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At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
NASA NTRSDate not supplied

Modular Assembled Radiators for Nuclear Electric Propulsion (NEP) VehicLes (MARVL) Project Thermal Technology Development

Nuclear Electric Propulsion (NEP) has been considered for crew and cargo class missions to Mars for more than six decades. The Primary Heat Rejection Subsystem (PHRS) of an NEP vehicle expels the megawatts of waste heat generated by the Power Conversion Subsystem (PCS). In previous studies, integrating the majority of the NEP vehicle within a single launch fairing required the use of complex deployable elements and resulted in reduced PCS efficiency. To achieve the necessary radiating area, past design studies have utilized massive multi-wing folding radiators resulting in conceptual PHRS designs that have been widely acknowledged as non-optimal. The Modular Assembled Radiators for NEP VehicLes (MARVL) project is incorporating in-Space Assembly (iSA) into the design process to realize lightweight, modular, and more capable PHRS radiators. Since the project’s Authority to Proceed (ATP) date of October 2024, the project team has been developing a modular design for the PHRS radiators that will be suitable for both the Mars mission environments and iSA commissioning operations. This document captures the current progress of the incremental technology development for the thermal components of the MARVL System of Interest (SOI).

Modular Radiator↗
NASA NTRS2026

A Review of Pump Package Assembly Anomalies and Demonstration to Gain Motor Controller Replacement Capability

The International Space Station (ISS) generates heat within the US modules from systems used to operate the station, maintain crew health, and conduct science experiments in the orbiting laboratory. The Internal Thermal Control System (ITCS) collects this heat in water loops and transfers it to the external ammonia system for rejection to space. The ITCS uses six Pump Package Assemblies (PPA) as the motive force to circulate the water within the Lab, Node 2, and Node 3 (Node 1 is parasitic to the Lab and Node 3 ITCS loops). In the life of ISS, 3 PPAs have had to be replaced and 2 PPAs are currently showing signs of degradation. The first PPA to fail in 2003 was due to early coolant chemistry issues; the remaining pump failures and current pump degradations are determined to be related to the Pump/Fan Motor Controller (PFMC), with their individual signatures all being unique to this point. The PFMC of one of the failed PPAs was returned and testing confirmed its sub-components to be the root cause of the PPA failure. Due to the relative size and weight of the standalone PFMC in comparison to the integrated assembly, it is desired, from a launch and on-orbit stowage perspective, to replace the PFMC component of the degraded PPA instead of the entire PPA. The PPA was not designed for this capability, so on the ground demonstrations are planned using crew tools to show that the PFMC can be replaced on-orbit while maintaining full functionality. This paper will focus on the signatures of the two most recent PPA failures and the two PPAs showing signs of degradation, in addition to presenting the findings from the PFMC demonstration effort.

Aaron Rodriguez↗
NASA NTRS2026

A Review of Pump Package Assembly Anomalies and Demonstration to Gain Motor Controller Replacement Capability

The International Space Station (ISS) generates heat within the US modules from systems used to operate the station, maintain crew health, and conduct science experiments in the orbiting laboratory. The Internal Thermal Control System (ITCS) collects this heat in water loops and transfers it to the external ammonia system for rejection to space. The ITCS uses six Pump Package Assemblies (PPA) as the motive force to circulate the water within the Lab, Node 2, and Node 3 (Node 1 is parasitic to the Lab and Node 3 ITCS loops). In the life of ISS, 3 PPAs have had to be replaced and 2 PPAs are currently showing signs of degradation. The first PPA to fail in 2003 was due to early coolant chemistry issues; the remaining pump failures and current pump degradations are determined to be related to the Pump/Fan Motor Controller (PFMC), with their individual signatures all being unique to this point. The PFMC of one of the failed PPAs was returned and testing confirmed its sub-components to be the root cause of the PPA failure. Due to the relative size and weight of the standalone PFMC in comparison to the integrated assembly, it is desired, from a launch and on-orbit stowage perspective, to replace the PFMC component of the degraded PPA instead of the entire PPA. The PPA was not designed for this capability, so on the ground demonstrations are planned using crew tools to show that the PFMC can be replaced on-orbit while maintaining full functionality. This paper will focus on the signatures of the two most recent PPA failures and the two PPAs showing signs of degradation, in addition to presenting the findings from the PFMC demonstration effort.

Aaron Rodriguez↗
DOE OSTI2020

Refrigeration system

In one aspect, a refrigeration system is provided. The refrigeration system includes a refrigeration circuit configured to condition an air supply, a subcooling circuit configured to cool the refrigeration circuit, the subcooling circuit including a subcooling condenser, a subcooling heat exchanger, and at least one adsorption bed, and a heat generation system thermally coupled to the subcooling circuit.

42 ENGINEERING↗
DOE OSTI2021

Electronic assembly having sectional thermal management

The first board section comprises a switching module with heat-generating semiconductor switches associated with a first operational temperature range. The first board section has a first conductive layer of a first thickness. The second board section comprises a plurality of capacitors mounted on a second circuit board. The second board section has conductive traces for interconnecting the capacitors as a network. The capacitors are associated with a second operational temperature range that is lower than the first operational temperature range. A thermal isolation intermediary forms a barrier between, or adjoining, the first board section and the second board section, where the first board section and the second board section are spaced part from each other by the thermal isolation intermediary.

Schmit, Christopher J.↗
DOE OSTI2021

CTE-tuned pyrolytic graphite (PG) substrate to minimize joining stress between laser diode and the substrate

A pyrolytic graphite (PG) substrate and laser diode package includes a substrate body having a PG crystalline structure with a basal plane oriented at a pre-determined orientation angle as measured from a longitudinal axis of a heat generating material, such as a laser diode, mounted on a surface of the PG substrate, so that a coefficient of thermal expansion (CTE) of the PG substrate is substantially matched with a CTE of the material.

Patra, Susant K.↗
DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

This study focuses on requirements and challenges pertaining to advanced transportable nuclear energy systems, which are currently in development and feature varying degrees of mobility. These reactors mostly use high-assay low-enriched uranium (HALEU) fuel, and only limited refueling is required during their lifecycles. Due to the smaller source terms, higher burnup fuel, and potentially higher thermal efficiency involved, the waste inventory will be less than that generated by conventional light-water reactors. However, the relevant regulatory requirements must be met when transporting HALEU fuel (whether fresh or used) inside the reactor vessel—mainly the fuel material transportation, siting, and physical protection requirements outlined in US Nuclear Regulatory Commission (NRC) regulations 10 CFR 71 and 73. Spent nuclear fuel (SNF), as well as irradiated fuel (used during any interval of reactor operation), contains highly radioactive materials that generate heat and necessitate special precautions during transportation. This study addresses the challenges that this incurs, and outlines potential solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

HIGH-LOW FIDELITY THERMAL HYDRAULIC COUPLING USING AI/MACHINE LEARNING ALGORITHMS

The primary goal of the US Department of Energy (DOE) office of Nuclear Energy Integrated Energy Systems (IES) program is to develop the tools and framework for coupling multi-scale and multi-physical thermal and electrical energy usage and storage systems. High- and low-fidelity (high–low) coupling is a key feature of multi-scale, multi-component systems and has been an important focus of research in the nuclear energy community for the past two decades. An essential feature of demonstrating the capability to couple high-fidelity and low-fidelity systems for real-time applications are surrogate/reduced order models (ROM). For the purposes of this study, surrogate models are essentially Blackbox models, typically developed using supervised Machine learning (ML) algorithms. The surrogate models can be used to mimic the response of high-fidelity models to represent large historical datasets and coupled with more general low-fidelity system models distributed as Functional Mock-up Interface (FMI) or Functional Mock-up Units (FMU) modules. The example is demonstrated with Spallation Neutron Source (SNS) First Target Station flow loop data. The flow loop is a liquid mercury loop with a pump, piping, heat exchange, and internal heat generation in the target window. This work elucidates some of the potential benefits and future needs of developing tools for high–low system coupling of energy systems.

Williams, Wesley↗
DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

This study focuses on requirements and challenges pertaining to advanced transportable nuclear energy systems, which are currently in development and feature varying degrees of mobility. These reactors mostly use high-assay low-enriched uranium (HALEU) fuel, and only limited refueling is required during their lifecycles. Due to the smaller source terms, higher burnup fuel, and potentially higher thermal efficiency involved, the waste inventory will be less than that generated by conventional light-water reactors. However, the relevant regulatory requirements must be met when transporting HALEU fuel (whether fresh or used) inside the reactor vessel—mainly the fuel material transportation, siting, and physical protection requirements outlined in US Nuclear Regulatory Commission (NRC) regulations 10 CFR 71 and 73. Spent nuclear fuel (SNF), as well as irradiated fuel (used during any interval of reactor operation), contains highly radioactive materials that generate heat and necessitate special precautions during transportation. This study addresses the challenges that this incurs, and outlines potential solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Heat sink for luminaire and luminaire arrangements having a heat sink

A heat sink for a luminaire includes a central portion having a top surface and a bottom surface. The bottom surface is adapted to receive a lighting arrangement. The heat sink further includes a plurality of arms configured to dissipate heat generated by the lighting arrangement. The plurality of arms extend radially outward from the central portion. Each one of the plurality of arms is substantially arcuate between a proximal end and a distal end.

Trublowski, John↗
DOE OSTI2023

Multiphysics Analysis of the MSRE Experiment Using Griffin-SAM Coupled Code System

Molten salt reactors (MSRs) with flowing fuel have the unique feature of utilizing the fuel salt for heat generation and extraction at the same time since the fuel salt is circulating through the whole primary loop of the reactor. This movement of the fuel salt results in a partial decay of the delayed neutron precursors (DNPs) outside the core and corresponding redistribution in the active core region. To capture this phenomenon accurately, neutronics and thermal hydraulics computational tools need to be modified to handle the movement of the DNPs and their decay. Idaho and Argonne National Laboratories are actively working on developing neutronics and thermal hydraulics tools to model MSRs with flowing fuel. The neutronics code Griffin and the thermal hydraulics code SAM were extended to handle flowing fuel with the drift of the DNPs under the MOOSE framework. In this work, a simplified Multiphysics model of the Molten Salt Reactor Experiment (MSRE) is developed and utilized to perform steady-state and transient analyses. The MSRE experiments were designed to show the impact of the DNPs losses on core reactivity. The following section provides a description of the MSRE experiment with the core main parameters used to develop the model. Then, the developed Multiphysics model of the MSRE is presented. Finally, steady-state verification tests and unprotected loss of flow transient tests are discussed followed by a summary and conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

CNRS Molten Salt Reactor Benchmark Analysis Using Griffin-Pronghorn Coupled Multi-physics Code System

Molten salt reactors (MSRs) with flowing fuel have the unique feature of utilizing the fuel salt for heat generation and extraction at the same time, since the fuel salt is circulating through the whole primary loop. This movement of fuel salt results in a partial decay of the delayed neutron precursors (DNPs) outside the core and the corresponding redistribution of the DNPs in the active region of the core. To capture this effect, neutronics and thermal-hydraulics (T/H) codes need to be able to handle the movement of DNPs and their decay [1, 2]. Idaho National Laboratory (INL) is actively working on developing the neutronics and T/H codes to model the MSRs with flowing fuel. As one of the efforts, the neutronics code Griffin [3] and the T/H code Pronghorn [4] that are built upon the MOOSE framework [5] have been extended to handle the flowing fuel involving the drift of DNPs, and this capability is being tested and verified with MSR benchmark problems including the CNRS benchmark problem [6]. Additionally, a multi-physics analysis of the molten salt reactor experiment (MSRE) is being conducted to verify and validate the codes against available experimental data [7]. This work presents the analysis of the CNRS benchmark problem using the Griffin-Pronghorn coupled multi-physics code system of INL. Both steady-state and transient problems are analyzed, and the results are compared with those of other participants presented in [6]. The following section provides a description of the benchmark, along with input parameters and observables for each step. Then, the results for each step are presented in comparison with other participants’ results, followed by summary and conclusion.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

CTE-tuned pyrolytic graphite (PG) substrate to minimize joining stress between laser diode and the substrate

A pyrolytic graphite (PG) substrate and laser diode package includes a substrate body having a PG crystalline structure with a basal plane oriented at a pre-determined orientation angle as measured from a longitudinal axis of a heat generating material, such as a laser diode, mounted on a surface of the PG substrate, so that a coefficient of thermal expansion (CTE) of the PG substrate is substantially matched with a CTE of the material.

Patra, Susant↗
DOE OSTI2023

Ultra-compact configurable double-sided manifold micro-channel cold plate

A configurable, double sided, manifold micro-channel cold plate includes a one or more manifold micro-channel cold plate cells. Each cold plate cell includes a manifold body, a manifold insert, a plurality of heat sinks, and a plurality of longitudinal openings configured for fluidic communication with the fluid channels. The manifold body has fluid channels to permit flow of a fluid coolant. The manifold insert has a plurality of manifold insert fluid channels, and is configured for receipt in the one of the modular body recesses. The heat sinks are configured for receipt in the modular body recesses, and include an impingement surface configured for fluidic communication with the manifold insert fluid channels, and a heat transfer surface for thermal communication with one or more heat generating devices.

Zhou, Feng↗
DOE OSTI2020

Key Technical Issues for Greater-Than-Class-C (GTCC) Waste Disposal - 20164

Greater-than-Class C (GTCC) waste is low-level waste (LLW) that exceeds the Class C concentrations tabulated in Title 10, Code of Federal Regulations (CFR), Part 61. Disposal of GTCC waste in the near surface (i.e. upper 30 m of the earth's surface) is prohibited in the United States. Though GTCC waste disposal is generally prohibited, the Commission can approve disposal according to 10 CFR 61.55(a)(2)(iv) on a case-by-case basis. The waste classification tables, developed in the early 1980's, were based on model projections of dose to an inadvertent intruder in agricultural, construction, and discovery exposure scenarios. The assumptions and parameters used were documented in a series of public reports. Though different disposal facility designs and site conditions were considered, the waste classification tables were based on a disposal facility design located in a specific environment. The tables were based on shallow (i.e., top few meters) trench facility designs and did not consider deeper facilities. To determine the suitability of GTCC waste disposal in the near-surface, site-specific analyses must account for differences between GTCC waste and Class A, B, and C LLW (hereafter, traditional LLW). GTCC waste can have concentrations of radionuclides that are much higher than traditional LLW. Because of these higher concentrations, processes that are typically not significant for traditional LLW may be significant with respect to disposal of GTCC waste. These processes include, but are not limited to, heat generation, criticality, and radiolysis. The form of the waste as well as the barriers to release of the waste (e.g. waste package) could be substantially different than they are for traditional LLW. These barriers need to be considered when assessing the impacts of accidents during receipt and placement and in evaluating long-term performance. Some GTCC radioactivity is either embedded in stainless steel or contained in stainless steel barriers. Stainless steel can have very low corrosion rates under a variety of environmental conditions. Finally, and possibly most importantly, GTCC waste would likely need to be disposed deeper than traditional LLW to reduce the probability of disturbance. The waste classification table values of traditional LLW are based on an inadvertent intruder excavating into the waste and bringing some of the material to the land surface. If waste is deeper than approximately 5 m, the excavation scenario becomes very unlikely. Therefore, other intruder scenarios, such as drilling exposure scenarios, need to be evaluated. This paper summarizes key technical issues for the disposal of GTCC waste. Our previous study shows that certain GTCC waste may be suitable for near surface disposal whereas others may not. This study may support ongoing technical analyses assessing potential disposal of GTCC in a near surface disposal facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Review of the Safety Analyses of the Konrad Repository - 20530

The Bundesgesellschaft fuer Endlagerung (BGE) is the license holder of all geological repository sites in Germany since 2017. Besides the construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository, BGE is responsible to carry out the operative tasks of selecting a site for a high-level waste (HLW) repository. The Konrad repository is Germany's first repository for radioactive waste with negligible heat generation, which was fully licensed under atomic law within the scope of a plan approval procedure from the original application in 1982 until the license was granted in 2002 and finally confirmed by the Federal Administrative Court in 2007. The emplacement of LILW was licensed up to a volume of 303,000 m{sup 3} with a total β/γ-activity of 5 x 10{sup 18} Bq and a total α-activity of 1.5 x 10{sup 17} Bq. Currently, the former iron ore mine Konrad is being converted to a repository and is expected to come into operation in 2027. But what if there is no safety case in place nor a regulatory framework tailored for a change of documents in order to ensure the safety according to the state-of-the-art of science and technology during the evolution of a long-term project such as a geological repository? In this presentation, a review procedure of the safety analyses, that have been the basis of the plan approval decision for the Konrad repository at that time, will be described. Furthermore, the first results of this graduated review procedure for updating the safety analyses will be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Borehole Disposal: Update on the Science, Technology, and Potentially Suitable Radioactive Wastes - 20170

A recent Special Issue of the Energies Journal on Deep Borehole Disposal of Nuclear Waste has delivered a timely update on the science and technology of borehole disposal and the types of radioactive wastes it could potentially accommodate. The Special Issue papers discuss (i) circumstances under which a national waste management programme might wish to consider deep borehole disposal [1], (ii) a status report of deep borehole disposal options in Germany [2], (iii) the use of deep horizontal drill-holes in sedimentary, metamorphic or igneous rocks for disposal of spent nuclear fuel and high-level radioactive waste [3], (iv) the analysis of corrosion performance of engineered barrier systems [4], (v) a review of the potential cementing systems suitable for deep borehole disposal [5], (vi) the thermal evolution around heat-generating waste for a wide range of material properties and disposal configurations [6], (vii) a geochemical analysis of deep brines focussed on fluid-rock interactions [7], (viii) post-closure performance assessment calculations for deep borehole disposal of Cs/Sr capsules [8], and (ix) an example safety case for deep borehole disposal of nuclear wastes [9]. This paper provides an overview of the main findings from these publications and highlights some of the technological and scientific challenges that need to be overcome when developing a deep borehole disposal concept. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗