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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.

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At least 199 records · Page 11
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DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a HTGR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts over the past decades have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near-term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR ha s undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform high-temperature gas reactor (HTGR) safety evaluations. The generic plant model is based on publicly available PMBR-400 design information. For plant aspects (e.g., reactor building size and leak rate) that are not described in the PBMR-400 references, the analysts made assumptions needed to construct a MELCOR full-plant model. The HTGR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to a recuperative heat exchange r. Surrounding the reactor vessel is a reactor cavity contained within a confinement room cooled by the Reactor Cavity Cooling System (RCCS). Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, primary system heat removal and reactor vessel decay heat removal, and confinement cooling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Modeling a Hypothetical Fusion Blanket Design Using MELCOR-TMAP

MELCOR (not an acronym) is a nuclear safety code developed by Sandia National Laboratories for the Nuclear Regulatory Commission. Idaho National Laboratory collaborates with Sandia to maintain and develop a version of MELCOR specifically for fusion device applications. Most recently, the tritium migration analysis program (TMAP) has been incorporated into MELCOR for fusion, in a version called MELCOR-TMAP. We show the results applying MELCOR-TMAP to model steady-state inventory and tritium retention in a liquid immersion blanket design for a hypothetical fusion device under steady-state operation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Radiological Releases from Novel Fuel Forms in Advanced Reactors During Severe Accidents for Consequence Analyses

Various advanced reactor developers are exploring the potential for reductions in the size of physical security forces and emergency planning zones. These reductions are based on robust fuel forms and inherently safe reactor designs. However, such reductions in physical protection measures could increase the risk of sabotage. To assess the possibility of reducing these measures, sabotage-induced radiological consequence analyses were carried out. These analyses considered accident scenarios that were beyond design basis accidents and overly conservative (Shah, 2025a; Shah, 2025b; Shah and Hartanto, 2026), yielding very large release fractions. These fractions, which can be used to evaluate physical protection and emergency planning requirements, have been crudely determined and applied as demonstrations for a sodium-cooled fast reactor (SFR) (Shah and Hartanto, 2025a), a high-temperature gas-cooled reactor (HTGR) (Shah and Hartanto, 2025b), a heat pipe–cooled reactor (HPR) (Shah and Hartanto, 2025c), and a molten salt–cooled reactor (MSR) (Shah et al., 2026). A Sandia National Laboratories (SNL) team used MELCOR—a fully integrated severe accident analysis code—to demonstrate the code’s capability to analyze advanced (i.e., not light water–cooled) reactors (including a fluoride salt–cooled high-temperature reactor [FHR]) and calculate radiological releases to the environment during severe accidents (Wagner et al., 2022a, 2022b, 2022c, 2023a, and 2023b). Although the analyses were carried out to demonstrate MELCOR’s growing capability, the release source terms were estimated for advanced reactors, providing valuable insights into the accident progression and radiological releases. These findings from prior SNL studies, including estimated source terms and related sensitivity studies, were leveraged to derive source terms for postulated sabotage-induced accidents. Insights from these sensitivity studies informed the scaling of SNL’s estimated source terms for the defined accident scenarios. The derived release fractions for the severe accident scenarios for the respective reactor designs can be used to perform more nuanced dose consequence analyses to evaluate the reactors’ physical protection and emergency planning zone requirements. These analyses are in accordance with the risk-informed, performance-based approach proposed under 10 CFR Part 53. This study builds on the prior source term analyses and associated sensitivity studies by SNL to derive time-dependent and design-informed release fractions. Section 2 describes the diverse advanced reactor designs analyzed by the SNL team. Section 3 discusses the severe accident analyses, the release fractions calculated, and the limitations and assumptions of the demonstration project. Section 4 presents the release percentages derived for the hypothetical sabotage-induced severe accidents at the advanced reactors. Section 5 summarizes the study’s findings and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Assessment of MELCOR condensation models with the presence of noncondensable gas in natural convection flow regime

Condensation of steam vapor is an important mode of energy removal from the reactor containment. The presence of noncondensable gas complicates the process and makes it difficult to model. MELCOR, one of the more widely used system codes for containment analyses, uses the heat and mass transfer analogy to model condensation heat transfer. To investigate previously reported nodalization-dependence in natural convection flow regime, MELCOR condensation model as well as other models are studied. The nodalization-dependence issue is resolved by using physical length from the actual geometry rather than node size of each control volume as the characteristic length scale for MELCOR containment analyses. At the transition to turbulent natural convection regime, the McAdams correlation for convective heat transfer produces a better prediction compared to the original MELCOR model. The McAdams correlation is implemented in MELCOR and the prediction is validated against a set of experiments on a scaled AP600 containment. The MELCOR with our implemented model produces improved predictions. For steam molar fractions in the gas mixture greater than about 0.58, the predictions are within the uncertainty margin of the measurements. The simulation results still underestimate the heat transfer from the gas-steam mixture, implying that conservative predictions are provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Sodium Fire Collaborative Study Progress (CNWG FY 2019)

This report discusses the progress on the collaboration between Sandia National Laboratories (Sandia) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2019. First, the current sodium pool fire model in MELCOR, which is adapted from CONTAIN-LMR code, is discussed. The associated sodium fire input requirements are also presented. A proposed model improvement developed at Sandia is discussed. Finally, the validation study of the sodium pool fire model in MELCOR carried out by a JAEA's staff is described. To validate this model, a JAEA sodium pool fire experiment (F7-1 test) is used. A preliminary calculation is performed using a modified MELCOR model from a previous experiment simulation. The results of the calculation are discussed as well as suggestions for improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2020.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Sodium Fire Collaborative Study Progress -- CNWG Fiscal Year 2020

This report discusses the progress on the collaboration between Sandia National Laboratories (Sandia) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2020. First, the current sodium pool fire model in MELCOR, which is adapted from CONTAIN-LMR code, is discussed. The associated sodium fire input requirements are also presented. These input requirements are flexible enough to permit further model development via control functions to enhance the current model without modifying the source code. The theoretical pool fire model improvement developed at Sandia is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR carried out by both Sandia and JAEA’s staff is described. To validate this pool fire model with the enhancement, a JAEA sodium pool fire experiment (F7-1 test) is used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2021.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modifications to MELCOR 1.8.6 for modeling Accident Tolerant Fuel in Boiling Water Reactors

This work describes further modifications made to a previously implemented User-Defined Generalized Coating (UDGC) model in MELCOR 1.8.6. The UDGC version of the code allows for the modeling of cladding and canister materials other than Zircaloy (without overwriting Zircaloy properties) as well as coated Zircaloy cladding and canisters, with the thermophysical and oxidation properties of the clad or coating material specified by the user. Specific changes outlined in this report are the extension of the UDGC models to BWR canisters, and generalization of the oxidation model for all UDGC components to accommodate different oxidation rate correlations in two different temperature ranges. User input requirements are specified and an example case is described that illustrates the potential impact of model assumptions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Probabilistic Economic Valuation of Safety Margin Management

The goal of this project was to explore a new methodology for evaluation of potential safety margin upgrades to add a new instrument to the Risk Informed Safety Margin Characterization (RISMC) toolkit. To this end, the RAVEN code developed by the Idaho National Laboratory was used to perform a cost-benefit analysis of several potential power plant safety upgrades to compare the material cost of the upgrades with the probabilistic benefit of avoiding accident scenario consequences to find the most cost-effective safety upgrades. By coupling RAVEN to MELCOR to drive Monte-Carlo method stochastic parameter analysis and using data from existing nuclear codes, as well as MELCOR’s radionuclide relocation tracking capabilities, the radionuclide release source term activity in a severe accident scenario can be estimated. Severe Accident Mitigation upgrades were benchmarked against the Hardened Containment Ventilation upgrade mandated by the USNRC for Mk1 BWRs using dollars spent per Curie of release prevented as a metric by which to evaluate the economic efficacy of the upgrades. Additionally, Fault Tree/Event Tree-based analyses were performed to compare older methods of evaluation to the methodology pioneered in this work, and Monthly Operating Report (MORP) outage data was analyzed for insights into power plant outage optimization and reduction.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Long-term Prediction of Emissivity of Structural materials for High Temperature Reactor Systems

The purpose of this research was to extend the recent short-term studies of emissivity of reactor materials performed independently at the University of Missouri and University of Wisconsin to tests aimed at longerterm predictions of emissivity. The University of Missouri measured hemispherical emissivity, which provides an integrated measurement, while the University of Wisconsin measured spectral emissivity (2 to 10 μm wavelength) from which the integrated quantity can be obtained. Hemispherical emissivity is needed for MELCOR, RELAP and FLUENT type heat transfer code calculations, while spectral emissivity is needed for a detailed Monte Carlo Transport theory computation of heat transfer and fundamental modeling verification. The objectives of the study were: Objective 1: Defining materials for long-term emissivity in cross-cutting reactor applications - procurement, preparation, and analyses of materials. Objective 2: High temperature materials testing in various environments and post-test surface analysis Objective 3: Independent hemispherical and spectral emissivity testing. Objective 4: Development of models to gain a predictive capability for long-term emissivity and application and validation in current RCCS research programs. The research has been performed at two universities over a four-year term with the participation of students, scientific staff, and faculty. The report describes progress in realizing the objectives, and the accomplishments at the two universities. There was considerable progress with respect to Objectives 1-3. The model development was redirected towards modeling of interrelationships between the spectra and hemispherical emissivities and uncertainty analysis of total hemispherical emissivity measurements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s ability to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

MELCOR

MELCOR is a fully integrated, engineering-level computer code for modeling the progression of severe accidents in light water reactors (LWR) at nuclear power plants and nuclear fuel cycle facilities. Originally developed to assess severe accidents following Three Mile Island, MELCOR’s flexible modeling framework has enabled it to be applied to safety assessments of a much broader range of nuclear power reactor designs and other types of nuclear facilities processing radioactive material. Further, MELCOR can model a broad spectrum of severe accident phenomena such as thermal-hydraulic response in a reactor coolant system; core heat-up, degradation, and relocation; and transport behavior in both boiling water and pressurized water reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Uncertainty quantification in MELCOR Safety analysis of ARIES reactor designs

MELCOR-TMAP is a combined thermal-hydraulics and tritium tracking code developed to simulate severe accident scenarios in fission and fusion power plants. Here, we demonstrate the results of MELCOR-TMAP analyses on historical ARIES program reference designs. By coupling MELCOR-TMAP with the open source RAVEN probabilistic risk analysis framework’s Bayesian UQ capabilities, we also demonstrate key uncertainties in material properties with the highest impact on tritium inventory and plant risk.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Analysis of Time Dependent Data and PRA

In the past decades, several numerical simulation codes have been employed to simulate accident dynamics (e.g., RELAP5-3D, MELCOR, MAAP). In order to evaluate the impact of uncertainties into accident dynamics, several stochastic methodologies have been coupled with these codes. These stochastic methods range from classical Monte-Carlo and Latin Hypercube sampling to stochastic polynomial methods. Similar approaches have been introduced into the risk and safety community where stochastic methods (such as RAVEN, ADAPT, MCDET, ADS) have been coupled with safety analysis codes in order to evaluate the safety impact of timing and sequencing of events. These approaches are usually called Dynamic PRA or simulation-based PRA methods. These uncertainties and safety methods usually generate a large number of simulation runs (database storage may be on the order of gigabytes or higher). The scope of this paper is to present a broad overview of methods and algorithms that can be used to analyze and extract information from large data sets containing time dependent data. In this context, 'extracting information' means constructing input-output correlations, finding commonalities, and identifying outliers. Some of the algorithms presented here have been developed or are under development within the RAVEN statistical framework. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

The MELCOR Plot File Format

This work provides a detailed explanation of the MELCOR Plot File format. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants, and it primarily exports its simulation data to a binary file in the Plot File format. This work documents the Plot File's overall structure and the data types of exported information. The process of interpreting and associating time series information with MELCOR's Plot Variables for complete time histories is also presented. The format and meaning of time-independent information, called Special information, is also discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Heat up and failure of BWR upper internals during a severe accident

In boiling water reactors, the shroud dome, separators, and dryers above the core are made of approximately 100,000 kg of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. In this scenario, the upper internals can also be heated by thermal radiation from the hot degrading core. Historically, models of the upper internals have been relatively simple in severe accident codes. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics and no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. The MELCOR modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. Consistent with a previous study, the results indicate that the upper internals can reach high temperatures during a severe accident sufficient to lose their structural integrity and relocate. Finally, the additional 100 metric tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Deuterium Extraction from Helium with a Vanadium Vacuum Permeator: Commissioning of the Tritium Extraction eXperiment (TEX)

Here, the Tritium Extraction eXperiment (TEX) is a forced-convection lead-lithium (PbLi) loop in the Safety and Tritium Applied Research (STAR) facility at Idaho National Laboratory (INL) with the purpose of providing validation data for the vacuum permeator tritium extraction concept. A vanadium tube of 1000 mm length, 12.7 mm outside diameter, and 0.50 mm wall thickness is installed in the test section of TEX. The installed vanadium tube is characterized to quantify impurity concentrations, surface chemistry, and microstructure to elucidate permeation phenomena observed in experimentation. Herein, the permeation properties of the vanadium tube are characterized by measuring deuterium permeation at 300 °C, 325 °C, and 350 °C at 100 kPa, 125 kPa, and 150 kPa total pressures with 5000 ppm deuterium in helium gas mixture in a once-through flow configuration. The hydrogen isotope permeation through the vanadium tube in the test section is measured with quadrupole mass spectrometers and the hydrogen isotope concentration in the feed and retentate gas stream is measured with gas chromatography. The transient permeation results are modeled with MELCOR-TMAP, a thermal-hydraulic tritium transport code, and compared well with literature data.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Initial development of a generic fluoride salt-cooled reactor model

Fluoride high-temperature reactors (FHRs) are high-temperature, low-pressure reactor concepts that use tri-structural isotropic (TRISO) fuel and molten fluoride salt coolant. These reactors have the potential to provide both electrical power and high-temperature process heat. We used generic FHR parameters for a pebble-bed FHR to develop an initial model with fresh fuel for a generic FHR (gFHR) in MELCOR and SCALE (NEWT and KENO). In this paper, we present the development of our gFHR models, which will serve as the baseline for a sensitivity and uncertainty analysis to quantify the range of possible source terms for FHRs in severe accidents. We present MELCOR results for fuel and coolant temperatures through the core, a nodalization study for the steady-state thermal hydraulic model, and development of reactor physics models in SCALE. As this work progresses, these models will be used to calculate source terms for a loss-of-forced-flow accident and to conduct a sensitivity study on this accident to establish a range of possible source terms. SCALE will provide reactor physics parameters like isotopic inventory, decay heat generation, and temperature coefficients of reactivity. Using the uncertainty quantification tools within SCALE, we will generate distributions for those parameters and will use the uncertainty quantification code RAVEN or DAKOTA to sample those distributions in MELCOR to quantify the impact of reactor physics and thermal hydraulic uncertainties on FHR source terms. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Data Analysis Approaches for the Risk-Informed Safety Margins Characterization Toolkit

In the past decades, several numerical simulation codes have been employed to simulate accident dynamics (e.g., RELAP5-3D, RELAP-7, MELCOR, MAAP). In order to evaluate the impact of uncertainties into accident dynamics, several stochastic methodologies have been coupled with these codes. These stochastic methods range from classical Monte-Carlo and Latin Hypercube sampling to stochastic polynomial methods. Similar approaches have been introduced into the risk and safety community where stochastic methods (such as RAVEN, ADAPT, MCDET, ADS) have been coupled with safety analysis codes in order to evaluate the safety impact of timing and sequencing of events. These approaches are usually called Dynamic PRA or simulation-based PRA methods. These uncertainties and safety methods usually generate a large number of simulation runs (database storage may be on the order of gigabytes or higher). The scope of this paper is to present a broad overview of methods and algorithms that can be used to analyze and extract information from large data sets containing time dependent data. In this context, “extracting information” means constructing input-output correlations, finding commonalities, and identifying outliers. Some of the algorithms presented here have been developed or are under development within the RAVEN statistical framework.

97 MATHEMATICS AND COMPUTING↗