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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Assessment of the Griffin Reactor Multiphysics Application Using the Empire Micro Reactor Design Concept

In late 2019, INL and ANL agreed to jointly develop the reactor physics code named Griffin based on the integration of the two code suites, MAMMOTH/Rattlesnake (INL) and MC2 - 3/PROTEUS (ANL). Griffin is being developed based on the MOOSE framework and MOOSE quality assurance procedures. This decision was made to be able to allow DOE-NE to efficiently invest funding to this area and to provide effective and timely support for existing and potential users; the latter includes industry and government organizations who are developing various types of advanced reactors in the near and long term. Since MAMMOTH/Rattlesnake has been developed based on the MOOSE framework, the INL/ANL Griffin development team agreed to build Griffin beginning with a merger of MAMMOTH and Rattlesnake into a single code and moving forward by implementing capabilities from the PROTEUS suite into Griffin. Moving forward, both ANL and INL efforts are equally invested in the Griffin project, with management support, to provide an advanced reactor multiphysics tool to assist in reactor design, optimization, and safety analysis. Much work remains in moving Griffin forward to migrate PROTEUS capabilities and to optimize performance to meet user needs. The main objective of this work is to assess the current status of Griffin capabilities in terms of performance and accuracy, to determine priorities for PROTEUS migration, and to identify capabilities and features to improve for supporting the code integration effort. For this assessment, the Empire micro reactor problem that was developed in the ARPA-E MEITNER program was selected as an advance reactor concept of interest to the technical community. The Empire reactor problem was expanded from its original incomplete specification to be a small heat-pipe-cooled micro reactor core with ~113 cm radius and 70 cm in height, composed of 18 fuel assemblies, 12 control drums, and beryllium radial and axial reflectors. In the current model, using 5 cm axial reflectors specified in the original Empire assembly model, more than 10% of neutrons leak axially and through the empty center safety hole, as well as through heat pipe channels in fuel assembly elements that extend through the top reflector region. Several calculation models of the core were defined for systematic assessment, including 2-D and 3-D fuel assemblies and whole cores with cylindrical boundaries. Cross sections were generated using Serpent 2, and meshes were produced using the Argonne mesh tool or the INL neutronics meshing tools combined with CUBIT. Cross sections and meshes were converted to the ISOXML and Exodus formats, respectively, so that Griffin and PROTEUS could use consistent data for solving the reactor problems. With the prepared cross sections and meshes, PROTEUS was run first to ensure that all input data were correctly generated and input options in terms of angle, mesh, and energy group were accurately determined. Comparisons against Serpent 2 solutions were made in terms of eigenvalue and pin power. The same calculations and comparisons were then conducted using Griffin. For the fuel assembly and whole core problems, the PROTEUS eigenvalues agreed well with reference Serpent 2 solutions within 100 and 30 pcm, respectively, and pin power differences relative to Serpent 2 were overall less than 2.2% and RMS 0.8% for the whole core models. This indicated that all input data were properly prepared. Using the same data, Griffin was run selecting the SAAF-CFEM SN solver with Legendre-Gaussian quadrature and NDA and DSA for acceleration. It was found that the SAAF-CFEM solver of Griffin required finer meshes to achieve eigenvalue and pin power solutions in good agreement with Serpent 2, consequently requiring more memory requirement and longer computation time. On the other hand, the SPH-Diffusion 2-D core calculations performed using Griffin were able to recover the exact eigenvalue from the reference Serpent 2 solutions, resulting in a pin-power distribution with an RMS of 0.6% and maximum absolute difference of less than 1.4%. The runtimes for SPH-Diffusion for the 2-D core were less than 3 minutes on 40 cores. During this evolution of this evaluation, many updates were made in Griffin by the Griffin development team of INL (focusing on software updates) and ANL (reviewing and supporting software updates) to complete this assessment. Observations from the code assessment are presented in the conclusion section of this report, followed by a discussion of recommendations for future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Path Integral Monte Carlo Simulations of Iron Plasmas (Final Technical Report)

This documents is the final technical report for our grant entitled "Path Integral Monte Carlo Simulations of Iron Plasmas" that focused on developing path integral Monte Carlo (PIMC) computer simulations. This techniques will be developed to study plasmas composed of heavier elements including iron and other third row elements. Equations of state (EOS) and transport properties will be derived in the regime of warm dense matter (WDM) and dense plasmas where existing first-principles methods cannot be applied. While standard density functional theory (DFT) has been used to accurately predict the structure of many solids and liquids up to temperatures on the order of 100,000 K, this method is not applicable at much higher temperature because the number of partially occupied electronic orbitals reaches intractably large numbers or the use of finite-temperature free energy functionals in orbital-free DFT introduces an uncontrolled approximation. Here we focus on PIMC methods that become more and more efficient with increasing temperatures and still include all electronic correlation effects. In this approach, electronic excitations increase the efficiency rather than reduce it. While it had commonly been assumed this method could only be applied to elements without core electrons, we showed that PIMC with free-particle nodes works well for first-row elements (PRL 108 (2012) 115502). Most recently, we extended the applicability range of all-electron PIMC to second-row elements by adopting localized nodal surfaces (PRL 115 (2015) 176403). To simulate third-row elements efficiently under WDM conditions, we propose a new method to remove core electrons by introducing pseudo-nodes. We explain our approach step by step and present preliminary results. We focus our method development on getting PIMC simulations of iron to work because of its fundamental importance for WDM and astrophysics. Then we move on to krypton and copper-doped beryllium, a ICF ablator material. We plan to continue working on key second-row material such as Na, Mg, MgO, Al, silica, and silicon-doped plastic ablators. Our collaborators at LLNL, will use our PIMC EOS data both as comparisons to existing semi-empirical, EOS-generating schemes, and as input for continuum radiation hydrodynamics simulations. We will establish an efficient pipeline from PIMC to macroscopic continuum studies of materials response. An emphasis will be placed on benchmarking such methods for plasmas of heavy elements at the very high temperatures (~100 eV) and low densities that are generated when Hohlraum radiation heats the ablator material in indirect drive laser experiments. Results from changes to the EOS will be of immeasurable importance to the designers at the National Ignition Facility (NIF) and at other facilities. Starting with our EOS of Cu-doped Be, our second collaborator at LLE, will perform real-time simulations of laser fusion experiments at the Omega laser and at the NIF to determine how sensitive the compression path depends on the ablator EOS. Since our collaborator also has experience in performing orbital-free DFT calculations, we propose to compare predictions from this method with PIMC results. In joint publications, we plan to analyze the accuracy of different free-energy functionals in order to understand why existing orbital-free DFT calculations do not predict compression peaks along the shock Hugoniot curve that we see with PIMC. The peaks are caused by the ionization of various electron shells. Their accurate characterization is important to compare with experimental results. We will break new ground by developing PIMC techniques that can simulate iron and all other third row elements in the plasma and WDM regimes. We introduce the concept of pseudo-nodes for the efficient treatment of the core-electrons. The EOS and transport properties will be derived and published online in the form of a new WDM database. Our PIMC EOS calculations will benchmark and possibly replace semi-analytical EOS tables like QEOS or SESAME, which will impact the hydrocode simulation community and will affect the design of NIF targets. Our PIMC results will help to improve the accuracy of orbital-free DFT simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Gamma ray analysis of the K2 containers

The K2 containers contain PuO2 with 238 Pu being the most abundant isotope of all plutonium isotopes. The purpose of this analysis is to determine the light element impurities in the PuO2. These light elements can interact with the alpha particles from plutonium decay to release gamma rays, protons, and neutrons through the reactions (α,α’), (α,p), and (α,n) respectively. The gamma rays released by the interactions can be measured and compared with the gamma rays released by the plutonium. The ratios of the gamma rays from the interactions and the plutonium gamma rays allow for the determination of the impurities in the items. The elements that can be detected are Lithium (Z=3), Beryllium (Z=4), Boron (Z=5), Fluorine (Z=9), Sodium (Z=11), Magnesium (Z=12), Aluminum (Z=13), Silicon (Z=14), Phosphorous (Z=15), Chlorine (Z=17), and Potassium (Z=19).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Integral Experiment Request 523 Feasibility Study (Summary Report)

This report documents the feasibility phase of the Critical Experiment Design (CED) conducted as part of integral experiment request (IER) 523. The purpose of IER-523 is to explore the effects of using 35 weight percent enriched uranium dioxide-beryllium oxide (UO 2 -BeO) material on critical configurations using the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories (Sandia). Preliminary experiment design concepts, neutronic analysis results, and proposed paths for continuing the CED process are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effects of Electron Irradiation on Candidate Materials for Target Window in Accelerator-Driven Mo-99 Production

Room-temperature (RT) axial tensile tests were conducted in 2018 and 2021 with flat, sub-sized, dog-boned samples of non-irradiated and irradiated Inconel 718 and irradiated beryllium (Be). The sample gauge length was 7.62 mm. Tensile properties of interest are the engineering yield stress (YS), ultimate tensile stress (UTS), uniform elongation (UE) and total elongation (TE). The electron dose of the 2021 samples was ≈2 times the dose of samples tested in 2018. Relative dose rates of 0, 1 (2018), and 2 (2021) are assigned to these samples. The 2021 electron-irradiated Be samples were brittle with a low failure stress of 338±64 MPa based on two samples. The failures occurred in the gauge section. The 2018 results for electron-irradiated Be indicated a slight decrease in ductility and strength for the one electron-irradiated sample that failed in the gauge section.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

A study of sacrificial mirrors for use prior to a laser wakefield accelerator driven by the Z-Petawatt laser

Many experiments at Sandia’s Z Pulsed Power Facility require x-ray backlighting diagnostics to understand experiment performance. Due to limitations in present-day source/detection modalities, most x-ray diagnostics at Z are restricted to photon energies <20 keV, ultimately limiting the density, amount, and atomic number of targets diagnosable in experiments. These limitations force the use of low-Z materials like Beryllium, and they prevent acquisition of important backlighting data for materials/densities that are opaque to soft x-rays and where background emission from the Z load and transmission lines overwhelm diagnostics. In this LDRD project, we have investigated the design and development of a laser wakefield acceleration platform driven by the Z-Petawatt laser – a platform that would enable the generation of a pulsed, collimated beam of high energy x-rays up to 100 keV. Geometrical considerations for implementation on the Z Machine require the use of sacrificial mirrors, which have been tested in offline experiments in the Chama target chamber in building 983. Our results suggest the use of sacrificial mirrors would not necessarily inhibit the laser wakefield x-ray process, particularly with the benefits stemming from planned laser upgrades. These conclusions support the continuation of laser wakefield source research and the development of the necessary infrastructure to deliver the Z-Petawatt laser to the Z center section along the appropriate lines of sight. Ultimately, this new capability will provide unprecedented views through dense states of matter, enabling the use of previously incompatible target materials/designs, and uncovering a new set of observables accessible through diffraction and spectroscopy in the hard x-ray regime. These will amplify the data return on precious Z shots and enhance Sandia’s ability to investigate fundamental physics in support of national security.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

A User Guide to PARET/ANL

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA since the inception of the Reactor Conversion Program. Since Reduced Enrichment for Research and Test Reactors (RERTR) began in 1978, PARET/ANL has been benchmarked to experimental data including SPERT testing, and used to determine the expected transient behavior of a large number of reactors regardless of enrichment both inside and outside the Reactor Conversion Program. This document provides the pertinent information for the use of PARET/ANL Version 7.6. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

SCALE Modeling of the Fast Spectrum Heat Pipe Reactor

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

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Responses to Nonconformances, AIHA Onsite Assessment August 29-31, 2022

A nonconformance report (NCR) was not generated to document the original Beryllium (Be) positive result being changed to negative due to an oversight by the lab. The sample was re-analyzed on the Agilent ICP-OES which includes software to deal with interferences and resulted in a negative for Be. In order to determine the cause of the false positive, which was originally thought to be due to the Titanium interference in the sample, the laboratory analyzed a blank and a reporting limit standard with 5 ppm and 10 ppm of Titanium added in. These were analyzed on the Perkin Elmer ICP 5 instrument to see if the positive Be was reproducible.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

A Search for Sterile Neutrino Dark Matter (Final Report)

Decades of dark matter searches with increasingly higher sensitivity have so far not been able to detect the most-favored hypothetical dark matter particles. This has increased interest in sterile neutrinos with masses in the kilo-electronvolt (keV) range as possible dark matter candidates. We have established a and new and world-leading approach to search for keV-scale sterile neutrinos with superconducting radiation detectors. For this, we have implanted radioactive beryllium-7 (Be-7) nuclei into our superconducting tunnel junction (STJ) detectors and measured the recoil energy of the lithium-7 (Li-7) daughter nucleus with high accuracy. If the associated electron neutrino wave function contains some admixture of a keV sterile flavor, the Li-7 recoil energy will be shifted and produce an additional peak in the spectrum whose energy depends on the sterile neutrino mass. While we have not observed any evidence for sterile neutrinos in these precision decay experiments, the experiment has increased the sensitivity of sterile neutrino searches by a factor of ~10 over competing approaches in the mass range from ~10 to ~800 keV. The next generation of STJ detectors that we have developed as part of this project will be able further increase the sensitivity of the sterile neutrino search by at least another two orders of magnitude.

42 ENGINEERING↗
DOE OSTI2022

Thermal Neutron Cross Section Measurements at the RPI LINAC [Slides]

Cold polyethylene shows adequate capacity for enhancing sub-thermal neutron flux. Cold moderator system produced high quality thermal cross section measurements. Measurements were performed for polyethylene, polystyrene, Plexiglas, yttrium hydride and beryllium and compared with ENDF/B-VIII.0 TSL evaluations. All data reduction for all measurements was finalized. Future work includes completion of necessary journal publications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Assembly of MiniFuel Targets for Irradiation of U-Mo Fuel Specimens in the High Flux Isotope Reactor

To support the development of advanced low-enriched uranium for use in nuclear reactors, irradiation testing of U-Mo disk specimens was performed at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) to collect experimental irradiation data on this type of fuel at pressurized water reactor–relevant temperatures. U-Mo is a uranium alloy that has superior dimensional stability relative to alpha-phase uranium metal and has a substantially higher uranium density compared to UO 2 . U-Mo disks specimens were fabricated at Idaho National Laboratory and inserted into MiniFuel targets for HFIR irradiation. Three MiniFuel targets were successfully assembled, welded, tested, and delivered to HFIR, along with their quality assurance documentation. The targets were inserted into HFIR’s inner vertical experiment facility within the permanent beryllium reflector. Each target contains six disk specimens and will be irradiated in HFIR for three, four, and eight cycles, with target temperatures between 250 and 500°C. This report summarizes the experiment design, test matrix, pre-characterization of specimens, and experiment assembly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

IER - 551: Experiment for Unresolved Resonance of Plutonium Actinides (EUROPA): CED-1 for True Intermediate Pu Benchmark

IER-551 or the Experiment for Unresolved Resonance Of Plutonium Actinides (EUROPA), is an experiment campaign to design and execute a series of true intermediate energy plutonium critical configurations. EUROPA is a collaboration between Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), and the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to validate current and future evaluations of 239,240 Pu isotopes. Preliminary designs were optimized using Particle Swarm Optimization to save on computation time and the configuration with highest percentage of intermediate neutrons causing fission found was BeO-Cd. This combination has over 70% of fissions occurring from intermediate energy neutrons. Secondary considerations regarding end benchmark uncertainties and quality of nuclear data resulted in a beryllium metal only experiment proposed as the final system. This final preliminary design achieves 68% intermediate fissions, sensitivities of fission and capture cross sections of 239 Pu of 0.3021 and -0.1610 respectively.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

LEU Fission Plate: Experimental Prototype Design Document

The Ultracold Neutron (UCN) source at LANSCE is currently the source of the highest density of UCN in the world. We intend to use innovative new solutions to boost our UCN density in order to maintain our leadership in this competitive field. Prior studies suggest that a fission plate neutron multiplier made from low enriched uranium (LEU) installed between our spallation target and UCN source moderator can boost the density by up to a factor of ten. Here we present a prototype of this idea that uses an existing one liter volume of LEU that can boost the UCN density by roughly a factor of 2. Additional simulation studies presented here also suggest that we may find more gains by replacing some of the moderator or reflector geometries with beryllium.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

A Fission-like Neutron Spectrum Shaping Assembly based on D-T Generator

For nearly five decades, 252 Cf sources have been used for testing and calibration over a wide range of neutron detection devices used for nuclear safety and radiation protection. However, the vastly increased cost of 252 Cf sources, its short half-life, and concerns about future shortage have prompted nuclear facilities and radiological calibration laboratories that rely on high-intensity neutron sources to seek alternatives. While there appear to be no other radionuclide neutron sources that would match the 252 Cf performance characteristics, all of them, including 252 Cf, present similar concerns in terms of hazards, safety, and security. As a possible solution to this issue, the Pacific Northwest National Laboratory (PNNL) developed a low-risk, cost-efficient alternative to produce fission-like spectra based on commercially available deuterium-tritium (D-T) neutron generators. The alternative of producing a fission-like neutron spectrum without any fissionable materials or beryllium would enable the nuclear facilities across the entire complex to eliminate their reliance on high intensity 252 Cf sources for nuclear safety and radiation protection applications, substantially reducing risks and associated costs. This report describes efforts of designing the spectrum shaping assembly, its construction and characterization of the neutron field produced. Performance of this newly developed D-T generator based neutron field is demonstrated via calibration of a criticality detector, neutron survey instruments, and personal dosimeters verified against current standardized fields produced by 252 Cf spontaneous fission sources.

42 ENGINEERING↗
DOE OSTI2023

Comparison of the Neutronic Performance of IRP-1 and IRP-2 at SNS

The Spallation Neutron Source (SNS) operated from start-up, in May 2006, through December, 2017 before replacing the Inner Reflector Plug (IRP). This component is the most costly and time-consuming “regular maintenance” replacement foreseen in the operation of the First Target Station (FTS). The IRP includes beryllium reflector blocks, steel and aluminum support structures, aluminum moderator vessels (which in turn include gadolinium and cadmium neutron poisons) and water coolant systems. The lifetime of this multi-million dollar component is driven by poison and decoupler burnup in moderator vessels. The moderators include these neutron poisons to best match neutron beam performance with scattering instrument needs. The burnup that comes with use changes the performance of the moderators until they no longer server their intended purpose. The reasonable desire to maximize the IRP lifetime means that the moderator performance at the Beginning Of Life (BOL) is significantly “over poisoned” and different from the performance at moderator End Of Life (EOL). Upon installation of the second IRP article, IRP-2, the SNS Neutronics team characterized the performance of the neutron beamlines in order to assess the differences between them, primarily by comparing the EOL performance of IRP-1 to the BOL performance of IRP-2. While IRP-2 was largely similar to IRP-1 in its conceptual design, there were some differences in the specific design of IRP-2 to support enhanced manufacturability, and the change from IPR-1 to IRP-2 coincided with the replacement of the light water in the IRP cooling loop with heavy water, as had been intended in the SNS design but not implemented during construction because of heavy water availability.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Be 2 C synthesis, properties, and ion-beam irradiation damage characterization

This technical report is in submission of completion of the Level 2 milestone number M2AT 23OR1101022, titled “Be 2 C synthesis, properties, and ion-beam irradiation damage characterization,” for the AT-23OR110102 “Beryllium Carbide as moderator for MSRs—ORNL” WBS number 2.04.11.01 within the larger Advanced Reactor Technologies program at Oak Ridge National Laboratory. This effort is an initial study of the viability of Be 2 C as a future neutron moderator for molten salt reactors and other high-temperature reactors. The research was awarded funding in September 2022, and this Level 2 milestone presents the work that completed by the end of July 2023. This work is a collaboration between Oak Ridge National Laboratory and the University of Michigan. This report discusses the status of this effort and planned work before the final reporting.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

NB3 Bio-SANS preliminary optics design

This report describes the calculations and simulations to determine the optimal guide coatings for the NB3 (Bio-SANS) beam line post beryllium reflector changeout. The reported guide performance meets the Bio-SANS scientific requirements document.

43 PARTICLE ACCELERATORS↗