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NASA NTRS1984

Radioisotope Space Power Programs

Radioisotope Thermoelectric Generators (RTGs) were extensively used in past space missions with great success. An improved generation of RTGs employing a new lightweight, modular heat source is being built by the DOE for NASA launches in 1986. More advanced modular RTGs which promise power-to-weight ratios of over 8.8 watts(e)/kg (4 watts(e)/lb) are currently under development by the DOE and could be flight-ready within five years. Dynamic Isotope Power Systems (DIPS) which offer 18-20% conversion efficiencies were demonstrated by the DOE in ground tests. DIPS would be useful for several military missions requiring power in the low-kilowatt range. These systems could also be brought to flight readiness by the DOE following receipt of firm user requirements.

Rock, B. J.↗

NASA NTRS2006

Thirty Years After: The Science of the Viking Program and the Discovery of a 'New Mars'

Viking discovered a Mars that was very different from the Mars found by Mariner 4, 6 and 7. The new, exciting, more Earth-like Mars was hinted at by the Mariner 9 orbiter and confirmed by Viking. Viking discovered some very fundamental things about Mars. Viking discovered the presence of nitrogen in the atmosphere. A key ingredient needed for life. Viking made the first measurements of the isotopic composition of carbon, oxygen, nitrogen and the noble gases in the atmosphere of Mars. The ratio of 15N to 14N suggested that Mars may have lost more than 99% of the total mass of its atmosphere. The denser atmosphere in the past may explain the presence of flowing water earlier in the history of Mars first discovered by Mariner 9 with additional and higher spatial resolution examples provided by the Viking Orbiters. Viking did not measure organics or life at the surface of Mars. But, Viking did discover a surface unlike any other on the Solar System--a surface exhibiting very high chemical reactivity, most probably formed by the deposition of chemically active atmospheric gases, like hydrogen peroxide (H2O2) and ozone (O3), onto the surface of Mars.

Levine, Joel S.↗

DOE OSTI2020

Proposed guidance for preparing and reviewing a molten salt non-power production or utilization facility application

Development of non-power molten salt reactors (MSRs) are under consideration to further establish an MSR experience base, support the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.43(e), and provide any additional analyses needed for development of a full-scale MSR. Guidance provided in this report is based on MSRs operating with liquid fuel (i.e., fuel dissolved within a molten salt). These reactors, unless owned by the DOE or DOD, will require licensing by the US Nuclear Regulatory Commission (NRC) staff. Standard review plan (SRP) guidance for large light water reactors (LWRs) is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants; Light Water Reactor (LWR) Edition. However, NRC staff observed that NUREG-0800 is very cumbersome to apply to non-power reactors “because of the great differences in complexity and hazards between non-power reactors and nuclear power plants.” Therefore, a program to develop performance-based guidance applicable to non-power reactors was initiated. In 1996, NUREG-1537, Parts 1 and 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, was published. Part 1, the format and content guide, suggests a uniform format for presenting information in non-power reactor applications that is acceptable to the NRC staff, but conformance with the format and content is not required. Part 2, the SRP, ensures the quality and uniformity of the staff review of an application. Unfortunately, the application guidelines and SRP do not provide adequate guidance for all advanced non-LWR technologies and applications. This discrepancy eventually led to the 2012 development of interim staff guidance (ISG) for NUREG-1537, which includes criteria for describing and reviewing aqueous homogeneous reactors (AHRs). Specifically, NUREG-1537 ISG, 2012 expanded the original document to address three areas: 1. updated criteria for heterogeneous non-power reactors, 2. criteria for licensing AHRs, and 3. criteria for licensing a Part 50-licensed isotope production facility. In 2015, the US Department of Energy (DOE) opted to build on the AHR NUREG-1537 ISG experience by performing a gap analysis of the guidance that would be used to license a non-power MSR. MSRs represent one of the advanced non-LWR technologies selected by DOE for development through a multiyear cost share award with Southern Company Services. Under this Advanced Reactor Concepts 2015 (DOE Advanced Reactor Concepts [ARC] 15) award program, the DOE tasked Oak Ridge National Laboratory (ORNL) to evaluate the guidance changes that the NRC may need to consider when licensing an MSR non-power reactor. ORNL staff, with support from Boston Government Services, LLC, focused on five system-related chapters in NUREG-1537 that were considered most relevant to inform the effort that would be required for a non-power MSR applicant. ORNL documented this review in a technical report, ORNL/TM-2018/834, Proposed Guidance for Preparing and Reviewing Molten Salt Non-Power Reactor License Applications (NUREG-1537). The report was subsequently shared with industry and the NRC. The 2018 review was limited in scope, focusing on key system chapters based on the expected significance of each chapter relative to expected differences in addressing advanced non-LWR technologies, specifically non-power MSRs, compared with heterogeneously fueled non-power reactors. In the ORNL report, proposed generic adaptations were suggested for the following NUREG-1537 chapters: Chapter 4, “Reactor Description”; Chapter 5, “Reactor Coolant Systems”; Chapter 6, “Engineered Safety Features”; Chapter 9, “Auxiliary Systems”; Chapter 11, “Radiation Protection Program and Waste Management” The inclusion of Chapter 11 in the previous review effort was intended to provide guidance for categorizing the waste-handling process for an MSR operating with homogenous fuel. The introductions from Parts 1 and 2 of the 2012 NUREG-1537 ISG provide guidance for the application and review of production facilities. After a period of operation, non-power MSRs with homogenous fuel will include gaseous and soluble fission products. The gaseous fission products will be collected and held for decay in an off-gas system. There might also be an initiative to polish or filter the soluble fission products in the fuel salt by some mechanical or chemical means. The treatment and handling of fission products in the non-power MSR fuel salt and the description of this process in the safety analysis report (SAR) must be very precise to avoid the waste treatment facility being construed as a co-located special nuclear material (SNM) fuel cycle facility (see Section 2.3 of this report). Subsequent to the release of ORNL/TM-2018/834, NRC staff expressed a desire to continue the regulatory gap analysis that was begun in that report. This would provide additional clarity and information addressed in certain sections of the original report, while also providing new guidance on certain topics not addressed in the original report. This revision would benefit the NRC staff reviewing applications involving non-power MSR designs and would help developers understand how the NRC staff might approach the review of such applications. The focus of this report is to provide infrastructure support to the NRC staff for the regulatory review of non-power MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2020

Nano-ALSEP: Use of Nano-Sat Technology to Advance Lunar Exploration Science/Engineering

During the Apollo program, unique ALSEP (Apollo Lunar Surface Experiments Package) stations were deployed and left on the surface for long term data collection. Powered by a SNAP-27 electrical power source, the packages continued to transmit useful data until September 30, 1977 (the end of the Fiscal year and termination of the budget). The instruments were varied but included an active seismic experiment (ASE), charged particle experiments, magnetometer experiment, micro-meteoroide/bolide detection experiments. Only only one, the Laser Ranging Retro-Reflector (LRRR) was continued to be used to the present day due to the passive nature of the device. After a dearth of data during an absence of 50 years, is it possible to restart similar packages using the advances in nano-sat technologies? An innovation effort is proposed that would first look at a 2U nano-sat package evolving to an eventual radio-isotope powered system that could be either part of a lunar lander or as an independent package. In the first phase, the packages would only survive during the Lunar day. In a second phase, the later radio-isotope options would place the surface packages in a semi-dormant mode during the Lunar night cycle. Communication capability developed in the TechEdSat-n flight series (UHF, S, X and optical bands) would help to independently close the link and potentially explore IoT devices that were powered only in the sunlight. In addition, limited mobility in the form factor of a nano-rover might also greatly extend the utility of such packages while adding evolutionary capability - and finally continue where the Apollo program was forced to leave off.

M Murbach↗

DOE OSTI2020

Initial Radiation of Nuclear Power Plant Components (CRADA NFE-17-06892 Final Report)

As advanced reactor concepts are developed, instrumentation and control (I&C) cables and sensors will play a significant role in assuring their operation is well understood, and safety-related process information can be collected reliably over the course of their design life. As a first step to develop acceptance criteria and define the limits of currently available nuclear I&C cable and sensor technology, a collaboration between Oak Ridge National Laboratory and Analysis and Measurement Services (AMS) Corporation was initiated under the Gateway for Accelerated Innovation in Nuclear (GAIN) Program. This collaborative effort focused on providing technical information related to the combined thermal and radiation aging of nuclear power plant (NPP) cables, companion NPP cable insulation samples, and two prototype temperature and two pressure sensors. This work was carried out using the instrumented gamma irradiation system that is part of the High Flux Isotope Reactor Gamma Irradiation Facility (HFIR GIF). This system provided the opportunity for the condition of the prototype sensors to be tracked during the irradiation in order to determine their functional limits with greater granularity when compared to a simple irradiation insertion at the HFIRGIF. Experimental assemblies for sensor irradiation were successfully completed, and sensor irradiation to accumulated dose of 3000 kGy at a rate of approximately 6 kGy/hr was carried out. Two prototype temperature assemblies were continuously monitored during exposure to 3000 kGy and found to have no observable degradation from the irradiation. Two pressure sensors were first irradiated to 1000 kGy. At which point, one of the sensors was found to operate normally; however, a second sensor failed the operability test. The operable pressure sensor output was then continuously monitored from 1000 kGy to 3000 kGy. The sensor output exhibited steady trendable drift versus the accumulated radiation dose. At the conclusion of the irradiation, the four test sensors were released for laboratory calibration checks. The calibration check of the temperature sensors found no significant change after irradiation compared to baseline measurements before irradiation; however, both pressure sensors failed the post-test calibration checks and both sensors were not functioning properly. These results will help inform sensor manufacturers and I&C design engineers of potential limitations of these types of sensors for use in high radiation environments associated with the compact configuration of new Small Modular Reactors (SMRs) such as NuScale’s reactor power module. Irradiations on NPP cables and companion samples were not completed due to programmatic delays, however a path forward that was based on lessons learned from this project was defined to accommodate future, comparable instrumented gamma irradiations at the HFIR GIF.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Performance Improvements of the Griffin Solvers in FY24

The Griffin code is a MOOSE-based reactor physics application jointly developed by Idaho National Laboratory and Argonne National Laboratory under the Department of Energy Office of Nuclear Energy Nuclear Energy Advanced Modeling and Simulation Program. This fiscal year, we have made significant efforts to improve the performance of transport solver options and cross-section generation for the efficient use of Griffin in advanced reactor applications. For the HFEM-PN solver, the residual evaluations of HFEM kernels were optimized by utilizing the pre- computed averaged cross sections for individual elements. Numerical integration involving the evaluation of basis functions at quadrature points was bypassed by facilitating precomputed element mass matrices for response matrices. Red-black iterations were improved by introducing a new generalized minimum residual based solver. The memory usage of response matrix storage was significantly reduced by applying basis function rotations on interfaces and calculating volumetric odd-parity moments on the fly. Additionally, the adjoint flux and transient calculation capabilities of the HFEM-PN solver were successfully implemented and verified using the TWIGL benchmark problem. For the DFEM-SN solver, memory footprint and computation time were significantly reduced by not treating angular flux vectors as the MOOSE nonlinear system vectors. Specifically for IQS, scalar adjoint weighting was introduced to further eliminate angular adjoint flux storage in the MOOSE auxiliary system. It was demonstrated through the three-dimensional Advanced Burner Test Reactor core problem that the memory usage for transient calculations with the IQS method was reduced by over 7.5× compared to before the optimizations. For the self-shielding application programming interface, a new double-heterogeneity treatment method, named the Bell Function-Based Analytic Two-Region Slowing Down Method, was developed to efficiently flux-volume homogenize TRISO particles with the matrix. Additionally, optimizations were made to hyper- fine group (HFG) slowing down calculations by pretabulating collision probability coefficients and grouping isotopes, significantly reducing the computational time for calculating scattering sources per HFG. Lastly, the pin power reconstruction module was extended to account for temporal behavior in a microreactor analysis problem, specifically for a control drum transient. Verification tests for each of these improvements demonstrated significant performance enhancements and memory reduction.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1990

Microbial mats and the early evolution of life

Microbial mats have descended from perhaps the oldest and most widespread biological communities known. Mats harbor microbes that are crucial for studies of bacterial phylogeny and physiology. They illustrate how several oxygen-sensitive biochemical processes have adapted to oxygen, and they show how life adapted to dry land long before the rise of plants. The search for the earliest grazing protists and metazoa in stromatolites is aided by observations of mats: in them, organic compounds characteristic of ancient photosynthetic protists can be identified. Recent mat studies suggest that the 13C/12C increase observed over geological time in stromatolitic organic matter was driven at least in part by a long-term decline in atmospheric carbon dioxide levels.

NASA Center ARC↗

NASA NTRS1990

Origin of petroporphyrins. 2. Evidence from stable carbon isotopes

Compared with the carbon-13 isotopic composition of the ubiquitous C32DPEP (DPEP, deoxophylloerythroetioporphyrin) the heavy but equivalent carbon-13 isotopic composition for the porphyrin structures 15(2)-methyl-15,17-ethano-17-nor-H-C30DPEP and 15,17-butano-, 13,15-ethano-13(2),17-propano-, and 13(1)-methyl-13,15-ethano-13(2),17-propanoporphyrin suggests a common precursor, presumably chlorophyll c, for these petroporphyrins isolated from the marine Julia Creek oil shale and the lacustrine Condor oil shale. Similarly, the heavy but variable carbon-13 isotopic composition of 7-nor-H-C31DPEP compared with C32DPEP is consistent with an origin from both chlorophyll b and chlorophyll c3. The equivalent carbon-13 isotopic composition for 13(2)-methyl-C33DPEP compared with C32DPEP suggests a common origin resulting from a weighted average of chlorophyll inputs.

NASA Program Exobiology↗

DOE OSTI2020

VENTSAR V3.0: A Python GUI for Estimating Contaminant Concentrations on or Near Buildings Due to Building Effects and Plume Rise and For Calculating Inhalation and Plume Shine Doses

VENTSAR: Originated as VENTAX (Smith and Weber 1983) on the IBM Mainframe at the Savannah River Site (SRS) as a Fortran program. Updated to VENTSAR XL V1.0 (Simpkins 1997) as a spreadsheet version using macros created in Microsoft Excel. Later updated to VENTSAR XL V2.0 (Dixon 2018) as a spreadsheet executing a modified version of the original VENTAX Fortran code. Estimates contaminant concentrations on or near a building from a release at a nearby location. Calculates concentrations for a given meteorological exceedance probability or for a given stability and wind speed combination. Can model a single building with or without a penthouse on top or a ground location from either a stack or ground release. Plume rise can be considered. Contaminant releases can be chemical or radioactive with downwind concentrations determined at user-specified distances. Wind passing over and around buildings creates a complicated dispersion pattern. Air-intake vents may be located on building roofs or near the ground downwind of a release source. An estimation of pollutant concentrations on or near a structure is important in determining expected pollutant levels. Meteorological data are selected based on a specific area of the site. Fortran-based VENTAX able to make fast, complex mathematical calculations. Not user-friendly VENTSAR XL V1.0 no longer supported due to obsolete Excel macros. VENTSAR XL V2.0 an interim solution using an Excel spreadsheet to interface with the modified VENTAX Fortran code. Requires user to have Microsoft Excel. Not an intuitive interface for someone unfamiliar with VENTSAR. Does not perform dose calculations. VENTSAR V3.0 goal to combine the strengths of previous versions of VENTSAR into a single, powerful yet user-friendly program with a Graphical User Interface (GUI). Not dependent upon a specific program or plug-ins. Self-contained package to be used on any computer running Microsoft Windows. User-friendly, intuitive GUI created in Python for parameter input. Executes same modified VENTAX Fortran code as VENTSAR XL V2.0. Outputs formatted text file of completed calculations for easy review and dissemination. Performs inhalation and plume shine dose calculations for up to 11 user-selected nuclides from a nuclide dose factor library of almost 500 nuclides. 12 test cases were created for verification of V1.0 and V2.0. Same test cases run in V3.0 to verify correct performance. V3.0 produced similar results to V1.0. Confirmed GUI did not alter calculations in any way. Comparisons of the test case results confirm the V3.0 GUI does not influence the VENTSAR calculations. Simply passes same input parameters to VENTAX Fortran code for execution. Provides end user an easy-to-use, intuitive tool to quickly make building effect and plume rise calculations, as well as, inhalation and plume shine dose calculations. No experience with or working knowledge of Fortran, Excel, command line, or macros required. Self-contained package allows VENTSAR be deployed to any Windows computer without the need for specific programs or plug-ins.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Laser ablation spectrometry for studies of uranium plasmas, reactor monitoring, and spent fuel safety

Nuclear security is one of the defining challenges of our time. Nuclear threats range from deliberate dispersal of radioactive material to contaminate the vital infrastructure to diversion and smuggling of special nuclear material for clandestine nuclear programs and nuclear terrorism, respectively. There is an associated need to develop and sustain the nuclear forensics capabilities, which requires the understanding of complex processes that occur in plasmas of nuclear materials. The area of nuclear safety has seen a resurgence of public interest, and there is a concomitant need to safely store used nuclear fuel and detect structural material failure in nuclear power systems, especially in innovative reactor designs envisioned for future adoption. Laser-produced plasmas are complicated extreme environments that can generate intense and rich, highly specific signatures of nuclear and radiological materials, which can then be explored in a wide range of applications. They include interdiction and rapid detection of nuclear materials, including their isotopic composition, detection over long distances, laboratory simulation of weapons effects, monitoring the condition of structural materials in dry cask storage containers, and novel instrumentation for nuclear power systems. We present a compilation of recent representative examples of the application of laser spectroscopy, and laser-induced breakdown spectroscopy in particular, to nuclear safety and security problems. A case is made that spectroscopic techniques based on laser-produced plasmas offer complementary, and sometimes unique, capabilities that motivate the continued exploration of their efficient production and understanding of the signatures they produce

(020.3260) Isotope shifts, (140.3440) Laser-induce↗

NASA NTRS1976

Heavy ion fragmentation experiments at the bevatron

Collaborative research efforts to study the fragmentation processes of heavy nuclei in matter using heavy ion beams of the Bevatron/Bevalac are described. The goal of the program is to obtain the single particle inclusive spectra of secondary nuclei produced at 0 deg by the fragmentation of heavy ion beam projectiles. The process being examined is B+T yields F + anything, where B is the beam nucleus, T is the target nucleus, and F is the detected fragment. The fragments F are isotopically identified by experimental procedures involving magnetic analysis, energy loss and time-of-flight measurements. Effects were also made to: (a) study processes of heavy nuclei in matter, (b) measure the total and partial production cross section for all isotopes, (c) test the applicability of high energy multiparticle interaction theory to nuclear fragmentation, (d) apply the cross section data and fragmentation probabilities to cosmic ray transport theory, and (e) search for systematic behavior of fragment production as a means to improve existing semi-empirical theories of cross-sections.

Heckman, H. H.↗

DOE OSTI2013

The neutron skin in neutron-rich nuclei at Jefferson Lab

The Jefferson Lab program to measure the symmetry energy of neutron-rich nuclear matter, using precision electroweak methods, is progressing well. The initial measurement by the PREX experiment, leading to a 2-sigma determination of the 'neutron skin' in {sup 208}Pb, has been published. Design and preparation for a further, more-precise measurement on {sup 208}Pb is progressing well and there is general acceptance of the great advantage to a further measurement on {sup 48}Ca. The surprising ancillary result that the beam-normal single-spin asymmetry for {sup 208}Pb is consistent with zero is also now in the literature. This paper will discuss the current experimental situation of the program.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2018

Analytical chemistry of nuclear material: case studies from Los Alamos National Laboratory

Presented here are three case studies that demonstrate the breadth of bulk actinide analytical chemistry at Los Alamos National Laboratory (LANL). This laboratory was established during the Manhattan Project and is still considered one of the premier analytical laboratories for special nuclear materials characterization around the globe. First, the analytical chemistry of 238 Pu is presented, showcasing the unique challenges of this high activity isotope. Next, the analysis of 241 AmO 2 is discussed from the perspective of supporting nascent production at LANL. Finally, the analytical chemistry support of the production of plutonium non-destructive assay standards is discussed. Furthermore these examples paint a picture of the analytical capabilities needed to meet the dynamic challenges of LANL’s nuclear programs.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2024

The hypernuclear physics program at Jefferson Lab

Missing mass spectroscopy of Λ hypernuclei using the (e, e' K⁺) re action has been performed in the past at the Thomas Jefferson National Accelerator Facility (Jefferson Lab) by several experiments in Halls A and C. A new experimen tal campaign is expected to start running in 2026 in Hall C and to provide the first study of the isospin dependence in medium-mass hyperisotopes by populating $^{40}_Λ{K}$ and $^{48}_Λ{K}$ using isotopically enriched calcium targets [E12-15-008]. During this cam paign, it will be possible to study Λ interactions in nuclear matter using a lead target [E12-20-013]. Solid-state targets made of lithium, beryllium, boron [LOI12-23-013], and aluminum [LOI12-23-016] are considered to be included in the campaign. The use of the HKS and HES spectrometers together with thin target foils and high beam currents leads to a sub-MeV energy resolution, significantly better than in hadron beam experiments. Valuable information on few-body hyperon systems would be gained if helium targets were used [E12-19-002]. Using an additional spectrometer for measuring the decay-pion spectra could give access to the binding energies of light hyperfragments [LOI12-23-011]. The measurement of precise and accurate en ergy spectra of different hyperisotopes probes the Λ-N interaction in nuclei including the Λ-N-N interaction. The latter is assumed to play a key role for the stiffness of the nuclear equation of state relevant for the stability of neutron stars.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

PU-ICE Summary Information.

The Generator Knowledge Report for the Plutonium Isentropic Compression Experiment Containment Systems (GK Report) provides information for the Plutonium Isentropic Compression Experiment (Pu- ICE) program to support waste management and characterization efforts. Attachment 3-18 presents generator knowledge (GK) information specific to the eighteenth Pu-ICE conducted in August 2015, also known as ‘Shot 18 (Aug 2015) and Pu-ICE Z-2841 (1).’ Shot 18 (Aug 2015) was generated on August 28, 2015 (1). Calculations based on the isotopic content of Shot 18 (Aug 2015) and the measured mass of the containment system demonstrate the post-shot containment system is low-level waste (LLW). Therefore, this containment system will be managed at Sandia National Laboratory/New Mexico (SNL/NM) as LLW. Attachment 3-18 provides documentation of the TRU concentration and documents the concentration of any hazardous constituents.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Effect of Reactor Radiation on the Thermal Conductivity of Graphite-based Dispersion Fuel

The Fukushima Daiichi nuclear accident in 2011 stimulated materials development efforts to improve the current fuel design for Light Water Reactors (LWRs). From fuel element to fuel cladding, new concepts in advanced nuclear fuel development have been proposed, studied, and tested. The new fuel systems are expected to have better resistance to damage, melting, and chemical reaction during accident scenarios. To test the performance and accident tolerance of the advanced fuel, it is highly preferable to simulate off-normal transients in addition to standard in-core fuel performance tests (available at High Flux Isotope Reactor at Oak Ridge National Laboratory and Advanced Test Reactor at Idaho National Laboratory). The Transient Reactor Test Facility (TREAT) is now being prepared to restart its operation, performing transient testing to assist advanced fuel development. TREAT was operated from 1959 to 1994, and was extensively used to generate thousands of reactor transients for different programs. The reactor has not operated since 1994. The original fuel used at TREAT was graphite-based dispersion fuel with highly-enriched uranium (HEU, 93.1% enriched uranium). For nonproliferation purposes, a fuel conversion program, supported by U.S. Department of Energy, National Nuclear Safety Administration (NNSA), Office of Material Management and Minimization (NA-23) Reactor Conversion Program, is underway to develop a replacement core in which the UO{sub 2} contains only low-enriched uranium (LEU). The preliminary LEU core design has shown a much lower carbon/uranium atomic ratio and enrichment than in the HEU core design. To achieve performance similar to the HEU core, the prospective LEU core will contain a much higher density of UO{sub 2} particles. As a result, a higher level of radiation damage, particularly from fission fragments emerging from the UO{sub 2} particles, can be expected dependent on the UO{sub 2} particle size within the fuel. To support LEU fuel design, the present study investigates the effect of reactor radiations on the thermal conductivity of graphite-based dispersion fuel. Both neutron- and fission-radiation induced degradations of the fuel are modeled, and validated based on the available experimental data. In this study, the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel was analyzed. A semi-empirical method to model the relative thermal conductivity with reactor radiations was proposed and validated. The proposed method can be further applied to predict the changes in the thermal conductivity of the prospective LEU TREAT fuel during long-term operation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

SU-F-BRE-11: Neutron Measurements Around the Varian TrueBeam Linac

Purpose: With the emergence of flattening filter free (FFF) photon beams, several authors have noted many advantages to their use. One such advantage is the decrease in neutron production by photonuclear reactions in the linac head. In the present work we investigate the reduction in neutrons from a Varian TrueBeam linac using the Nested Neutron Spectrometer (NNS, Detec). The neutron spectrum, total fluence and source strength were measured and compared for 10 MV with and without flattening filter and the effect of moderation by the room and maze was studied for the 15 MV beam. Methods: The NNS, similar to traditional Bonner sphere detectors but operated in current mode, was used to measure the neutron fluence and spectrum. The NNS was validated for use in high dose rate environments using Monte Carlo simulations and calibrated at NIST and NRC Canada. Measurements were performed at several positions within the treatment room and maze with the linac jaws closed to maximize neutron production. Results: The measurements showed a total fluence reduction between 35-40% in the room and maze when the flattening filter was removed. The neutron source strength Qn was calculated from in-room fluence measurements and was found to be 0.042 × 10{sup 2} n/Gy, 0.026 × 10{sup 2} n/Gy and 0.59 × 101{sup 2} n/Gy for the 10 MV, the 10 MV FFF and 15 MV beams, respectively. We measured ambient equivalent doses of 11 mSv/hr, 7 mSv/hr and 218 mSv/hr for the 10 MV, 10 MV FFF and 15 MV by the head. Conclusion: Our measurements revealed a decrease in total fluence, neutron source strength and equivalent dose of approximately 35-40% across the treatment room for the FFF compared to FF modes. This demonstrates, as expected, that the flattening filter is a major component of the neutron production for the TrueBeam. The authors greatly acknowledge support form the Canadian Nuclear Commission and the Natural Sciences and Engineering Research Council of Canada through the CREATE program. Co-authors Dubeau and Witharana are employees of Detec (Gatineau, Quebec), manufacturer of the Nested Neutron Spectrometer.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2019

Two Delayed Critical 7-Inch-Diameter Interacting Enriched (93.14) Uranium Metal Cylinders without Moderator and Reflector

This report documents very accurately the configuration and the materials for experiments with two unmoderated, unreflected, interacting, coaxial, highly enriched, 7-in.-diameter, uranium metal cylinders performed at the Oak Ridge Critical Experiments Facility (ORCEF) in April 1963 and described in logbook E-19 associated with experiments in the East cell of ORCEF. The information is sufficiently accurate that it can be used as the basis for preparation of benchmarks for International Criticality Safety Benchmark Program (ICSBEP) at Idaho National laboratory. The thickness of the cylinders was varied and the spacing between them was adjusted to achieve a delayed critical configuration. The average enrichment of the uranium metal was 94.15 wt. % 235 U. The heights of the 7-in.-diameter, equal-height cylinders varied from 2 5/8 to 3 1/4 inches. All interacting cylinders were assembled coaxially with their flat faces parallel and their combined masses varied between 62 and 76 kilograms of HEU metal. The data from these six experiments described are judged to be acceptable for use as criticality safety benchmark experiments for the ICSBEP and EURATOM’s Nuclear Energy Agency nuclear criticality safety benchmark program, once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, it is expected that the uncertainties in k eff could be as low as ± 0.0002.

07 ISOTOPE AND RADIATION SOURCES↗