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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 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Pipe Overpack Container Fire Testing: Phase I & II

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a new series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016, and described herein, were done in two phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. This report will describe the various tests conducted in phase I and II, present preliminary results from these tests, and discuss implications for the POCs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Pipe Overpack Container Fire Testing: Phase I II & III.

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016 were done in three phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. The goal of the third phase was to see if surrogate aerosol gets released from the PC when the drum lid is off. This report will describe the various tests conducted in phase I, II, and III, present preliminary results from these tests, and discuss implications for the POCs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Pipe Overpack Container Fire Testing: Phase II-A

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible Transuranic (TRU) waste at Department of Energy (DOE) sites. At the request of DOE's Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), SNL started conducting a new series of fire tests in 2015 to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to overpressurize and release its inner content. In 2016, Phase II tests showed that POCs tested in a pool fire failed within 3 minutes of ignition with the POC lid ejecting. These POC lids were fitted with an all-metal (NUCFIL019DS) filter and revealed that this specific filter did not relieve sufficient pressure to prevent lid ejection. For the test phase discussed in this report, Phase II-A, the POCs are exposed to a 30-minute pool fire, with similar configurations to those tested in Phase II, except that the POC lids are fitted with a hybrid metal-polyethylene (UT9424S) filter instead. This report will: describe the various tests conducted in Phase II-A, present results from these tests, and discuss implications for the POCs based on the test results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Pipe Overpack Container Fire Testing: Phase II-A

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible Transuranic (TRU) waste at Department of Energy (DOE) sites. At the request of DOE's Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), SNL started conducting a new series of fire tests in 2015 to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to overpressurize and release its inner content. In 2016, Phase II tests showed that POCs tested in a pool fire failed within 3 minutes of ignition with the POC lid ejecting. These POC lids were fitted with an all-metal (NUCFIL019DS) filter and revealed that this specific filter did not relieve sufficient pressure to prevent lid ejection. For the test phase discussed in this report, Phase II-A, the POCs are exposed to a 30-minute pool fire, with similar configurations to those tested in Phase II, except that the POC lids are fitted with a hybrid metal-polyethylene (UT9424S) filter instead. This report will: describe the various tests conducted in Phase II-A, present results from these tests, and discuss implications for the POCs based on the test results.

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

The effects of a finite beam pipe on resistive wall instabilities

The effect of resistive wall instabilities on beam break-up is investigated for beam pipes with finite and infinite thickness walls. Analytic solutions for the electric field along the beam axis are determined for both finite and infinite thickness walls using complex analysis. The longitudinal and transverse impedances are also determined; variations from the infinite wall case are observed when the wall thickness is of the order of the skin depth of the beam pipe material.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Development of the PROTEUS and ANSYS Coupled System for Simulating Heat Pipe Cooled Micro Reactors

The objective of this project is to develop the coupled system of PROTEUS/ANSYS for simulating a heat pipe cooled micro reactor. As an initial effort, while figuring out controlling ANSYS Mechanical with Workbench, the coupled system of PROTEUS/FLUENT in the CORBA environment were tried to demonstrate the coupled simulation for a transient problem of a heat pipe cooled micro reactor. Finally, the coupled system of PROTEUS/ANSYS was constructed using APDL commands in the CORBA environment, instead of ANSYS Workbench, since we found that ANSYS with the current version of Workbench Python scripting did not provide capabilities necessary for simulating a transient problem.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Pipe Overpack Container Fire Testing: Phase II-B

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire. This test resulted in one of the POCs generating sufficient internal pressure to pop off its drum lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials that would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible Transuranic (TRU) waste at Department of Energy (DOE) sites. At the request of DOE's Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), SNL started conducting a new series of fire tests in 2015 to examine whether PCs with combustibles would reach a temperature that could result in: (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner contents. In 2016, Phase II of the tests showed that POCs tested in a pool fire failed within 3 minutes of ignition with the POC lid ejecting. These POC lids were fitted with a NUCFIL-019DS filter and revealed that this specific filter did not relieve sufficient pressure to prevent lid ejection. In the Fall of 2017, Phase II-A was conducted to expose POCs to a 30-minute pool fire with similar configurations to those tested in Phase II, except that the POC lids were fitted with an UltraTech (UT) 9424S filter instead. That specific filter was chosen because of its design to help relieve internal pressure during the fire and thus prevent lid ejection. In Phase II-A, however, setups of two POCs stacked upon one another were never tested, which led to this phase of tests, Phase II-B. This report will describe the various tests conducted in Phase II-B, present results from these tests, and implications for the POCs based on the test results will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor

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 have expanded its application scope to includ e non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR techno logies under Strategy 2 of the NRC?s near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enha ncements 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 heat pipe reactor (HPR) safety evaluations. The generic plant mode l is based on a publicly-available Los Alamos National Laboratory (LANL) Megapower design as modified in the Idaho National Laboratory (INL) Design A description. For plant aspects (e.g., reactor building size and leak rate) that are not described in the LANL and INL references , the analysts made assumptions needed to construct a MELCOR full-plant model. The HP R uses high assay, low-enrichment uranium (HALEU) fuel with steel cladding that uses heat pipes to transfer heat to a secondary Brayton air cycle. The core region is surrounded by a stainless-steel shroud, alumina reflector, core barrel and boron carbide neutron shield. The reactor is secured inside a below-grade cavity, with the operating floor located above the cavity. 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 re activity control, the primary and secondary system heat removal, and the effectiveness of th e confinement natural circulation flow into the reactor cavity (i.e., a flow blockage).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

SCALE Modeling of the Fast Spectrum Heat Pipe Reactor

As part of the severe accident analysis collaboration with Sandia National Laboratories (SNL) and the Nuclear Regulatory Commission (NRC), SCALE models were developed for a fast-spectrum heat pipe reactor. These models were based on the Idaho National Laboratory (INL) Design A concept, which is an alternative design to the Los Alamos National Laboratory (LANL) Special Purpose Reactor (SPR), also known as the Megapower reactor. The model contains 1,134 heat pipes, surrounded by hexagonal fuel elements, with a potassium working fluid; the fuel is UO 2 with 19.75 wt% 235 U enrichment. The model contains axial beryllium oxide (BeO) reflectors above and below the active fuel region along with a radial alumina reflector containing 12 B 4 C control drums. The center of the core is left unfueled to make room for two shutdown control rods, one annular and one solid. The active region of the core was discretized into twenty axial and five radial zones to analyze spatial variations in power and burnup. Infinite lattice unit cell sensitivity studies were used to perform verification between the SCALE and INL models. The eigenvalue results agreed well with the reported results to within roughly 50 percent mille (pcm). Full-core model verification was performed by analyzing system eigenvalues with differing configurations of control drum and shutdown rod positions. These full core results all had eigenvalue differences less than 310 pcm. Control drum and shutdown rod worths were also compared, with differences of 3.2% or less. Using the verified model, the isotopic inventory and decay heat, as well as temperature feedback coefficients, were calculated and provided to SNL as input to the MELCOR severe accident code to analyze potential releases from this class of reactor. The results of the MELCOR analysis are provided in a different report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

High-Fidelity Neutronics Model of a Realistic Heat Pipe Microreactor Report

The creation of a digital twin requires two parallel products: a virtual model and a physical asset. This report details the development the high-fidelity reactor physics model of a realistic heat pipe micro reactor for the virtual model. The realistic heat pipe micro reactor provides a generic model that research can utilize to make safeguards considerations. The model will be leveraged for capturing diversion and misuse scenarios expected to take place in a microreactor. A python wrapper has been developed to rapidly explore various acquisition pathways by manipulating the reactor physics model on-the-fly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Reference Piping and Instrumentation Diagrams for Heat Transport Systems for Methanol Plants

This report aims to bridge the gap between advanced nuclear reactor vendors and methanol producers seeking decarbonization. It equips both parties with tools and preliminary information for integrating clean heat from advanced reactors into novel methanol production processes. The proposed investigation in this report involves a three-step approach: 1. Design of Current and Decarbonized Process Models for Methanol Production. This report initiates process modeling of two novel methanol production pathways, with their overall energy requirements comparative to classical methanol production (see Figure ES1). The two process models provide the foundation for analyzing integration of hermos-electric generation from advanced nuclear reactors. 2. Identification of Coupling Points to Integrate Advanced Reactors. This report investigates the integration with two advanced reactor technologies: light-water reactors (LWRs) and high-temperature gas reactors (HTGRs). The key coupling points between advanced nuclear reactors and the production facilities associated with the two decarbonized methanol process pathways are identified. 3. Preliminary Heat and Electricity Transfer Design from Reactors to Methanol Production Sites. The preliminary designs for piping and instrumentation to transfer heat from the boundary of advanced reactor technology location to the boundary of methanol production site are evaluated within the context of decarbonized pathways. The pipe analysis for heat transfer adheres to relevant codes and specifications from American Society of Mechanical Engineering. A simplified design for transmission of electricity to the industrial site has been provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

SCALE Analyses of Scenarios in the TRISO-based Heat Pipe Microreactor Fuel Cycle

This report documents the application of the SCALE code to the analysis of a TRistructural-ISOtropic (TRISO)-based heat pipe microreactor (HPMR) within the context of its nuclear fuel cycle stages. The evaluation was conducted in support of the US Nuclear Regulatory Commission’s ongoing efforts to assess modeling capabilities for advanced non–light-water reactor technologies. The generic HPMR selected as a representative microreactor concept features a compact core design that incorporates TRISO fuel compacts, passive heat removal via heat pipes, and a transportable configuration intended for deployment in remote environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Andrieslombaardite, RhSbS, a new platinum-group mineral from the platiniferous Onverwacht Pipe, Republic of South Africa

A hundred years after the discovery of the Merensky Reef in 1924, it is appropriate to present the new mineral andrieslombaardite in honour of Andries Frederik Lombaard who was instrumental in its discovery. Andrieslombaardite, RhSbS, was first described as an unknown mineral from placer deposits associated with the Tulameen Alaskan-Uralian type complex, British Colombia, Canada (Raicevic and Cabri, 1976) but has since been reported from several other deposits including the platiniferous Driekop, Mooihoek, and Onverwacht pipes in the eastern Bushveld Complex, South Africa. The mineral and the name were approved by the Commission on New Minerals Nomenclature and Classification (CNMNC) of the International Mineralogical Association (IMA no. 2022-076) based on data in the co-type samples from Onverwacht and a co-type sample from the Yubdo stream, Birbir River, Ethiopia. Andrieslombaardite in the Onverwacht sample is a single 8 x 20 μm grain attached to laurite in a matrix of altered silicate and Fe-oxyhydroxide minerals. In the Yubdo samples, there are many grains of pale brownish gray andrieslombaardite up to 25 x 55 μm in size, included in Pt-Fe alloys, some associated with erlichmanite, and others attached to bornite and chalcopyrite. The reflectance values (R%) measured in air and in oil at the COM wavelengths are 48.3 and 33.0 (470 nm), 49.3 and 34.0 (546 nm), 51.0 and 35.9 (589 nm), and 51.8 and 36.7 (650 nm). The colour values x, y, Y, λd, and Pe in air are 0.317, 0.322, 50.3, 580, and 3.2, and in oil are 0.319, 0.324, 35.6, 579, and 4.5. The composition of andrieslombaardite is ideally RhSbS, but it contains variable amounts of Fe, Pt, Pd, and Ir that may substitute for Rh. The mineral is cubic with unit-cell dimensions of a = 6.0278(4) Å, V = 219.01(6) Å 3 and Z = 4. It was synthesised at 400 and 550°C using stoichiometric elemental amounts. It is a member of the cobaltite group. The mineralisation of the intrusive dunite pipes was probably introduced at high temperatures, under magmatic conditions. The primary assemblages were to a certain degree overprinted and redistributed by low-temperature hydrothermal fluids. The Pt-Fe alloys from Yubdo containing PGM inclusions such as andrieslombaardite in the Yubdo-Alaskan-type complex were formed at some post-magmatic stage owing to PGE remobilisation during hydrothermal or metamorphic episodes.

58 GEOSCIENCES↗
DOE OSTI2024

Assessment of a Grade 91 Steel Forging and Seamless Pipe Section After 141,000-Hours of Operation in a Superheat Outlet Header

In this work, two unique heats of 9Cr creep strength enhanced ferritic (CSEF) steels extracted from a retired superheat outlet header after 141,000 hours of service were evaluated. These two CSEF steels were a forging manufactured to SA-182 F91 (F91) reducer and a seamless pipe produced to SA-335 P91 (P91) pipe. Their creep deformation and fracture behavior were assessed using a lever arm creep frame integrated with in-situ high-temperature digital image correlation (DIC) system. Critical metallurgical and microstructure factors, including composition, service damage, grain matrix degradation, precipitates, and inclusions were quantitatively characterized to link the performance of the two service aged F91 and P91 CSEF steels. The creep test results show the F91 and P91 steels exhibit a large variation in creep strength and creep ductility. The F91 steel fractured at 572 hours while P91 steel fractured at 1,901 hours when subjected to a test condition of 650 °C and 100 MPa. The nominal creep strains at fracture were 12.5% (F91) and 14.5% (P91), respectively. The high-resolution DIC strain measurements reveal the local creep strain in F91 was about 50% while the local creep strain in P91 was >80%. The characterization results show that the F91 steel possessed pre-existing creep damage from its time in service, a higher fraction of inclusions, and a faster matrix grain coarsening rate. These features contribute to the observed reduction in performance for the F91 steel. The context for these findings, and the importance of metallurgical risk in an integrated life management approach will be emphasized.

Wang, Yiyu↗
DOE OSTI2020

Theoretical Analysis of Fractional Viscoelastic Flow in Circular Pipes: General Solutions

Fractional calculus is a relatively old yet emerging field of mathematics with the widest range of engineering and biomedical applications. Despite being an incredibly powerful tool, it, however, requires promotion in the engineering community. Rheology is undoubtedly one of the fields where fractional calculus has become an integral part of cutting-edge research. There exists extensive literature on the theoretical, experimental, and numerical treatment of various fractional viscoelastic flows in constraint geometries. However, the general theoretical approach that unites several most commonly used models is missing. Here we present exact analytical solutions for fractional viscoelastic flow in a circular pipe. We find velocity profiles and shear stresses for fractional Maxwell, Kelvin–Voigt, Zener, Poynting–Thomson, and Burgers models. The dynamics of these quantities are studied with respect to normalized pipe radius, fractional orders, and elastic moduli ratio. Three different types of behavior are identified: monotonic increase, resonant, and aperiodic oscillations. The models developed are applicable in the widest material range and allow for the alteration of the balance between viscous and elastic properties of the materials.

Gritsenko, Dmitry (ORCID:0000000248181512)↗
DOE OSTI2026

Nondestructive Modular Leak Detection in 3D Printed 316L Stainless Steel Pipes via Laser Powder Bed Fusion

This research investigates the leak detection features of 316L Stainless Steel pipe structures manufactured via Laser Powder Bed Fusion (LPBF). This work involves the design of a modular sensor system integrating nondestructive evaluation (NDE) methods, including thermal imaging and ultrasonic frequency detection to detect and characterize leaks in components. This aims to improve leak detection sensitivity within medium-pressure gas systems, during continuous operation without halting flow or introducing safety risks. The system could be adaptable for use on unmanned aerial vehicles (UAVs), enabling remote leak detection in active environments. A custom pneumatic system incorporating temperature and pressure sensors was assembled to detect leaks in LPBF-printed 316L SS tee pipes. Experimental results and simulations confirm the system’s effectiveness in leak detection and material evaluation. This research program also integrated a Python-based image recognition platform based on a metallography and optical microscopy to assess the porosity and complement the leak detection data on the printed structures. This allows a detailed analysis of pore distribution and internal leak paths, which could compromise structural integrity, critical for quality control during manufacturing. Findings suggest that the investigated approach holds potential for enhancing leak detection technologies and adapt them for advanced manufactured parts.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Importance of Pipe Deposits to Lead and Copper Rule Compliance

When Madison, Wis., exceeded the lead action level in 1992, residential and off-line tests suggested that lead release into the water was more complex than a lead solubility mechanism. Scale analyses (color and texture as well as mineralogical and elemental composition) of five excavated lead service lines (LSLs) revealed that accumulation of manganese (and iron) onto pipe walls had implications for lead corrosion by providing a high-capacity sink for lead. Manganese that accumulated from source well water onto pipe scales (up to 10% by weight of scale composition) served to capture and eventually transport lead to consumer taps. In addition, manganese sometimes obstructed the predominance of an insoluble (and thus potentially protective) plattnerite [Pb(IV) solid] scale layer. Full LSL replacement in Madison achieved Lead and Copper Rule compliance and a major reduction in lead contamination and exposure, supplemented by unidirectional flushing of water mains and manganese control in the source well water.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
NASA NTRS1966

Pipe cutting tool is useful in limited space

Portable pipe cutting tool is used in areas of limited space. The pipe is clamped in the tool and then cut by a rotating cutter assembly that is internally connected to a drive shaft engaged in the chuck of a portable electric drill. The tool is held in a fixed position during the cutting operation.

Headley, C. A.↗