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DOE OSTI2021

Comprehensive Physical Activity Assessment During U.S. Army Basic Combat Training

Abstract Alemany, JA, Pierce, JR, Bornstein, DB, Grier, TL, Jones, BH, and Glover, SH. Comprehensive physical activity assessment during U.S. Army Basic Combat Training. J Strength Cond Res 36(12): 3505–3512, 2022—Physical activity (PA) volume, intensity, and qualitative contextual information regarding activity type and loads carried are limited during U.S. Army Basic Combat Training (BCT). The purpose of this study was to characterize daily (05:00–20:00 hours) PA during BCT using a comprehensive approach. During 2 10-week BCT cycles ( n = 40 trainees per cycle), pedometers, accelerometers, and direct observation were used to estimate daily step count, PA volume, and intensity. Physical activity intensity was categorized by metabolic equivalents (METs) such as “sedentary” (1–2 METs), “light” (2–3 METs), “moderate” (3–6 METs), or “vigorous” (≥6 METs). Daily PA data were analyzed longitudinally using linear mixed models, with significance set at p ≤ 0.05. The mean daily step count was 13,459 ± 4,376 steps, and the mean daily accelerometer-assessed PA volume and intensity were as follows: sedentary: 505 ± 98 minutes, light: 190 ± 78 minutes, moderate: 168 ± 51 minutes, and vigorous: 14 ± 14 minutes, with no differences between cycles for all measures ( p > 0.50). Cumulative time on feet (∼50%) and sitting (20–25%) accounted for most daily activity types during both cycles. Trainees, on average, carried between 3 and 9 kg, and ≥9 kg, for 60% and 10% of the monitored day, respectively. Basic Combat Training's physical demands are high, where trainees achieved 1.7 to 2.7 times greater daily ambulation and 6 times the recommended weekly moderate-to-vigorous PA compared with civilian counterparts and performed weight-bearing load carriage for nearly half of the day. Basic Combat Training-associated PA may increase injury risk among trainees unaccustomed to arduous PA and exercise. Implementing national PA policies to improve physical fitness and facilitate acclimatization to BCT's high physical demands could reduce public health burdens and military nonreadiness.

Sport Sciences↗
DOE OSTI2012

Use of Treated Municipal Wastewater as Power Plant Cooling System Makeup Water: Tertiary Treatment versus Expanded Chemical Regimen for Recirculating Water Quality Management

Treated municipal wastewater is a common, widely available alternative source of cooling water for thermoelectric power plants across the U.S. However, the biodegradable organic matter, ammonia-nitrogen, carbonate and phosphates in the treated wastewater pose challenges with respect to enhanced biofouling, corrosion, and scaling, respectively. The overall objective of this study was to evaluate the benefits and life cycle costs of implementing tertiary treatment of secondary treated municipal wastewater prior to use in recirculating cooling systems. The study comprised bench- and pilot-scale experimental studies with three different tertiary treated municipal wastewaters, and life cycle costing and environmental analyses of various tertiary treatment schemes. Sustainability factors and metrics for reuse of treated wastewater in power plant cooling systems were also evaluated. The three tertiary treated wastewaters studied were: secondary treated municipal wastewater subjected to acid addition for pH control (MWW_pH); secondary treated municipal wastewater subjected to nitrification and sand filtration (MWW_NF); and secondary treated municipal wastewater subjected nitrification, sand filtration, and GAC adsorption (MWW_NFG). Tertiary treatment was determined to be essential to achieve appropriate corrosion, scaling, and biofouling control for use of secondary treated municipal wastewater in power plant cooling systems. The ability to control scaling, in particular, was found to be significantly enhanced with tertiary treated wastewater compared to secondary treated wastewater. MWW_pH treated water (adjustment to pH 7.8) was effective in reducing scale formation, but increased corrosion and the amount of biocide required to achieve appropriate biofouling control. Corrosion could be adequately controlled with tolytriazole addition (4-5 ppm TTA), however, which was the case for all of the tertiary treated waters. For MWW_NF treated water, the removal of ammonia by nitrification helped to reduce the corrosivity and biocide demand. Also, the lower pH and alkalinity resulting from nitrification reduced the scaling to an acceptable level, without the addition of anti-scalant chemicals. Additional GAC adsorption treatment, MWW_NFG, yielded no net benefit. Removal of organic matter resulted in pitting corrosion in copper and cupronickel alloys. Negligible improvement was observed in scaling control and biofouling control. For all of the tertiary treatments, biofouling control was achievable, and most effectively with pre-formed monochloramine (2-3 ppm) in comparison with NaOCl and ClO2. Life cycle cost (LCC) analyses were performed for the tertiary treatment systems studied experimentally and for several other treatment options. A public domain conceptual costing tool (LC3 model) was developed for this purpose. MWW_SF (lime softening and sand filtration) and MWW_NF were the most cost-effective treatment options among the tertiary treatment alternatives considered because of the higher effluent quality with moderate infrastructure costs and the relatively low doses of conditioning chemicals required. Life cycle inventory (LCI) analysis along with integration of external costs of emissions with direct costs was performed to evaluate relative emissions to the environment and external costs associated with construction and operation of tertiary treatment alternatives. Integrated LCI and LCC analysis indicated that three-tiered treatment alternatives such as MWW_NSF and MWW_NFG, with regular chemical addition for treatment and conditioning and/or regeneration, tend to increase the impact costs and in turn the overall costs of tertiary treatment. River water supply and MWW_F alternatives with a single step of tertiary treatment were associated with lower impact costs, but the contribution of impact costs to overall annual costs was higher than all other treatment alternatives. MWW_NF and MWW_SF alternatives exhibited moderate external impact costs with moderate infrastructure and chemical conditioner dosing, which makes them (especially MWW_NF) better treatment alternatives from the environmental sustainability perspective since they exhibited minimal contribution to environmental damage from emissions.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2014

Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and affects mechanical properties of aluminum including density, neutron irradiation hardening, swelling, and loss of ductility. Because slightly greater quantities of silicon will be produced in the cold source moderator vessel for the LEU core, these effects will be slightly greater for the LEU core than for the HEU core. Three-group (thermal, epithermal, and fast) neutron flux results tallied in the cold source LH2 hemisphere show greater values for the LEU core under both BOC and EOC conditions. The thermal neutron flux in the LH2 hemisphere for the LEU core is about 12.4% greater at BOC and 2.7% greater at EOC than for the HEU core. Therefore, cold neutron scattering will not be adversely affected and the 4 12 neutrons conveyed to the cold neutron guide hall for research applications will be enhanced.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

High-performance sorbents for carbon dioxide capture from air

This project has focused on capture of CO 2 from ambient air (“air capture”). If this process is technically and economically feasible, it could potentially contribute to net reduction of CO 2 emissions in ways that are complementary to better developed techniques for CO 2 from concentrated point sources. We focused on cyclic adsorption processes for CO 2 capture from air in which the entire cycle is performed at moderate temperatures. The project involved both experimental studies of sorbent materials and process level modeling of cyclic air capture processes. In our experimental work, a series of amine-functionalized silica adsorbents were prepared and characterized to determine the impact of molecular architecture on CO 2 capture. Some key findings were: • Amine functionalized silicas can be prepared with high enough CO 2 capacities under ambient conditions to merit consideration for use in air capture processes. • Primary amines are better candidates for CO 2 capture than secondary or tertiary amines, both in terms of amine efficiency for CO 2 adsorption and enhanced water affinity. • Mechanistic understanding of degradation of these materials can enable control of molecular architecture to significantly improve material stability. Our process modeling work provided the first publically available cost and energy estimates for cyclic adsorption processes for air capture of CO 2 . Some key findings were: • Cycles based on diurnal ambient heating and cooling cannot yield useful purities or amounts of captured CO 2 . • Cycles based on steam desorption at 110 oC can yield CO 2 purities of ~88%. • The energy requirements for cycles using steam desorption are dominated by needs for thermal input, which results in lower costs than energy input in the form of electricity. Cyclic processes with operational costs of less than $\$$100 tCO 2 -net were described, and these results point to process and material improvements that could substantially reduce these costs. The most critical conclusions from our work are that (i) CO 2 capture from ambient air using moderate temperature cyclic adsorption processes is technically feasible and (ii) the operational costs of realistic versions of these processes are moderate enough to encourage future development of this technology. Because of the very modest net investment that has been made in R&D associated with this approach from all sources worldwide (relative to the massive public and private investment that has been made in technologies for CO 2 from concentrated point sources), our results strongly suggest that continued development of air capture is justified.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Final Report: DE- FC36-05GO15063, Fundamental Studies of Advanced High-Capacity, Reversible Metal Hydrides

The project was component of the US DOE, Metal Hydride Center of Excellence (MHCoE). The Sandia National Laboratory led center was established to conduct highly collaborative and multi-disciplinary applied R&D to develop new reversible hydrogen storage materials that meet or exceed DOE/FreedomCAR 2010 and 2015 system targets for hydrogen storage materials. Our approach entailed a wide variety of activities ranging from synthesis, characterization, and evaluation of new candidate hydrogen storage materials; screening of catalysts for high capacity materials requiring kinetics enhancement; development of low temperature methods for nano-confinement of hydrides and determining its effects on the kinetics and thermodynamics of hydrides; and development of novel processes for the direct re-hydrogenation of materials. These efforts have resulted in several advancements the development of hydrogen storage materials. We have greatly extended the fundamental knowledge about the highly promising hydrogen storage carrier, alane (AlH 3 ), by carrying out the first crystal structure determinations and the first determination of the heats of dehydrogenation of β–AlH 3 and γ-AlD 3 . A low-temperature homogenous organometallic approach to incorporation of Al and Mg based hydrides into carbon aerogels has been developed that that allows high loadings without degradation of the nano-porous scaffold. Nano-confinement was found to significantly improve the dehydrogenation kinetics but not effect the enthalpy of dehydrogenation. We conceived, characterized, and synthesized a novel class of potential hydrogen storage materials, bimetallic borohydrides. These novel compounds were found to have many favorable properties including release of significant amounts of hydrogen at moderate temperatures (75-190 º C). However, in situ IR studies in tandem with thermal gravimetric analysis have shown that about 0.5 equivalents of diborane are released during the dehydrogenation making re-hydrogenation effectively impossible and precluding these compounds from further consideration as hydrogen storage materials. The hydrogen cycling of >11 wt % between MgB 2 to Mg(BH 4 ) 2 was achieved but required very forcing conditions. Under moderate conditions (dehydrogenation 200°C; re-hydrogenation 250°C, 120 atm H 2 ), Mg(BH 4 ) 2 undergoes reversible dehydrogenation to Mg(B 3 H 8 ) 2 . Although the 2.5 wt% cycling capacity does not meet current on-board storage targets, this result provides first example of direct hydrogen cycling of a borohydride under moderate conditions and demonstrates the plausibility of finding mild, PEM fuel cell relevant conditions for the high capacity, reversible dehydrogenation of borohydrides. A method was developed for the room temperature, direct hydrogenation of Ti-doped LiH/Al in liquefied dimethyl ether under 100 atm of H 2 . The process has been optimized such that Ti-doped LiAlH 4 is obtained in >95% yield. The WTT energy efficiency our direct synthesis process has been estimated to approach the 60% U.S. DOE target. Thus our simplification of the hydrogenation half-cycle may provide the key to harnessing the long-recognized potential of this lightweight, high capacity material as a practical hydrogen carrier. Finally, we have gained insight into the fundamental basis of the enhanced hydrogen cycling kinetics of Ti-doped NaAlH 4 through studies by solid state 1 H NMR, anelastic spectroscopy; muon spin rotation; and positron annihilation.

08 HYDROGEN↗
DOE OSTI2016

Simulations of a PSD Plastic Neutron Collar for Assaying Fresh Fuel

The potential performance of a notional active coincidence collar for assaying uranium fuel based on segmented detectors constructed from the new PSD plastic fast organic scintillator with pulse shape discrimination capability was investigated in simulation. Like the International Atomic Energy Agency's present Uranium Neutron Collar for LEU (UNCL), the PSD plastic collar would also function by stimulating fission in the 235 U content of the fuel with a moderated 241 Am/Li neutron source and detecting instances of induced fission via neutron coincidence counting. In contrast to the moderated detectors of the UNCL, the fast time scale of detection in the scintillator eliminates statistical errors due to accidental coincidences that limit the performance of the UNCL. However, the potential to detect a single neutron multiple times historically has been one of the properties of organic scintillator detectors that has prevented their adoption for international safeguards applications. Consequently, as part of the analysis of simulated data, a method was developed by which true neutron-neutron coincidences can be distinguished from inter-detector scatter that takes advantage of the position and timing resolution of segmented detectors. Then, the performance of the notional simulated coincidence collar was evaluated for assaying a variety of fresh fuels, including some containing burnable poisons and partial defects. In these simulations, particular attention was paid to the analysis of fast mode measurements. In fast mode, a Cd liner is placed inside the collar to shield the fuel from the interrogating source and detector moderators, thereby eliminating the thermal neutron flux that is most sensitive to the presence of burnable poisons that are ubiquitous in modern nuclear fuels. The simulations indicate that the predicted precision of fast mode measurements is similar to what can be achieved by the present UNCL in thermal mode. For example, the statistical accuracy of a ten-minute measurement of fission coincidences collected in fast mode will be approximately 1% for most fuels of interest, yielding a ~1.4% error after subtraction of a five minute measurement of the spontaneous fissions from 238 U in the fuel, a ~2% error in analyzed linear density after accounting for the slope of the calibration curve, and a ~2.9% total error after addition of an assumed systematic error of 2%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Photonuclear Contributions to SNS Pulse Shapes

Short-pulsed sources like the Spallation Neutron Source (SNS) and ISIS produce bursts of neutron pulses at rates of 10-60 Hz, with sub-microsecond proton pulses impacting on high-Z target materials. Moderators are grouped around the target to receive the fast neutrons generated from spallation reactions to moderate them effciently to thermal and sub-thermal energies and to feed narrow neutron pulses to neutron scattering instruments. The scattering instruments use the neutrons as a probe for material investigations, and make use of time-of-flight (TOF) methods for resolving the neutron energy. The energy resolution of scattering instruments depends on the narrow time-structure of the neutron pulses, while neutrons in the long tail of the emission time distributions can degrade the instrument performance and add undesired background to measurements. The SNS neutronics team is investigating a possible source term impacting the background at short-pulsed spallation sources. The ISIS TS2 project claims to have significantly reduced neutron scattering instrument background levels by the elimination or reduction of iron shielding in the target-moderator-reflector assembly. An alternative hypothesis, also proposed by ISIS, suggests that this apparent reduction arises from moving beamline shielding away from the neutron guide channels, reducing albedo down the beamlines. In both hypotheses, the background neutrons in question are believed to be generated by photonuclear reactions. If the background neutrons are indeed generated via photonuclear channels, then they are generated in a time-dependent fashion, since most of the high-energy photons capable of inducing photonuclear production are gone within a few microseconds following the proton pulse. To evaluate this e ect, we have enabled photonuclear reactions in a series of studies for the SNS first target station (FTS) taking advantage of its Monte Carlo model. Using a mixture of ENDF/B VII.0 and TENDL-2014 photonuclear cross sections available and the CEM03 physics model within MCNPX 2.6.0 in the simulation, we are able to estimate the impact of photoneutron production on both overall neutron production and delayed neutron production. We find that a significant number of photon-induced neutrons are produced a few milliseconds after the proton pulse, following prompt gamma emission through the capture of neutrons in the slowing-down and thermalization processes. We name these "slowing-down delayed neutrons" to distinguish them from either "activation-delayed neutrons" or "beta-delayed neutrons." The beta-delayed and activation-delayed neutrons were not part of this study, and will be addressed elsewhere. While these other delayed neutron channels result in the time-independent (constant) production of fast neutrons outside of the prompt pulse, the slowing-down delayed neutrons also a ect the shape of the pulses. Although numerically insignificant in most cases, we describe a set of scenarios related to T0-chopper operation in which the slowing-down delayed neutrons may be important.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2017

Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

The High Temperature Test Reactor (HTTR) is a graphite moderated thermal reactor with a nominal power output of 30 MWth. The purpose of this work is to prepare a set of neutron cross sections for use in the transient analyses of three Loss of Forced Cooling (LOFC) events planned for this reactor. The cross sections are generated with the Serpent 2 3 D Monte Carlo code to include the axial heterogeneity and the strong axial coupling of this core. The approach yields good agreement of results when compared to previously used 2 D lattice models, which require a fine spatial discretization and, surprisingly, comes at a higher computational cost. In addition, the temperature distribution must be considered to obtain reasonable cross sections and the axial leakage component dominates the spectral effects in the core. Consequently, a temperature model as a function of the fuel and moderator temperatures is adopted, and the core is depleted to 390 effective full-power days (EFPDs). A full tabulation of cross sections is prepared at this burnup point for a variety of fuel and moderator temperatures. Next, the multiphysics reactor application MAMMOTH is used to evaluate the quality of these cross sections using a diffusion solver. As expected, the homogenization error in the cross sections is significant and the Super Homogenization (SPH) correction from MAMMOTH is necessary to preserve key reaction rates. The SPH corrected MAMMOTH results are in excellent agreement with the Serpent results and reproduce both the reference power profile and temperature coefficients at each tabulation point. This fact confirms the accuracy of the SPH corrected cross sections and demonstrates the maturity of using Serpent 2 and MAMMOTH for 3 D cross section generation, even for reactors as complex as the HTTR. Finally, the authors recommend that the cross sections should be parametrized with the local burnup and that a full core burnup calculation with coupled thermal fluids would provide a better estimate of the initial condition for any subsequent transient analysis. Unfortunately, this was not within the scope of this work, and direct analysis with these tabulations should be considered as a first order approximation, at best.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of Critical Experiments to Provide Validation Data for Multiphysics Coupling Methods (Final Report)

Computational simulation has become one of the most important tools in the modern toolbox for nuclear reactor design. Reactor simulations, like the reactors themselves, involve multiple physical processes including neutronics, fluid flow and heat transfer (thermal-hydraulics), thermo-mechanics, fuel behavior, chemistry, and balance of plant. Each process can affect the behavior of the others, and advanced computational tools have been developed that couple separate physics models together to simulate the feedback effects. The objective of this work is to generate experimental data to support the validation of multi-physics reactor codes. It specifically targets the SHARP toolset from the NEAMS Reactor Product Line, but it is intended to be useful in any set of coded that couples thermal-hydraulics with neutronics simulations. The project takes advantage of the inherent flexibility of the RPI Reactor Critical Facility (RCF), a low power open-pool research reactor. The standard configuration of the RCF is used to explore the effects of moderator temperature on neutronic behavior and reactivity. Because the low power of the reactor generates no appreciable temperature change in the moderator, the moderator temperature is artificially changed with two 18 kW electric immersion heaters. The temperature feedback effects can be observed by recording the change in the reactor power as the reactor goes from a supercritical state, to critical, to a subcritical state. To expand the temperature range and temperature rate of change that can be investigated, an experimental apparatus consisting of a heated water loop was designed and installed in the RCF. In this setup, water from a heated reservoir tank is pumped through a pipe placed through the center of the reactor core. The water in the reservoir can be heated to around 70 °C and pumped through the system. The pump has a variable flow rate and electronic control, so a number of scenarios can be developed. A slug of water can be pushed through the system to generate a reasonably fast change in reactivity. The pump speed can be increased or decreased over the course of a measurement, or even oscillated at different periods, to observe the response of the reactor. To complement the experimental data generated by the RCF, a set of computational models have been developed. They are intended to illustrate the usefulness of the experimental data generated in validating coupled multi-physics simulations. A model previously constructed in MCNP has been used in conjunction with safety analysis at the RCF for many years and was modified to support the design of the heated water loop experiment. This model was translated into Serpent 2, a similar Monte Carlo code, which provided the added advantage of being able to produce collapsed cross-section data files for use in the deterministic neutron transport codes that are part of the SHARP toolset. A complete reconstruction of the RCF model, both with and without the heated fluid loop experiment, were necessary for use with the SHARP codes, PROTEUS (neutronics) and Nek5000 (thermal hydraulics), as these require an unstructured mesh format to operate. The computational requirements for fully coupled three dimensional simulations are significant, and challenges to full implementation and comparison with RCF experiments remain. A wide variety of experiments have been conducted for eventual benchmarking and validation of multi-physics reactor codes. A complete database of detailed time- and temperature-dependent measurements have been generated and stored. Several examples of the experimental results are presented here, with refinement and analysis of experiments ongoing. While these experiments still only address a finite range of operating conditions, in particular being limited to water in the sub-boiling regime, they represent a unique (to date) look at the feedback between temperature and neutronic behavior suited for validation of the coupling routines in multi-physics codes. It is hoped that these will be the foundation of future benchmarks that will provide confidence in the tools that will be used to develop the next generation of nuclear reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Nuclear Data and Benchmarking Program: Nuclear Data Performance Assessment for Advanced Reactors

Over the last several Evaluated Nuclear Data File (ENDF)/B releases, many updates have been introduced to the nuclear data libraries that can significantly change computational results. These changes can be particularly important for advanced reactor concepts, which are not as thoroughly investigated as those used in light water reactor (LWR) systems. Therefore, a performance assessment of the ENDF/B libraries using the SCALE 6.2.3 code package was conducted under the auspices of the U.S., Department of Energy Office of Nuclear Energy (DOE-NE). The following systems relevant to the advanced reactor community were chosen as subject for investigation in this report: sodium cooled fast reactor systems, graphite moderated high temperature gas-cooled reactors, and several molten salt reactor models. At first, the similarity of the integral performance of the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries caused by compensating errors in important differential data was demonstrated. In calculations for a shipping container of high-assay low enriched uranium, the data of individual isotopes were systematically swapped between the two ENDF/B libraries. An eigenvalue difference of up to 450 pcm was found due to the use of 235 U and 238 U from one library and 1 H and 16 O from the other library. This clearly demonstrates a cross-correlation between reaction data sets of different isotopes within a library that should be reported in the evaluations. For sodium-cooled fast reactor (SFR) fuel assemblies, significant eigenvalue (k inf ) differences (200– 450 pcm) were found between calculations using the 2011 ENDF/B-VII.1 data and 2018 ENDF/B-VIII.0 data. These differences were mainly caused by updates of 238 U and 239 Pu neutron cross sections. Results from the multigroup (MG) calculations further revealed that the group structure of the applied MG library strongly influences the MG bias due to the importance of an appropriate energy resolution of the resonances in higher energy ranges. The application of previously used MG libraries requires new verification for advanced reactor simulation in comparisons with reference continuous-energy calculations. Nuclear data uncertainty analyses of these SFR systems resulted in eigenvalue uncertainties between 1,400 and 1,800 pcm, which are three to four times higher than corresponding uncertainties in light water reactor (LWR) systems. The main contributor to this uncertainty was found to be inelastic scattering on 238 U, which shows an uncertainty of up to 50% in the fast energy range. Other important contributors are the scattering reactions of 56 Fe and 23 Na which have so far not appeared in LWR analysis. Since the largest contribution to the eigenvalue uncertainty of SFR systems is coming from scattering reactions, it is expected that uncertainties in the angular scattering distributions also have a significant impact. Those uncertainties are, however, not available for the majority of nuclides, and the capabilities to determine sensitivities to this data are currently underdeveloped. For two graphite moderated high temperature gas-cooled reactor (HTGR) benchmarks, only the ENDF/BVII. 1 calculations resulted in consistent eigenvalues (k eff ) with the corresponding experimental measurements. Eigenvalue differences of about 1,000 pcm were found between the 2006 ENDF/B-VII.0 data and the 2011 ENDF/B-VII.1 data, primarily due to an updated carbon capture cross section in the thermal energy range in the later ENDF/B release. The ENDF/B-VIII.0 eigenvalue was larger than the ENDF/B-VII.1 eigenvalue by about 300 pcm mainly due to updates in multiple 235 U neutron cross sections, with offsetting updates in the 238 U cross sections. With ENDF/B-VIII.0, graphite can be modeled as perfect crystal or with two different porosities. The choice of the graphite evaluation can have a significant influence of the eigenvalue. For a HTGR benchmark, maximum eigenvalue differences as high as 650 pcm due to different porosities were observed. More detailed studies of the impact of the porosity are necessary while considering that the graphite porosity can vary between the used materials and changes as a function of neutron fluence. xii The uncertainty of the HTGR eigenvalues due to nuclear data uncertainties was found between 500 and 600 pcm, with the top contributor being the neutron multiplicity of 235 U. A gap in the form of missing uncertainties in graphite thermal scattering data, that might have significant impact on HTGR reactors, was identified. This gap is being investigated by the DOE-NE Nuclear Data and Benchmarking Program, as documented in a separate report The fast spectrum molten salt reactor calculations revealed similar comparisons as for the SFR assemblies. The eigenvalue uncertainties were significantly influenced by inelastic scattering on 238 U, while the impact of angular scattering distributions is unknown. For the graphite and zirconium hydride moderated system, eigenvalue uncertainties of up to 700 pcm were observed, requiring in depth analyses of the cross sections and corresponding uncertainties of the included materials. In general, it was observed that the flux spectra of the various molten salt reactor systems show significant differences, even between a fresh state and a depleted state. The choice of the energy group structure of MG calculations is therefore highly relevant; the applicability of previously used MG libraries needs to be verified. Due the unavailability of pin power measurements of advanced reactor systems, pin power calculations with both ENDF/B-VII.1 and VIII.0 data were performed for a light water reactor and compared to corresponding measurements. Both calculations show good agreement with the measurements. A comparison between the ENDF/B-VII.1 and VIII.0 results revealed small differences, mostly in the range of about 0.2%.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor

The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures. Both types of HTGR have the tristructural isotropic (TRISO) fuel kernel at the heart of the fuel design. For the prismatic block-type HTGR, the TRISO particles are overcoated with a resinated graphitic matrix and pressed into fuel compacts, which are then heat treated and placed in the fuel channels of the prismatic-block-shaped fuel assemblies. For the pebble-bed-type HTGR, the TRISO particles are dispersed in a graphitic-matrix sphere, which is the basic unit for the reactor core. Despite having very different fuel designs, both types of HTGR are graphite-moderated, gas-cooled, thermal reactors using many of the same materials. As a result, both prismatic-block-type and pebble-bed-type HTGRs have similar radioactive waste streams, all of which require safe and secure storage and eventual disposition. Modern HTGR designs are based on a long and rich operating history of several different graphite-moderated, gas-cooled, thermal reactors. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the spent nuclear fuel (SNF) and radioactive wastes associated with operating these reactors. This white paper will, (1) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste; and (2) identify the key radioactive waste streams from HTGRs and their characteristics. Idaho National Laboratory (INL) has significant experience in the management of SNF from HTGR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new HTGR-type small modular reactor, or microreactor intended for deployment at the INL site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Thermomechanical Processing of Uranium Alloys with 10 and 50 Weight Percent Zirconium

This work focuses on casting and thermomechanical processing for moderate sized (500-1000g) 10% by weight and 50% by weight uranium-zirconium alloys. Three castings of uranium – 10wt% zirconium and a fourth casting of uranium – 50wt% zirconium were completed using different castings processes. The first casting process relied on only induction mixing within the VIM system thereby eliminating the need for an arc-melting feedstock preparation step. The second casting process used the arc-melting process prior for feedstock to the same VIM system. In this study, the induction mixing process alone with a graphite crucible appears insufficient for the fabrication of UZr alloys. Graphite is suitable with the use of master alloy UZr feedstock produced by arc-melting. There was a moderate increase of approximately 100 ppm of carbon within the U10Zr alloys. However, there was a near 1000 ppm increase of carbon in the U50Zr alloy. This carbon content is considered too high for forming processes, and therefore, warrants the use of a ceramic crucible for UZr castings near 50wt% zirconium. Zirconia does not appear suitable for UZr castings near 50wt% zirconium. The use of Beryllia or Thoria may be a more suitable option. However, due to the significantly lower liquidus and lower zirconium content, Zirconia crucibles do appear viable for carbon reduction in the U10Zr system. Two thermomechanical processes are also studied in this work - hot extrusion of the U50Zr system and heat treatment in an argon atmosphere. This work demonstrated that extrusion of the U50Zr alloy near 680°C is a viable choice even with +1000 ppm carbon. Extrusion of the U50Zr system near 630°C was unsuccessful and does not appear viable unless significant force and tooling strength is available. There appears to be a moderate increase in extrudate hardness when compared to as-cast. Heat-treatment reduced the hardness of as-cast alloys in the U10Zr castings as well as resolve some variance in the microstructures. However, for the U50Zr billet that did not successfully extrude at 630°C, heat treatment at 900°C for 96 hours did not have a significant impact on hardness or microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

NEAMS Milestone Report: M2MS–20OR030102 FW–CADIS PWR Ex-Core Analysis with Shift Through VERA

This report presents the work completed for the NEAMS milestone M2MS–20OR030102 titled "FW–CADIS PWR Ex-Core Analysis with Shift through VERA." The work completed for this milestone includes the implementation, integration, and optimization of memory and performance improvement methods in Shift for fully coupled ex-core calculations through Virtual Environment for Reactor Applications (VERA). Fully coupled in this context means the transfer of moderator boron concentration, pin-wise fission source, depleted compositions, temperatures, and moderator densities from MPACT (with COBRA-TF (CTF)) to Shift. The ability to run ex-core calculations with VERA has been enabled and used for several years by Consortium for Advanced Simulation of Light Water Reactors (CASL) partners. However, this implementation was limited and potentially computationally burdensome. This work has enabled the ability to run higher-fidelity ex-core calculations on moderate computing clusters by focusing on multithreading, domain decomposition, and Forward-Weighted CADIS (FW-CADIS) variance reduction. Tests performed on small cores, a small modular reactor (SMR), and CASL progression problems show very promising memory reduction and computational performance. Recommendations for settings when running fully coupled high-fidelity ex-core calculations with VERA are documented. Without these optimization methods, many processors on a compute node would be left unused for the entire ex-core calculation. Therefore, these methods enable the user to better use the resources available and reduce computation time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Preliminary Design of a New Pulsed-Neutron Die--Away Experiments with Absorbers (IER-552 CED-1)

This report presents the preliminary design (CED-1) of IER-552, pulsed-neutron die-away (PNDA) experiments with absorbers, to be conducted by Lawrence Livermore National Laboratory (LLNL). The goal of the PNDA experiments with absorbing materials is to produce high-quality, low-cost integral benchmarks that can be used to validate cross sections that are high priority nuclear data for the Department of Energy’s Nuclear Criticality Safety Program (NCSP). PNDA experiments can be used to generate nuclear data validation benchmarks quickly and cheaply, as they do not require fissile material or a nuclear facility to conduct them. The original PNDA experiments were set up at LLNL under IER-501. Extending the PNDA experiments to include absorbing materials leverages existing facilities and resources at LLNL to validate cross sections, with a basic experiment design that uses a neutron generator to impinge a short, mono-energetic neutron pulse on a poisoned, moderating target sample. The neutron pulse lasts 100 µs, after which the neutron population in the sample reaches thermal equilibrium and later spatial equilibrium. In this fully equilibrized state, the neutron population has a characteristic decay-time eigenvalue. The eigenvalue can be used as an integral parameter to validate nuclear data involved with neutron scattering and absorption. Two observables can be measured: the neutron population and the population of secondary γ-rays produced by inelastic and capture reactions. This report explores the feasibility of both approaches. For γ-ray measurements, this allows for direct measurement of reactions of interest for validation. During the neutron burst time, inelastic (prompt) γ-rays dominate, and these data can be isolated for validation of neutron inelastic scattering data. After the pulse ends, neutron capture (delayed) γ-rays dominate and are produced by many reaction channels as neutrons slow down to thermal energies. Two sets of γ-ray spectra (inelastic, capture) can be produced for validation. Additionally, the time-dependent decay of the capture γ-ray population when the neutron population is thermalized and in its fundamental mode can be used for validation. Two experiments can be imagined where pulsed neutrons are used to validate the nuclear data of absorbing materials. The first integrates well into the existing PNDA testbed at LLNL. The absorber material would be introduced into a moderating target and the same neutron generator, neutron detectors, electronics, and shielding box used for IER-501 would be employed. From the same experimental setup, many absorbing materials can be validated by simply changing the neutron poison in the moderator. The only significant modification would be the introduction of a γ-ray detector. Such an experiment would primarily serve to validate nuclear data in the thermal energy range. The second kind of experiment would validate the fast and epithermal energy range. It would involve pulsing neutron into a bare assembly of the absorber material and measuring the die-away of neutrons. The fast leakage of such an assembly requires nanosecond neutron-generator pulse widths and fast detectors/electronics that could measure ~200 ns die-away curves. This kind of experiment would ideally be performed at a cyclotron, like that available at Lawrence Berkeley National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Automated Control for Nuclear Thermal Propulsion Start-Up using MOOSE-based Applications

This report presents a Griffin/Bison/RELAP-7 numerical model of a prototypical NTP system that features fuel assemblies arranged in rings, and which was designed to simulate rapid startup transients. The physics modeled include full-core neutronics, assembly-wise heat conduction, and conjugate heat transfer, with the balance of plant mainly imposed through boundary conditions. In addition, various forms of automated reactivity control were deployed by using the MOOSE to autonomously drive the model and simulate the reactor transitioning from assumed initial conditions to nominal power in a fraction of a minute. To generate the cross-sections of the neutronics model, and in an effort to simultaneously account for the tremendous axial temperature gradients in the reactor and to limit the number of state points required for cross-section generation, the average component temperatures and hydrogen densities in the cooling channels were correlated to the average fuel and moderator temperatures, and fixed axial profiles were derived for nominal conditions and then used during the transient. With this approximation, a tractable cross-section library tabulated with fuel/moderator temperatures and CD angles was generated using Serpent. The full-core SPH correction procedure and the CD decusping technology in Griffin, respectively, ensure preservation of the multiplication factor and reaction rates at state points, along with a reasonably accurate reactivity worth between tabulated CD angles, despite using a coarse mesh. Feedback from other physics was calculated by modeling one representative fuel assembly per ring, along with the corresponding fuel and moderator cooling channels. To limit power overshoots during startup, another layer of multiphysics coupling was added to the model in order to automatically control the drums. Two different technologies presented herein showed outstanding performance in this regard: (1) a novel hybrid PID controller based on both power and reactivity signals, and (2) a PGC that relies on kinetics parameters and reactivity coefficients to predict future behavior and adjust the desired signal accordingly. A challenging benchmark was devised, featuring a power demand curve that exponentially increases by a factor of 500 within 30 seconds, then levels out after that. Both control approaches create a simulated power curve that closely follows the power demand curve and limits power overshoots to 1% or less. While the former approach requires more tuning of the internal parameters, the latter requires additional knowledge of the reactivity feedback coefficients and rates of change of the corresponding variables, including fuel and moderator temperature, which could be difficult to dynamically measure for a real NTP system. Fortunately, some inaccuracy in these quantities will not drastically degrade the PGC performance. Subsequently, a more realistic startup sequence was considered, in which the mass flow rate and outlet pressures are ramped up to model bootstrap and thrust build-up phases prior to reaching steady-state conditions, demonstrating the ability of the hybrid PID and PGCs to handle such transients, with both types of controllers exhibiting very similar behavior. Nevertheless, a significant chamber temperature overshoot was observed, caused by the demanded power signal and assumed mass flow rate. This issue could be mitigated by deploying a reactor controller that follows the chamber temperature signal and actuates both the control valves and drums (rather than using a power signal based solely on the drums to control reactivity). Enhancement of the hydrogen fluid properties available in MOOSE, as well as a better understanding of prototypical initial conditions, are also needed to further enhance this startup model. Finally, a study was performed to model decay heat post-shutdown, and to prepare for extending this model to predict shutdown behavior and post-shutdown pulsed cooling requirements.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2023

Thermal Scattering Law Data Development for Paraffin Wax

Paraffin wax is often used as a nuclear moderator to slow down the fast neutrons in experimental critical assemblies [1]. It is a colorless and soft solid material that consists primarily of straight-chain alkanes (n-alkanes), which are hydrocarbons with the general formula CnH2n+2 [2-3]. The length of the hydrocarbon chain ranges from C20 to C30 and higher [2]. It is distinguished by its solid state at room temperature and begins to melt above approximately 310 K [4]. Paraffin wax is a commonly employed substance in the manufacture of shielding. One of its noteworthy characteristics is its ability to effectively absorb the neutrons. Also, it possesses a high macroscopic cross section, which enables it to efficiently moderate neutrons. As a result, paraffin wax is extensively utilized in various applications where moderation and shielding of neutrons are needed. For simulations, it is necessary to evaluate its thermal scattering law (TSL) and cross sections. Computationally, classical molecular dynamics (CMD) simulations provide the capability of simulating atomic details. For example, several unary, binary, and few multi component mixtures have been investigated of the paraffin model by using molecular dynamics simulations [5-12]. An assessment of thermal neutron scattering in a heavy paraffinic oil treated both as a solid and a viscous fluid containing 25% linear branched paraffin (C30H62), 35% one ring cycloalkane (C30H60), 15% two rings cycloalkane (C30H58), and 25% aromatic (C30H60) chains has been studied using CMD simulations for producing TSL data [13]. Nevertheless, there is lack of TSL and cross section data for paraffin wax as most of the reported analyses focus on the unary and binary mixture of n-alkanes, which is not consistent with actual paraffin wax [2]. In this work, we applied the equilibrium CMD simulations technique to explore the structure and dynamical properties of wax, which are fundamental input to calculate the TSL. A paraffin wax system was modeled using the CMD code LAMMPS (Large-scale Atomic/Molecular Massively Parallel Simulator) [14-15] with the semi-empirical COMPASS [16] force field. The density of state (DOS) was calculated from the normalized velocity autocorrelation function (VACF), which is the Fourier transform of the normalized VACF. The DOS was used for the calculation of the TSL and thermal scattering cross sections. The paraffin wax atomic system was constructed by using the MedeA material design platform [17], and was benchmarked using available properties (i.e., density, bond lengths, angles, diffusivity, and viscosity).

Nuclear Criticality Safety Program (NCSP)↗
DOE OSTI2024

NASA: Structure and Thermophysical Properties of Delta and Epsilon Zry4H x

Small-modular and micro-reactors are part of an emerging nuclear technology aimed to provide power to decentralized grids in civilian, industrial, and defense energy sectors. Resurgence in microreactor research is partly driven by demands for a mobile, safe, and compact energy platform for remote and strategic locations. The defining characteristics of the microreactor include its high-power density on a small footprint, its highly transportable design, and the use of high-assay low-enriched uranium (HALEU) fuel. Solid-state metal hydride neutron moderator components are considered for microreactor designs where the moderator functions to thermalize neutrons into lower energies for the fuel to achieve high fission rates. Zirconium hydride (ZrH x where x denotes the atomic hydrogen-to-metal ratio) is a leading moderator candidate material due to the combination of its low neutron absorption cross section, irradiation resistance, high hydrogen atomic density (upwards of 7.5 x 10 22 H atoms/cm 3 at low temperatures), and high hydrogen desorption (dissociation) temperature around 600 °C (873 K) compared to other metal hydrides. Its high-temperature stability enables operation of fuel temperatures into several hundred degrees Celsius for overall increased efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗