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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 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Certifying the Performance of Fixative Technologies under Open Air Demolition Activities for D and D - 20096

Department of Energy (DOE) facilities undergoing deactivation and decommissioning (D and D) activities struggle with safety concerns of residual radioactive contamination leftover after gross decontamination efforts have concluded. The positive effects of implementing fixative technologies for decommissioning and maintenance is known across the DOE complex. However, there is no standardized metric in place that quantifies the performance of these technologies such as incombustible fixative platforms that immobilize the contamination in a solid polymer material under normal operating conditions or when exposed to stressors. The primary focus of this research is to provide the empirical data necessary to properly characterize the effects of implementing fixative technologies on mitigating the release of contamination. In addition, this effort has been used to identify and establish uniform testing protocols to quantify operational parameters of fixating platforms and properly credit such systems under a variety of operational conditions and stressors that can arise during open air demolition efforts (e.g. impact, water, thermal, etc.). The established protocols will be provided to the ASTM International E10.03 Subcommittee on Radiological Protection for Decontamination and Decommissioning of Nuclear Facilities and Components for the potential of formal standardization. Attributes to be addressed during testing are material compromise from impact stress, thermal stress for incombustible fixative materials, and immobilization factors (how well an immobilizing platform is capable of retaining fixated material during thermal and impact stressors). In developing these test methods, a greater understanding of the material's behavior under anticipated and unanticipated events, such as decommissioning activities and contingency events as outlined in the Basis of Interim Operations (BIO), can be successfully characterized. There are several key components of the experimental methodology that is essential in empirically certifying fixative technologies. A surrogate contaminant with a unique signature has been utilized with the implementation of a uniform contamination process of test coupons that is quantifiable and replicable for direct fixative comparisons. A modular test chamber has also been used to apply the various stressors of interest. Collection of released contamination that includes airborne and resettled contaminant particles and analyzed using mass spectrometry. This research directly supports DOE complex-wide concerns in terms of final disposition of nuclear facilities and the considerations made during their time between ceasing facility operation and final disposition. The results of this effort aims to provide a standardized road map and logical decision flow to better facilitate fixative technologies into D and D activities required to achieve the desired facility end state. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Application of an Embedded Fracture and Borehole Modeling Approach to the Understanding of EGS Collab Experiment 1

EGS Collab is a series of meso-scale experiments and associated numerical simulation activities being funded by the United States Department of Energy, Geothermal Technologies Office (GTO) to investigate enhanced geothermal system processes under in-situ stress and slightly elevated temperature conditions. This project is designed to provide scientists and engineers with immediate access to impermeable rock at scales larger than possible in the laboratory, but generally smaller than those for commercial production. Immediate access to rock is provided via the existing drifts of the former Homestake Gold Mine, now operated as the Sanford Underground Research Facility in Lead, South Dakota. The objectives of the EGS Collab project are to develop well controlled fracture networks between injector and producer boreholes using normal and shear stimulation for permeability enhancement. The first experimental site was located off the West Access drift on the 4850 Level (4850 feet below ground surface) in phyllite of the Precambrian Poorman formation, and involved the creation of a fracture network comprising a combination of hydraulic and natural fractures. A second experimental site is now being considered near the battery alcove on the 4100 Level in amphibolite and rhyolite of the Yates Unit. Data generated during these experiments will be compared against predictions of a suite of computer codes specifically designed to solve problems involving coupled thermal, hydrological, geomechanical, and geochemical processes. Comparisons between experimental and numerical simulation results will provide code developers with direction for improvements and verification of process models, build confidence in the suite of available numerical tools, and ultimately identify critical future development needs for the geothermal modeling community. Moreover, conducting thorough comparisons of models, modelling approaches, measurement approaches and measured data, via the EGS Collab project, will serve to identify techniques that are most likely to succeed at the Frontier Observatory for Research in Geothermal Energy (FORGE), the GTO’s flagship EGS research effort. Experiment 1 has comprised a series of successful tests, including a long-term chilled-water circulation test, but the testbed has two atypical EGS elements. Active ventilation in the adjacent drift over a 50-year period cooled the testbed rock mass and hydraulic fracturing intersected monitoring boreholes making them conduits for fluid flow. Numerical simulations executed in support of the design of the EGS Collab Experiment 1, computed a radial temperature distribution and stress gradient orthogonal to the drift axis, resulting in the forecast of an oblong hydraulic fracture geometry, extended in the direction of the drift from the stimulation borehole. This paper describes the numerical simulation of long-term chilled-water circulation test with an embedded fracture and borehole modeling approach. The principal objective of the simulation work is show agreement between experimental observations in terms of production fluid temperatures, monitoring borehole temperatures, injection and production pressures, and tracer recoveries using a single conceptual model for the fracture network, using characterization and monitoring data generally available for EGS. A secondary objective is to improve the numerical simulation result comparisons with additional data available from the broader monitoring equipment within the testbed versus deeper, hotter, and more remote EGS. One complicating factor for comparisons between the numerical simulation results and experimental observations, is the Joule-Thomson heating associated with large pressure drops across the fracture network.

numerical simulation, embedded fractures and boreh↗
DOE OSTI2014

Influence of local capillary trapping on containment system effectiveness

Immobilization of CO 2 injected into deep subsurface storage reservoirs is a critical component of risk assessment for geologic CO 2 storage (GCS). Local capillary trapping (LCT) is a recently established mode of immobilization that arises when CO 2 migrates due to buoyancy through heterogeneous storage reservoirs. This project sought to assess the amount and extent of LCT expected in storage formations under a range of injection conditions, and to confirm the persistence of LCT if the seal overlying the reservoir were to lose its integrity. Numerical simulation using commercial reservoir simulation software was conducted to assess the influence of injection. Laboratory experiments, modeling and numerical simulation were conducted to assess the effect of compromised seal integrity. Bench-scale (0.6 m by 0.6 m by 0.03 m) experiments with surrogate fluids provided the first empirical confirmation of the key concepts underlying LCT: accumulation of buoyant nonwetting phase at above residual saturations beneath capillary barriers in a variety of structures, which remains immobile under normal capillary pressure gradients. Immobilization of above-residual saturations is a critical distinction between LCT and the more familiar “residual saturation trapping.” To estimate the possible extent of LCT in a storage reservoir an algorithm was developed to identify all potential local traps, given the spatial distribution of capillary entry pressure in the reservoir. The algorithm assumes that the driving force for CO 2 migration can be represented as a single value of “critical capillary entry pressure” P c,entry crit , such that cells with capillary entry pressure greater/less than P c,entry crit act as barriers/potential traps during CO 2 migration. At intermediate values of P c,entry crit , the barrier regions become more laterally extensive in the reservoir, approaching a percolation threshold while non-barrier regions remain numerous. The maximum possible extent of LCT thus occurs at P c,entry crit near this threshold. Testing predictions of this simple algorithm against full-physics simulations of buoyancy-driven CO 2 migration support the concept of critical capillary entry pressure. However, further research is needed to determine whether a single value of critical capillary entry pressure always applies and how that value can be determined a priori. Simulations of injection into high-resolution (cells 0.3 m on a side) 2D and 3D heterogeneous domains show two characteristic behaviors. At small gravity numbers (vertical flow velocity much less than horizontal flow velocity) the CO 2 fills local traps as well as regions that would act as local barriers if CO 2 were moving only due to buoyancy. When injection ceases, the CO 2 migrates vertically to establish large saturations within local traps and residual saturation elsewhere. At large gravity numbers, the CO 2 invades a smaller portion of the perforated interval. Within this smaller swept zone the local barriers are not invaded, but local traps are filled to large saturation during injection and remain during post-injection gravity-driven migration. The small gravity number behavior is expected in the region within 100 m of a vertical injection well at anticipated rates of injection for commercial GCS. Simulations of leakage scenarios (through-going region of large permeability imposed in overlying seal) indicate that LCT persists (i.e. CO 2 remains held in a large fraction of the local iv traps) and the persistence is independent of injection rate during storage. Simulations of leakage for the limiting case of CO 2 migrating vertically from an areally extensive emplacement in the lower portion of a reservoir showed similar strong persistence of LCT. This research has two broad implications for GCS. The first is that LCT can retain a significant fraction of the CO 2 stored in a reservoir – above and beyond the residual saturation -- if the overlying seal were to fail. Thus frameworks for risk assessment should be extended to account for LCT. The second implication is that compared to pressure driven flow in reservoirs, CO 2 migration and trapping behave in a qualitatively different manner in heterogeneous reservoirs when buoyancy is the dominant driving force for flow. Thus simulations of GCS that neglect capillary heterogeneity will fail to capture important features of the CO 2 plume. While commercial reservoir simulation software can account for fine scale capillary heterogeneity, it has not been designed to work efficiently with such domains, and no simulators can handle fine-scale resolution throughout the reservoir. A possible way to upscale the migration and trapping is to apply an “effective residual saturation” to coarse-scale grids. While the extent of overall immobilization can be correlated in this way, all coarser grids failed to capture the distance traveled by the migrating CO 2 for large gravity number. Thus it remains unclear how best to account for LCT in the routine simulation work-flow that will be needed for large-scale GCS. Alternatives meriting investigation include streamline methods, reduced-physics proxies (e.g. particle tracking), and biased invasion percolation algorithms, which are based on precisely the capillary heterogeneity essential for LCT.

58 GEOSCIENCES↗
DOE OSTI2012

Bonneville Powerhouse 2 Fish Guidance Efficiency Studies: CFD Model of the Forebay

In ongoing work, U.S. Army Corps of Engineers, Portland District (CENWP) is seeking to better understand and improve the conditions within the Bonneville Powerhouse 2 (B2) turbine intakes to improve survival of downstream migrant salmonid smolt. In this study, the existing B2 forebay computational fluid dynamics (CFD) model was modified to include a more detailed representation of all B2 turbine intakes. The modified model was validated to existing field-measured forebay ADCP velocities. The initial CFD model scenarios tested a single project operation and the impact of adding the Behavior Guidance System (BGS) or Corner Collector. These structures had impacts on forebay flows. Most notable was that the addition of the BGS and Corner Collector reduced the lateral extent of the recirculation areas on the Washington shore and Cascade Island and reduced the flow velocity parallel to the powerhouse in front of Units 11 and 12. For these same cases, at the turbine intakes across the powerhouse, there was very little difference in the flow volume into the gatewell for the clean forebay, and the forebay with the BGS in place and/or the Corner Collector operating. The largest differences were at Units 11 to 13. The CFD model cases testing the impact of the gatewell slot fillers showed no impact to the forebay flows, but large differences within the gatewells. With the slot fillers, the flow above the standard traveling screen and into the gatewell increased (about 100 cfs at each turbine intake) and the gap flow decreased across the powerhouse for all cases. The increased flow up the gatewell was further enhanced with only half the units operating. The flow into the gatewell slot was increased about 35 cfs for each bay of each intake across the powerhouse; this change was uniform across the powerhouse. The flows in the gatewell of Unit 12, the most impacted unit for the scenarios, was evaluated. In front of the vertical barrier screen, the CFD model with slot fillers showed reduced the maximum velocities (in spite of the increased the flow into the gatewell), and decreased the area of recirculation. The area near the VBS exceeding the normal velocity criteria of 1 ft/s was reduced and the flows were more balanced.

Bonneville↗
DOE OSTI2018

Gas Retention, Gas Release, and Fluidization of Spherical Resorcinol-Formaldehyde (sRF) Ion Exchange Resin

The Low-Activity Waste Pretreatment System (LAWPS) is being developed to provide treated supernatant liquid from the Hanford tank farms directly to the Low-Activity Waste (LAW) Vitrification Facility at the Hanford Tank Waste Treatment and Immobilization Plant. The design and development of the LAWPS is being conducted by Washington River Protection Solutions, LLC. A key process in LAWPS is the removal of radioactive Cs from LAW liquids using ion exchange (IX) columns filled with spherical resorcinol-formaldehyde (sRF) resin. When loaded with radioactive Cs, radiolysis of water in the LAW liquid will generate hydrogen gas. In normal operations, the generated hydrogen is expected to remain dissolved in the liquid and be continuously removed by liquid flow. One accident scenario being evaluated is the loss of liquid flow through the sRF resin bed after it has been loaded with radioactive Cs and hydrogen gas is being generated by radiolysis. For an accident scenario with a loss of flow, hydrogen gas can be retained within the IX column both in the sRF resin bed and below the bottom screen that supports the resin within the column, which creates a hydrogen flammability hazard. Because there is a potential for a large fraction of the retained hydrogen to be released over a short duration as a gas release event, there is a need to quantify the size and rate of potential gas release events. Due to the potential for a large, rapid gas release event, an evaluation of mitigation methods to eliminate the hydrogen hazard is also needed. One method being considered for mitigating the hydrogen hazard during a loss of flow accident is to have a secondary flow system, with two redundant pumps operating in series, that re-circulates liquid upwards through the bed and into a vented break tank where hydrogen gas is released from the liquid and removed by venting the headspace of the break tank. The mechanism for inducing release of gas from the sRF bed is to fluidize the bed, which should allow retained bubbles to rise and be carried to the break tank. Currently, neither the ability of fluidizing a settled sRF bed to release gas nor the quantity of hydrogen gas that can be retained and the corresponding potential for large, rapid gas release events has been tested and evaluated. Accordingly, the purpose of this study was to conduct experiments and develop models that evaluate the ability of a re-circulating flow system to fluidize an sRF resin bed as a method to release retained hydrogen gas and maintain hydrogen gas retention at low levels, thereby avoiding conditions where hydrogen gas poses a flammability hazard. The purpose of the study was also to determine if effective hydrogen gas release would occur using a single constant fluidization velocity, because this is the most simple safety system to implement, for the range of fluid and sRF resin conditions anticipated in the column (because a velocity that is effective for one fluid/resin pair can be ineffective for different resin/fluid pairs). Finally, the study also included quantifying the volume of gas that can be retained in an sRF bed and the behavior of spontaneous, rather than fluidization-induced, gas release events.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

First measurement of symmetric cumulants of hexagonal flow harmonics in Pb-Pb collisions at $\sqrt{s_{NN}}$ = 5.02 TeV

Correlations between event-by-event fluctuations of anisotropic flow harmonics are measured in Pb-Pb collisions at a center-of-mass energy per nucleon pair of 5.02 TeV, as recorded by the ALICE detector at the LHC. This study presents correlations up to the hexagonal flow harmonic 𝑣 6 , which was measured for the first time. The magnitudes of these higher-order correlations are found to vary as a function of collision centrality and harmonic order. These measurements are compared to viscous hydrodynamic model calculations with EKRT initial conditions and to the iEBE-VISHNU model with T R ⁢ENTo initial conditions. The observed discrepancies between the data and the model calculations vary depending on the harmonic combinations. Due to the sensitivity of model parameters estimated with Bayesian analyses to these higher-order observables, the results presented in this work provide new and independent constraints on the initial conditions and transport properties in theoretical models used to describe the system created in heavy-ion collisions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 1

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore fuel product. The long-term EnCore ® fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 2

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore ® fuel product. The long-term EnCore fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15 N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Pollutant Concentrations and Emission Rates from Scripted Natural Gas Cooking Burner Use in Nine Northern California Homes

METHODS: Combustion pollutant concentrations were measured during the scripted operation of natural gas cooking burners in nine homes. In addition to a base condition of closed windows, no forced air unit (FAU) use, and no mechanical exhaust, additional experiments were conducted while operating an FAU and/or vented range hood. Test homes included a 26m2 two-room apartment, a 134m2 first floor flat, and seven detached homes of 117–226m2. There were four single-story, four two-story and one 1.5 story homes. Cooktop use entailed boiling and simmering activities, using water as a heat sink. Oven and broiler use also were simulated. Time-resolved concentrations of carbon dioxide (CO2), nitric oxide (NO), nitrogen oxides (NOX), nitrogen dioxide (NO2), particles with diameters of 6 nm or larger (PN), carbon monoxide (CO), and fine particulate matter (PM2.5) were measured in the kitchen (K) and bedroom area (BR) of each home. CO2, NO, NO2, and PN data from sequential experiments were analyzed to quantify the contribution of burner use to the highest 1h and 4h time-integrated concentrations in each room. RESULTS: Four of the nine homes had kitchen 1h NO2 exceed the national ambient air quality standard (100 ppb). Two other homes had 1h NO2 exceed 50 ppb in the kitchen, and three had 1h NO2 above 50 ppb in the bedroom, suggesting substantial exposures to anyone at home when burners are used for a single substantial event. In all homes, the highest 1h kitchen PN exceeded 2 x105 cm-3-h, and the highest 4h PN exceeded 3 x105 cm-3-hr in all homes. The lowest 1h kitchen/bedroom ratios were 1.3–2.1 for NO in the apartment and two open floor plan homes. The largest K/BR ratios of 1h NO2 were in a two-story 1990s home retrofitted for deep energy savings: ratios in this home were 3.3 to 6.6. Kitchen 1h ratios of NO, NO2 and PN to CO2 were used to calculate fuel normalized emission factors (ng J-1). Range hood use substantially reduced cooking burner pollutant concentrations both in the kitchen and bedroom of several homes. A hood with large capture volume and a measured flow of 108 L/s reduced concentrations 80-95%. IMPLICATIONS: These measurements demonstrate that operation of natural gas cooking burners without venting can cause short-term kitchen concentrations of NO2 to exceed the US outdoor health standard, and can elevate concentrations of NO, NO2, and ultrafine particles throughout the home. Results are generally consistent with a recent simulation study that estimated widespread 1h NO2 exposures exceeding 100 ppb in homes that use gas burners without venting. While operating a venting range hood can greatly reduce pollutant levels from burner use (and presumably from cooking as well), performance varies widely across hoods. Increased awareness of the need to ventilate when cooking would substantially reduce in-home exposure to NO2 and ultrafine particles in California homes. Helping consumers select effective hoods, for example by publishing capture efficiency performance ratings, also would help reduce exposure.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

flowVS: channel-specific variance stabilization in flow cytometry

Abstract Background Comparing phenotypes of heterogeneous cell populations from multiple biological conditions is at the heart of scientific discovery based on flow cytometry (FC). When the biological signal is measured by the average expression of a biomarker, standard statistical methods require that variance be approximately stabilized in populations to be compared. Since the mean and variance of a cell population are often correlated in fluorescence-based FC measurements, a preprocessing step is needed to stabilize the within-population variances. Results We present a variance-stabilization algorithm, called flowVS, that removes the mean-variance correlations from cell populations identified in each fluorescence channel. flowVS transforms each channel from all samples of a data set by the inverse hyperbolic sine (asinh) transformation. For each channel, the parameters of the transformation are optimally selected by Bartlett’s likelihood-ratio test so that the populations attain homogeneous variances. The optimum parameters are then used to transform the corresponding channels in every sample. flowVS is therefore an explicit variance-stabilization method that stabilizes within-population variances in each channel by evaluating the homoskedasticity of clusters with a likelihood-ratio test. With two publicly available datasets, we show that flowVS removes the mean-variance dependence from raw FC data and makes the within-population variance relatively homogeneous. We demonstrate that alternative transformation techniques such as flowTrans, flowScape, logicle, and FCSTrans might not stabilize variance. Besides flow cytometry, flowVS can also be applied to stabilize variance in microarray data. With a publicly available data set we demonstrate that flowVS performs as well as the VSN software, a state-of-the-art approach developed for microarrays. Conclusions The homogeneity of variance in cell populations across FC samples is desirable when extracting features uniformly and comparing cell populations with different levels of marker expressions. The newly developed flowVS algorithm solves the variance-stabilization problem in FC and microarrays by optimally transforming data with the help of Bartlett’s likelihood-ratio test. On two publicly available FC datasets, flowVS stabilizes within-population variances more evenly than the available transformation and normalization techniques. flowVS-based variance stabilization can help in performing comparison and alignment of phenotypically identical cell populations across different samples. flowVS and the datasets used in this paper are publicly available in Bioconductor.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2025

Phenomena Identification and Ranking Table (PIRT) for heat pipes

Heat pipes are advanced passive thermal management devices that utilize phase change and capillary action to achieve efficient heat transfer. However, due to the complexity of the phenomena coupled in heat pipes, including capillary, phase change, turbulence, and compressibility effects, there are high uncertainties in the predictability of their operational regimes and performance. This PIRT exercise, conducted as a collaborative effort involving the Department of Energy (DOE) Microreactor Program (MRP), the Nuclear Regulatory Commission (NRC), and university partners systematically identifies, reviews, and prioritizes critical phenomena affecting the operation of heat pipes based on their importance and knowledge levels. Additional analyses and discussion are provided for phenomena with high importance and low knowledge, such as wick de-wetting, critical heat flux, contact angles, and pressure dynamics. The discussions included the recognizing challenges and proposing future research directions for both modeling and simulation and experimental efforts. Additionally, the report addresses phenomena with medium importance and low knowledge that could impact heat pipe operation during non-normal or transient operation, including frozen startup, laminar to turbulent transition, geyser boiling, wick priming, underfilling conditions, surface roughness of the wick, NCGs trapped in the wick, and the timescales of startup and shutdown. In conclusion, this comprehensive evaluation serves as a valuable resource for guiding future research and development efforts, supporting the successful integration of heat pipes into critical applications such as nuclear reactors, and contributing to the advancement of heat pipe technologies in safety-critical industries.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2012

Modeling the Structural Response from a Propagating High Explosive Using Smooth Particle Hydrodynamics

This report primarily concerns the use of two massively parallel finite element codes originally written and maintained at Lawrence Livermore National Laboratory. ALE3D is an explicit hydrodynamics code commonly employed to simulate wave propagation from high energy scenarios and the resulting interaction with nearby structures. This coupled response ensures that a structure is accurately applied with a blast loading varying both in space and time. Figure 1 illustrates the radial outward propagation of a pressure wave due to a center detonated spherical explosive originating from the lower left. The radial symmetry seen in this scenario is lost when instead a cylindrocal charge is detonated. Figure 2 indicates that a stronger, faster traveling pressure wave occurs in the direction of the normal axis to the cylinder. The ALE3D name is derived because of the use of arbitrary-Lagrange-Eulerian elements in which the mesh is allowed to advect; a process through which the mesh is modified to alleviate tanlging and general mesh distortion often cuased by high energy scenarios. The counterpart to an advecting element is a Lagrange element, whose mesh moves with the material. Ideally all structural components are kept Lagrange as long as possible to preserve accuracy of material variables and minimize advection related errors. Advection leads to mixed zoning, so using structural Lagrange elements also improves the visualization when post processing the results. A simplified representation of the advection process is shown in Figure 3. First the mesh is distorted due to material motion during the Lagrange step. The mesh is then shifted to an idealized and less distorted state to prevent irregular zones caused by the Lagrange motion. Lastly, the state variables are remapped to the elements of the newly constructed mesh. Note that Figure 3 represents a purely Eulerian mesh relaxation because the mesh is relocated back to the pre-Lagrange position. This is the case when the material flows through a still mesh. This is not typically done in an ALE3D analysis, especially if Lagrange elements exist. Deforming Lagrange elements would certainly tangle with a Eulerian mesh eventually. The best method in this case is to have an advecting mesh positioned as some relaxed version of the pre and post Lagrange step; this gives the best opportunity of modeling a high energy event with a combination of Lagrange and ALE elements. Dyne3D is another explicit dynamic analysis code, ParaDyn being the parallel version. ParaDyn is used for predicting the transient response of three dimensional structures using Lagrangian solid mechanics. Large deformation and mesh tangling is often resolved through the use of an element deletion scheme. This is useful to accommodate component failure, but if it is done purely as a means to preserve a useful mesh it can lead to problems because it does not maintain continuity of the material bulk response. Whatever medium exists between structural components is typically not modeled in ParaDyn. Instead, a structure either has a known loading profile applied or given initial conditions. The many included contact algorithms can calculate the loading response of materials if and when they collide. A recent implementation of an SPH module in which failed or deleted material nodes are converted to independent particles is currently being utilized for a variety of spall related problems and high velocity impact scenarios. Figure 4 shows an example of a projectile, given an initial velocity, and how it fails the first plate which generates SPH particles which then interact with and damage the second plate.

36 MATERIALS SCIENCE↗
DOE OSTI2010

The Discrete Equation Method (DEM) for Fully Compressible, Two-Phase Flows in Ducts of Spatially Varying Cross-Section

For the simulation of light water nuclear reactor coolant flows, general two-phase models (valid for all volume fractions) have been generally used which, while allowing for velocity disequilibrium, normally force pressure equilibrium between the phases (see, for example, the numerous models of this type described in H. Städtke, Gasdynamic Aspects of Two-Phase Flow, Wiley-VCH, 2006). These equations are not hyperbolic, their physical wave dynamics are incorrect, and their solution algorithms rely on dubious truncation error induced artificial viscosity to render them numerically well posed over a portion of the computational spectrum. The inherent problems of the traditional approach to multiphase modeling, which begins with an averaged system of (ill-posed) partial differential equations (PDEs) which are then discretized to form a numerical scheme, are avoided by employing a new homogenization method known as the Discrete Equation Method (DEM) (R. Abgrall and R. Saurel, Discrete Equations for Physical and Numerical Compressible Multiphase Mixtures, J. Comp. Phys. 186, 361-396, 2003). This method results in well-posed hyperbolic systems, this property being important for transient flows. This also allows a clear treatment of non-conservative terms (terms involving interfacial variables and volume fraction gradients) permitting the solution of interface problems without conservation errors, this feature being important for the direct numerical simulation of two-phase flows. Unlike conventional methods, the averaged system of PDEs for the mixture are not used, and the DEM method directly obtains a well-posed discrete equation system from the single-phase conservation laws, producing a numerical scheme which accurately computes fluxes for arbitrary number of phases and solves non-conservative products. The method effectively uses a sequence of single phase Riemann problem solutions. Phase interactions are accounted for by Riemann solvers at each interface. Non-conservative terms are correctly approximated. Some of the closure relations missing from the traditional approach are automatically obtained. Lastly, the continuous equation system resulting from the discrete equations can be identified by taking the continuous limit with weak-wave assumptions. In this work, this approach is tested by constructing a DEM model for the flow of two compressible phases in 1-D ducts of spatially varying cross-section with explicit time integration. An analytical equation of state is included for both water vapor and liquid phases, and a realistic interphase mass transfer model is developed based on interphase heat transfer. A robust compliment of boundary conditions are developed and discussed. Though originally conceived as a first step toward implict time integration of the DEM method (to relieve time step size restrictions due to stiffness and to achieve tighter coupling of equations) in multidimensions, this model offers some unique capabilities for incorporation into next generation light water reactor safety analysis codes. We demonstrate, on a converging-diverging two-phase nozzle, that this well-posed, 2-pressure, 2-velocity DEM model can be integrated to a realistic and meaningful steady-state with both phases treated as compressible.

42 ENGINEERING↗
DOE OSTI2014

AGR-2 Irradiation Test Final As-Run Report, Rev 2

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO 2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 10 25 n/m 2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 10 25 n/m 2 (E >0.18 MeV) for UO 2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO 2 -fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10 -6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10 -6 . In the UO 2 capsule (Capsule 3), the R/B values during the first three cycles were below 10 -7 . R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Failure Analysis–Informed Risk Assessment Framework for Geological Carbon Storage Using Numerical Simulation and Machine Learning

Geological carbon storage (GCS) is recognized as a critical technology for achieving large-scale reductions in anthropogenic carbon dioxide (CO 2 ) emissions. Ensuring long-term containment and safety requires robust risk assessment frameworks that account for geological uncertainty and identify potential failure scenarios. Among various indicators, the area of review (AoR) serves as a key metric for evaluating storage performance, regulatory compliance, and monitoring design, as it delineates the spatial extent impacted by pressure buildup and plume migration. However, conventional AoR-based risk assessments typically perturb parameters within narrow uncertainty bounds, potentially overlooking rare but high-impact events arising from extreme geological conditions. In this study, we present a failure analysis–informed risk assessment framework for large-scale GCS projects to improve site prescreening and monitoring design. A suite of 300 numerical simulations was generated using stochastic geological models that vary five key parameters: net-to-gross ratio, anisotropy azimuth, porosity multiplier, permeability multiplier, and vertical-to-horizontal permeability ratio. Among these, 200 realizations represent normal geological uncertainty, while 100 additional cases explore extreme yet plausible conditions for failure-case analysis. The AoR was simulated and computed from pressure and CO 2 saturation fields, where the baseline AoR boundary, representing the extent predicted under typical geological uncertainty, was defined as the union of 200 normal-range simulations, and failure was identified when extreme-range cases exceeded this baseline. Results show that incorporating broader parameter uncertainty produces significantly larger AoR extents, underscoring the potential underestimation of risk under conventional uncertainty ranges. Furthermore, spatial probability maps derived from failure-induced AoR exceedance identify regions requiring enhanced monitoring attention. Various machine learning (ML)–based classifiers were developed to predict failure occurrence from geological parameters, with the random forest model achieving the highest performance (F1-score of 0.986). Consistent findings from correlation coefficient, feature importance, and Sobol sensitivity analyses reveal that low net-to-gross ratios and permeability multipliers are the dominant risk drivers, reflecting reduced reservoir connectivity and limited pressure dissipation. Altogether, these results provide a novel framework for risk-informed site prescreening and monitoring design that explicitly considers rare but high-impact geological scenarios in GCS projects.

25 ENERGY STORAGE↗
DOE OSTI2017

TRACE Analysis of Loss of Flow for the APR1400

The APR1400 is a pressurized water reactor (PWR). The APR1400 nuclear steam supply system consists of a reactor pressure vessel (RPV), which contains the reactor core, and two primary reactor coolant system loops connected in parallel to the RPV. Each coolant loop consists of a hot leg (HL), a u-tube steam generator (SG), two reactor coolant pumps (RCPs) and two cold legs (CLs). A pressurizer is connected to the hot leg on one of the coolant loops. During normal operation, the RCPs circulate high pressure water through the reactor core to remove the heat. The heated water exits the RPV and passes through the hot legs into the steam generator tubes. Heat is transferred through the walls of the steam generator tubes into a lower pressure secondary coolant system. Cooled primary system water from the steam generators is returned through the RCPs and cold legs to the RPV. The RCPs operate at constant speed; the primary coolant system is designed to provide a prescribed rate of coolant flow through the reactor core. The APR1400 SG secondary coolant system pressure is controlled through modulation of valves in the steam lines to maintain a prescribed turbine inlet pressure. The SG secondary coolant inventory is controlled by using a feedwater control system to maintain a prescribed liquid level in the steam generator downcomer. This water inventory level is sufficient to cover the outer surfaces of the steam generator tubes while preventing overfill that would cause liquid to pass from the steam generators through the steam lines into the turbine system. In the current work, the Nuclear Regulatory Commission's reactor systems analysis code, TRACE?2 was used to perform an analysis of a loss of flow anticipated operational occurrence (AOO). The loss of flow is initiated by a loss of off-site power (LOOP), resulting in a trip of the RCPs. The regulatory purpose of the current work is to confirm safety analyses documented in the APR1400 Design Control Document (DCD)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

U.S. Fusion D-Li Neutron Irradiation Facility: Fusion Prototypic Neutron Source (FPNS) Technology Study

A primary engineering challenge for the realization of fusion energy involves the development of high-performance structural and plasma-facing materials with dimensional stability and resistance to neutron degradation of mechanical properties. The harsh fusion environment, including high levels of helium production and other transmutation products along with high operating temperatures, poses an extreme challenge for materials. Identifying materials that can withstand these conditions is a high priority for fusion energy research. Critical to resolving this gap in scientific understanding is developing a neutron source facility with the ability to mimic the extreme conditions in a fusion reactor. Working with the Department of Energy (DOE) Office of Fusion Energy Sciences (FES), Oak Ridge National Laboratory (ORNL) and the Virtual Laboratory for Technology (VLT) organized a workshop to discuss the need for a Fusion Prototypic Neutron Source (FPNS). [1] While there are a variety of uses for FPNS neutrons, the main driver of US interest is expanding the scientific understanding of materials degradation in the deuterium-tritium (D-T) fusion environment. FPNS will help build the scientific foundation required to enable design of a next step fusion device. Using specifications from previous projects as a guide, key parameters were agreed upon for a near term, moderate cost US facility that would be complementary to the more extensive plans of the International Fusion Materials Irradiation Facility (IFMIF). Types of facilities that could potentially meet those criteria were identified. A scoping study for three technologies was initiated to answer three critical questions: 1) Can the concept meet the key parameters, 2) What is the estimated cost, and 3) What is the estimated schedule. One approach selected for the scoping study is a D-Li stripping source similar to the design of the IFMIF but at a lower beam power and atomic displacement rate for cost and schedule reasons. The purpose of this report is to document the results of the D-Li stripping source option study, providing answers to the three questions. The approach to generating a fusion prototypic neutron source using a deuteron beam incident on a flowing lithium target has been studied for more than 40 years. As a result of the research and development recently executed primarily in Europe and Japan, the ability of this approach to meet key parameters (damage parameters, irradiation volume) outlined in the US workshop for a near-term, moderate cost fusion prototypic neutron source facility is not in doubt. The main question that must be answered is the cost of such a facility. As a result, a strong focus was placed on understanding alternatives for developing the facility and their relative costs. The project was broken into four components for the cost estimate: the ion source, accelerator, lithium target system, and conventional facilities. The cost of the accelerator is an important cost driver for the D-Li stripping concept. To fully understand the range of the accelerator cost, multiple alternatives were considered including normal conducting and superconducting linacs and the frequency of the radiofrequency quadrupole. Multiple previous design studies dating from 1989 through 2017 were examined (costs were escalated to 2019 dollars) and compared. After analysis, two options were selected for costing, one superconducting and one room temperature. We found similar project costs for the two accelerator options, on the order of $100M including design and project management but excluding contingency. Including everything except the conventional facilities, the estimate for the neutron source facility is estimated to be $220M without contingency. Additional significant costs for the facility are the conventional costs arising from the need to construct buildings for the accelerator and target facilities, bringing electrical power to the site including a new substation, and removing the heat from the facility using heat exchangers. While ORNL would oversee this work, much of the effort would be executed by contractors. Estimates for this effort was based on similar projects that are currently being executed at ORNL or were executed at ORNL in the recent past. The estimate for conventional facilities for the neutron facility is $173M without contingency, bringing the total project cost to $393M without contingency. With a 50% contingency, the estimated cost is $589M. The estimate range was established using guidance from the Association for the Advancement of Cost Engineering (AACE). Specifically, guidance from recommended practice 17R-97 provides insight on the expected accuracy range of estimates based on the degree of project definition and the position of the project in the stage-gate process or maturity timeline. An AACE Class 5 type estimate is appropriate for concept screening. The expected accuracy range includes a low end of -50% to -20% and a high end of +30% to +100% around the estimated cost. For the D-Li stripping neutron source, a -20% to +100% cost range was applied for a recommended cost range of $471M to $1,179M. Less emphasis was placed on determining the project schedule due to the fact that the project costs were large and may eliminate this option from consideration. The development of the accelerator facility may require on the order of five years without contingency. A more realistic schedule with contingency may be seven years. However, this estimate should be scrutinized in a preconceptual design activity if a more realistic estimate is needed. Finally, one lesson learned in this process is that the requirements for a near-term, moderate cost facility drives the alternative selection toward existing facilities and infrastructure. The D-Li stripping neutron source option requires a green field site due to the lack of existing buildings and infrastructure that can be reused for this effort. As a result, the conventional facilities costs are a significant fraction of the total facility cost. Other neutron source options where buildings, electrical power, and cooling infrastructure already exist may be able to meet the requirements for a near term, moderate cost facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗