fast Reactor Mechanistic Source Term Calculations

An accurate assessment of the radionuclide source term for sodium fast reactors requires a predictive understanding of the chemical speciation and phase transformations of key dose contributing radionuclides, including those released to the sodium pool and cover gas region. Developing models for chemical speciation and phase equilibria requires the compilation of thermodynamic constants and experimentally determined functions that account for non-ideal solution behavior. For the present study, a thermochemical database, including parameters accounting for non-ideal chemical mixing, was compiled and used for a set of mechanistic source term trial calculations for a metal-fueled, pool-type sodium fast reactor [1].

Jerden, James↗
DOE OSTI2020

Functional Requirements for the Modeling and Simulation of Advanced (Non-LWR) Reactor Mechanistic Source Term

A number of industry vendors are leading a resurgence in advanced reactor development. As the assurance of public safety from the accidental release of radionuclides to the environment is central to regulatory licensing and the reactor design process, the development of a mechanistic source term (MST) assessment has been identified by the advanced reactor industry as a foremost priority. In contrast to historic light-water reactor (LWR) source term analysis efforts, an MST attempts to realistically model the release and transport of radionuclides from the source to the environment for specific scenarios, while accounting for retention or transmutation phenomena and associated uncertainties. This work seeks to aid in the development of MST modeling and simulation tools by performing a preliminary identification of functional requirements for the major advanced reactor types.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Regulatory Technology Development Plan Sodium Fast Reactor. Mechanistic Source Term Development

Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and containment. Radionuclides released from a primary sodium fire are also considered as potential sources. Any available experimental data and pertinent results relevant to the aforementioned phenomena are discussed, and operating incidents at domestically operated facilities are also examined. Considering the extensive range of phenomena affecting the release of radionuclides, the existing state of knowledge generally appears to be substantial, and may be sufficient in most areas. For core damage accidents, high retention rates should be expected within the fuel matrix and primary sodium coolant for all radionuclides other than the noble gases. These factors greatly reduce the magnitude of possible radionuclide release to the environment. Several possible gaps within the knowledgebase were identified during this effort. First, there are uncertainties with regard to radionuclide release from metal fuel in the molten state. Another knowledge gap appears in the available thermodynamic data regarding the behavior of lanthanides and actinides in liquid sodium. While not necessarily a phenomenological knowledge gap, a determination of the data requirements for MST development should be formally made prior to the expenditure of significant future research efforts. That is, if additional experimentation is performed in support of MST development, it is important to identify the proper quality assurance requirements for licensing

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Trial Calculation. Work Plan

The overall objective of the SFR Regulatory Technology Development Plan (RTDP) effort is to identify and address potential impediments to the SFR regulatory licensing process. In FY14, an analysis by Argonne identified the development of an SFR-specific MST methodology as an existing licensing gap with high regulatory importance and a potentially long lead-time to closure. This work was followed by an initial examination of the current state-of-knowledge regarding SFR source term development (ANLART-3), which reported several potential gaps. Among these were the potential inadequacies of current computational tools to properly model and assess the transport and retention of radionuclides during a metal fuel pool-type SFR core damage incident. The objective of the current work is to determine the adequacy of existing computational tools, and the associated knowledge database, for the calculation of an SFR MST. To accomplish this task, a trial MST calculation will be performed using available computational tools to establish their limitations with regard to relevant radionuclide release/retention/transport phenomena. The application of existing modeling tools will provide a definitive test to assess their suitability for an SFR MST calculation, while also identifying potential gaps in the current knowledge base and providing insight into open issues regarding regulatory criteria/requirements. The findings of this analysis will assist in determining future research and development needs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the current state of knowledge is extensive, and in most areas may be sufficient. Several knowledge gaps were identified, such as uncertainty in release from molten fuel and availability of thermodynamic data for lanthanides and actinides in liquid sodium. However, the overall findings suggest that high retention rates can be expected within the fuel and primary sodium for all radionuclides other than noble gases.

Brunett, Acacia J.↗
DOE OSTI2016

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

Bucknor, Matthew↗
DOE OSTI2023

Progress Report on Identification and Resolution of Gaps in Mechanistic Source Term Modeling for Molten Salt Reactors

This report summarizes FY23 activities to improve mechanistic source term modeling for MSR concepts. Relevant MELCOR capability enhancements made during FY23 are summarized including development of a flexible python-based EOS generator (MELEOS), porous domain modeling capabilities for validation applications, and development of a MELCOR model for the LSTL facility in anticipation of upcoming molten salt experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

Grabaskas, David↗
DOE OSTI2022

State of Knowledge on Aerosols and Bubble Transport for Mechanistic Source Term Analysis of Molten Salt Reactors

A comprehensive review of the current state of knowledge surrounding molten salt aerosol behavior and bubble transport in molten salt systems relevant to mechanistic source term (MST) analyses of molten salt reactors (MSR) and fluoride high temperature reactors (FHR) was provided. Although relatively little experimental work has been carried out on investigating these phenomena in molten salts compared to that for other nuclear fluids of interest (e.g., aqueous, sodium), the amount of work that has been reported still provides a baseline understanding to inform and design future experimental campaigns on these topics. The bulk of this data is derived from the Molten Salt Reactor Program (MSRP) which was carried out at Oak Ridge National Laboratory (ORNL) nearly 50 years ago, leading to the successful operation of the Molten Salt Reactor Experiment (MSRE). Many other related research efforts are also highlighted in this review, such as recent efforts under the DOE:NE Advanced Reactor Technology (ART) MSR Campaign. The goal of this report is to provide a foundation for modelers to base assumptions on, as well as to make recommendations to experimentalists attempting to close information gaps. As such, Section 4 contains recommendations for future experimental work that may be needed to close current information gaps related to MST modeling of MSRs. Additionally, given the flowing fuel nature of MSRs, the relevance of aerosol and bubble transport behavior in molten salts goes beyond just MST modeling, evidenced by experiences with the pump bowl and off-gas system of the MSRE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗