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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
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↗
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↗
DOE OSTI2024

ROADRUNNER MiniFuel Experiment: Irradiation Target Design and Sample Characterization

High-density uranium nitride (UN) is a fuel candidate for several advanced nuclear reactor designs currently under development. Because there are limited UN performance data relative to fuel fabrication impurity and density variation, an irradiation campaign has been developed as part of a collaborative effort among the University of Texas at San Antonio (UTSA), Westinghouse Electric Company, Oak Ridge National Laboratory (ORNL), and Los Alamos National Laboratory (LANL) under the Nuclear Science User Facilities program. This project, entitled ROADRUNNER, or Research On ADvancing the peRformance of UraNium Nitrides in Extreme enviRonments, aimsto support UN fuel qualification for advanced reactors by investigating the impact of density and impurity variations on UN performance as a function of irradiation temperature and burnup. The MiniFuel experiment vehicle developed by ORNL, which leverages the High Flux Isotope Reactor, was selected to perform this accelerated separate-effects irradiation testing. The experiment test matrix consists of six MiniFuel targets containing miniature UN fuel disks, and targets three distinct burnup levels (37.5, 60, and 75 MWd/kg U) and three distinct temperatures (600, 900, and 1200°C). Neutronics and thermal analyses were performed to determine the experimental parameters needed to meet the desired irradiation conditions and to predict the experiment components temperatures. UN pellets were fabricated at LANL with tightly controlled parameters to produce specimens with three distinct densities and three levels of carbon content. The pellets were then thinned down by UTSA to the experiment-required thickness. The pre-characterization of the specimens includes density measurements, carbon and oxygen contents, microstructure analysis, and x-ray computed tomography. The selected specimens will be assembled into the MiniFuel experiment, and the first ROADRUNNER MiniFuel targets are intended for HFIR insertion during the Fall of 2024. After irradiation, the targets will be shipped to ORNL’s hot cell facility for disassembly. The post-irradiation examination on the fuel specimens includes fission gas release measurements, visual inspection, fuel swelling measurements, gamma spectroscopy, and microstructure analysis. The data collected post-irradiation will be used to develop fuel performance models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluating the Performance Envelope of Axial Flow HEPA Filter Designs - 17466

Nuclear grade high efficiency particulate air (HEPA) filters are routinely employed as the final control element in confinement ventilation systems. Approximately 6000 nuclear grade HEPA filters are purchased each year for use in the Department of Energy National Nuclear Security Administration (DOE/NNSA) complex. The traditional HEPA filter utilizes a pleated filter medium of fibrous glass that has been produced in a manner analogous to manufacturing paper. Containment ventilation systems are designed to condition the air-stream (temperature, relative humidity, moisture, particle size distribution of aerosols, and aerosol concentration) to levels and conditions that fibrous glass HEPA filters can handle. Axial Flow filters can be purchased with physical separators or embossed medium to maintain proper pleat geometry. Separator-less designs use media embossed with a corrugation pattern to retain pleat geometry. At least some versions of the separator-less filter packs have demonstrated unexpected and rapid failure under slightly elevated temperature and humidity conditions. The effects of age on HEPA filter performance have been a concern for at least two decades. A combination of physical evaluation of aged filter media and a review of filter failure records resulted in Appendix C of the Nuclear Air Cleaning Handbook (NACH), Determination of HEPA Filter Life. Mississippi State University is conducting studies to more clearly identify the operating envelope of separator and separator-less designs of ASME AG-1 Section FC axial flow HEPA filters under a set of generic design basis conditions. A parallel study is also being conducted to evaluate the effects of age on fibrous glass HEPA filters. Activities described in this paper have been conducted under Institute for Clean Energy Technology (ICET's) NQA-1 nuclear quality assurance program approved by DOE-EM. Testing meets the requirements of EM Quality Assurance Program (QAP) EM-QA-001, Rev. 1, June 11, 2012 that includes the applicable requirements of NQA-1-2008/2009a and DOE Order 414.1D. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Demonstration of closed shell breeding of cesium ions with an electron beam ion source

An electron beam ion source serves as a charge breeder for the Californium Rare Isotope Breeder Upgrade (CARIBU) at the Argonne Tandem Linac Accelerator System (ATLAS). The source accepts radioactive beams of 1+ or 2+ ions and raises their charge state for post-acceleration by ATLAS. The efficiency of this process impacts the length of each experiment and, hence, the type and number of experiments that can be run in each program cycle. Recent efforts to improve the charge breeding efficiency for the 80 < A < 160 species typical of CARIBU have focused on utilizing the closed shell breeding technique. Here, with this technique, the electron beam energy is manipulated to take advantage of the large gap in ionization energies at shell closures and selectively populate a single charge state. Charge breeding studies with stable cesium ions have demonstrated an absolute efficiency of 72% for Cs 27+ and a total efficiency of 93%. The Cs 27+ efficiency result represents a factor of 3 improvement over the previous best charge breeding result of 23% for Cs 27+ .

Vondrasek, R. [Argonne National Laboratory (ANL), ↗
DOE OSTI2026

Alternative Method for Performing As-Built, As-Irradiated Neutronics Analyses

Neutronics analyses have been used to assist in the design, development, and qualification of miniature fuel (MiniFuel) experiments irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) since their conception almost a decade ago. In recent years, upon completion of a MiniFuel target’s irradiation, neutronics analyses that reflect the as-built, as-irradiated conditions of an experiment have also been performed to validate postirradiation examination (PIE) efforts. Several radiation transport and depletion/activation software wrapper programs have been developed to increase the fidelity of these calculations; at ORNL, a program called HFIRCON is the primary software method employed. However, the utilization of these wrapper programs comes at a cost. To produce accurate results, the radiation transport portion of these analyses requires significant time and resources to execute. Therefore, this report discusses an alternative methodology for calculating some of the desired output for PIE validation—specifically, the output used in evaluating the experiment’s fuel burnup—without complete reliance on ORNL’s wrapper program. The alternative method is not a new software but rather a procedure for utilizing the depletion/activation software, SCALE, in conjunction with limited input data and reasonable assumptions to bypass most of the radiation transport calculations and thereby reduce the time required for the completion of a standard analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

SU-E-CAMPUS-J-03: Commissioning of the On-Board Cone-Beam CT System Equipped On the Rotating Gantry of a Proton Therapy System

Purpose: Proton therapy requires highly-precise image guidance in patient setup to ensure accurate dose delivery. Cone-beam CT (CBCT) is expected to play an important role to reduce uncertainties in patient setup. Hokkaido University has developed a new proton therapy system dedicated to spot-scanning under a collaborative work with Hitachi Ltd. In our system, an orthogonal X-ray imaging system is mounted on a full-rotating gantry. On-board CBCT imaging is therefore available. We have conducted commissioning of the CBCT system for clinical use in proton therapy. Methods: The orthogonal X-ray imaging system, which consists of two sets of X-ray tubes and flat panel detectors (FPDs), are equipped on the rotating gantry. The FPDs are mounted on the proton beam nozzle and can be retracted when not in use. The distance between the X-ray source and the FPD is about 2.1 m. The maximum rotation speed of the gantry is 1 rpm, so CBCT images can be acquired in approximately 1 minute. The maximum reconstruction volume is nearly 40 cm in diameter and 20 cm in axial length. For commissioning of the CBCT system, mechanical accuracy of the rotating gantry first was evaluated. Imaging performance was examined via quantitative evaluation of image quality. Results: Through the mechanical test, the isocentricity of the gantry was confirmed to be less than 1 mm. Moreover, it was improved to 0.5 mm with an appropriate correction. The accurate rotation of the gantry contributes to the CBCT image quality. In the image quality test, objects with 7 line-pairs per cm, which corresponds to a line spacing of 0.071 cm, could be discerned. Spatial linearity and uniformity were also sufficient. Conclusion: Clinical commissioning of the on-board CBCT system for proton therapy was conducted, and CBCT images with sufficient quality were successfully obtained. This research was supported by the Cabinet Office, Government of Japan and the Japan Society for the Promotion of Science (JSPS) through the Funding Program for World-Leading Innovative R and D on Science and Technology (FIRST Program), initiated by the Council for Science and Technology Policy (CSTP)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2026

Nuclear Data Libraries Sensitivity Studies for ORSA Using SCALE

Subcritical assemblies offer valuable training capabilities in nuclear criticality safety (NCS) for individuals handling fissile material. At Oak Ridge National Laboratory(ORNL), the Oak Ridge Subcritical Assembly (ORSA), a new experimental facility, is being established to provide hands-on training for the Nuclear Criticality Safety Program (NCSP).It is essential to accurately determine the neutron multiplication factor (keff) to ensure that ORSA remains subcritical and safe during operations. This study investigated the sensitivity of keff to variations across Evaluated Nuclear Data File (ENDF/B) libraries, consisting of ENDF/B-VII.1, ENDF/B-VIII.0, and ENDF/B-VIII.1. The analysis was conducted using the CSAS6 sequence in the SCALE-6.3 code system. Individual isotopes in the ORSA model were replaced one at a time with ENDF/B-VII.1 as the base library and changing to ENDF/B-VIII.0or ENDF/B-VIII.1. The results demonstrated that the changes in keffof the nuclides associated with the ORSA model were mostly within the uncertainty of the base model (~24 pcm), except for primary nuclides like Uranium-235and H-poly with few other nuclides. The relative delta keff values of Uranium-235 and H-poly, expressed in pcm, were +280 and -212 in ENDF/B-VIII.0 and +332 and -275 in ENDF/B-VIII.1, respectively, which were notable changes in reactivity. These results demonstrated that ORSA was largely insensitive to variations across these nuclear data libraries.

Hong, Evan [North Carolina State University]↗
DOE OSTI2022

Irradiation of MiniFuel Targets Bearing TRISO Fuel Compacts (Status Report)

Irradiation testing of MiniFuel compacts bearing tristructural isotropic (TRISO) fuel particles was performed at Oak Ridge National Laboratory (ORNL) to support the development of Kairos Power’s (KP’s) fluoride salt–cooled high-temperature reactor concept. The fuel compacts were fabricated with TRISO fuel particles of different types—including low-enriched uranium oxide, uranium carbide (LEUCO), natural uranium oxide, uranium carbide (NUCO), and low-enriched uranium dioxide (LEUO 2 )—and inserted into MiniFuel irradiation targets. Five targets were assembled and inserted in the High Flux Isotope Reactor (HFIR) for four cycles. The data collected post-irradiation will provide experimental input to validate TRISO fuel performance models for high particle power operations. This report summarizes the completion of the HFIR irradiation, the as-irradiated numerical analysis, and the post-irradiation work performed to date. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Diversion Path Analysis: A Proposed Methodology to Develop an MC&A Approach for Liquid-Fueled Molten Salt Reactors

Nuclear material control and accounting (MC&A) is a critical element of both the US Nuclear Regulatory Commission (NRC) and US Department of Energy (DOE)’s domestic safeguards and security requirements. NRC licensees are required, under Title 10 of the Code of Federal Regulations (10 CFR) Part 74 to establish and maintain an MC&A program that captures and records the quantities and locations of special nuclear material (SNM) at the facility. Along with physical protection, MC&A is a key element of domestic nuclear material safeguards that enables the NRC to ensure that SNM is controlled and accounted for. SNM, per 10 CFR Part 74, refers to plutonium, 233 U, and uranium enriched in the isotope 233 U or 235 U, but does not include source material. Periodic physical inventories, coupled with material balance evaluations, are effective and demonstrated tools to account for and detect theft or diversion of SNM in facilities containing SNM in bulk material form (i.e., not in discrete, countable items). Historically in the United States, these types of facilities have included fuel fabrication, conversion, and enrichment facilities. In comparison, reactors have relied on item counting of assemblies and control of SNM while in containment (e.g., a sealed reactor pressure vessel) because, to date, reactor fuel has been in item form. In liquid-fueled molten salt reactors (MSRs), unlike traditional light water reactors (LWRs) or bulk facilities, bulk SNM quantities can change significantly during operation as a result of depletion and transmutation. This introduces challenges to the use of traditional periodic physical inventories and material balance evaluations to detect theft or diversion of SNM in reactors that use SNM in bulk material form. Liquid-fueled (i.e., salt-fueled) MSR facilities are MSRs that use SNM within a salt eutectic as the fuel. The SNM is in a bulk material form any time it is outside of fresh or spent fuel storage containers. Some examples of when SNM will be in bulk form in the facility are during addition of fuel to the reactor system, while fuel is circulating in operation, and while fuel is in a drain tank. Periodic physical inventories and material balance evaluations can likely be effectively applied to many portions of an MSR facility, including all areas where depletion and transmutation are not significantly changing the quantities of SNM within the control area. Within an MSR facility, this would include fresh fuel receipt and loading, waste streams that may contain SNM, irradiated fuel storage outside of the reactor core, and any irradiated fuel processing that may happen after SNM has been removed from the reactor. All of these process steps could rely on measurements of SNM quantities compared with documented inventories. Any discrepancies from predicted (i.e., book) inventories and measured inventories could be quantified as inventory differences, consistent with traditional MC&A guidance from the NRC (e.g., in NUREG-1065 Revision 2, NUREG-2159 Revision 1, and RG 5.29 Revision 2). Within the reactor system, additions and removals to the book inventory include depletion of the SNM (e.g., fission of 235 U), which complicates the use of physical inventories. SNM control, however, can also likely be effectively applied to detect theft of SNM throughout a liquid-fueled MSR facility. To complement these approaches, prior technical reports have identified that a diversion path analysis may be a useful, risk-informed, and performance-based tool to determine suitable elements of an MC&A approach for the reactor system within a liquid-fueled MSR facility.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗