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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Solidification of SRPPF Aqueous Recovery Liquid: Process Disruptions

Savannah River National Laboratory has identified a magnesium oxysulfate grout formulation to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). The formulation uses reactive, light burnt MgO, anhydrous MgSO 4 , and dead burnt MgO and provides good mixability, similar density to the original Portland Cement-based mix, and a leachate pH of 9.39, within the assumed Waste Isolation Pilot Plant (WIPP) brine pH range. The formulation will solidify the effluent using a solid first, lost paddle design, within a 55-gallon drum. A series of seven different process disruptions, or irregular processing during the expected ARS liquid effluent solidification process, were completed to assess the severity of the impact of the process disruption and how the formulation behaves under these conditions. Seven process disruptions were investigated including: No Mixing, Partial Mixing, Compacted/Stratified, Sealed, Overfilled, Underfilled, and Undermixed. The impact on solidification, leachate pH, Er (a Pu/Am surrogate) distribution, and density of the resulting grout were established. Overall, four of the process disruptions, No Mixing, Partial Mixing, Compacted/Stratified, and Overfilled, resulted in basic, unabsorbed liquid that would require additional solidification to be acceptable for WIPP. Most solids generated in this testing met the expected WIPP leachate pH requirements. Solids from the Compacted/Stratified process disruption test, however, produced a leachate pH slightly above the assumed range in the WIPP performance assessment. The distribution of Er for most of the tests was relatively consistent. The No Mixing process disruption test, however, resulted in non-homogeneous Er distribution, indicating that radioactive material would not be evenly distributed throughout this solid’s matrix. The Overfilled test is the only mix with a density slightly below the acceptable value. The Sealed and Overmixed tests produced a solid closest to the standard mix with no unabsorbed liquid, expected leachate pH levels, consistent Er concentrations, and expected density, indicating these process disruptions do not have a negative impact on the final grout form. Based on the other process disruptions, it is critical to have premixed dry materials and a well-defined mixing process to produce a solid, homogeneous magnesium oxysulfate cement, with no unabsorbed liquid. These process disruption tests show that the lack of adequate mixing is fundamental to creating an improperly solidified material that may require remediation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Uncertainty Quantification in GADRAS Inverse Modeling

The Gamma Detector Response and Analysis Software (GADRAS) package includes an inverse modeling tool that is helpful in identifying characteristics of unknown radioactive materials. Traditionally, uncertainties in this analysis were derived solely from measurement data quality and the fit of synthetic spectra. This paper aims to rigorously quantify additional sources of uncertainty, focusing on uncertainties arising from measurements being analyzed, Detector Response Function (DRF) characterization, and DRF extrapolation. Applying these findings to the BeRPBall benchmark data set, we demonstrated the impact of these uncertainties on plutonium and polyethylene estimates. The results underscore the importance of incorporating diverse uncertainty sources to enhance the accuracy and reliability of GADRAS’s inverse modeling capabilities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Development of an L-Edge X-ray Absorbance Spectrometer for Monitoring Dissolver Solutions in H-Canyon

Savannah River National Laboratory has developed a monitor to measure plutonium and uranium concentrations in solutions of dissolved nuclear fuel. The monitor will be installed in the sample aisle location for the 6.3D Dissolver in the Savannah River Site’s H-Canyon and used in support of the electrolytic dissolution such as Fast Critical Assembly fuel. The monitor is based on the atomic absorbance of x-rays. Elements are differentiated by the appearance of absorbance features at specific energies of the x-ray spectrum that correspond to L-edge transitions of inner core electrons. Hence, the technique is called L-Edge X-Ray Absorbance Spectroscopy (L-XRAS). The technique is suitable for nuclear fuel processing due to its relative insensitivity to other components of the dissolver solution, such as nitric acid, transition metals (Fe, Cr, Ni, Mn) such as those from stainless steel, particulates, and catalysts and additives. The instrumentation consists of a commercially available x-ray source and detector, a sample cell designed to interface with the airlift sampler associated with H-Canyon Tank 6.3D, and a stainless steel enclosure. SRNL wrote instrument control software and developed chemometric models to interpret x-ray intensity spectra and estimate analyte concentrations and uncertainties in real time.

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

Effect of Anoxic Iron Corrosion on WIPP Brine Geochemistry FY23 Final Report (U)

A 280-day study was completed to evaluate the effect of zero-valent iron (Fe 0 ) on the Waste Isolation Pilot Plant (WIPP) brine geochemistry under anticipated reducing conditions. Hydrogen (H 2 ) gas is expected to be present in the repository after closure due to the anoxic corrosion of a vast quantity of iron contained in the waste forms disposed at WIPP; therefore, a background argon atmosphere containing H 2 was chosen for this study. WIPP groundwater brine pH and E h will impact the mobility and fate of plutonium within the repository. Modeling and laboratory results for Castile WIPP brine indicate that equilibrium E h values relative to the standard hydrogen electrode (SHE) are 40 mV more reducing (i.e., more negative) than those for Salado WIPP brine (-480 mV vs. -440 mV, respectively) because of the higher pH of the Castile brine (pH 9.3 for Castile vs. pH 8.8 for Salado). The E h and pH data were corrected for the effects of high ionic strength. The experimental results for both brines are consistent with thermodynamic predictions using OLI Systems' Mixed Solvent Electrolyte chemical equilibrium model. The measured and corrected pH and E h data from this study are provided in Table ES-1 and Table ES-2, respectively.

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

SAVY O-ring characterization for lifetime extension

Los Alamos National Laboratory and Nuclear Filter Technology, Inc. developed the SAVY-4000 nuclear material storage container, named after the designers Stone, Anderson, Viers, and Yarbro. The SAVY 4000 meets the requirements of Department of Energy (DOE) M 441.1-1,2 Nuclear Material Packaging Manual, to protect workers who handle nuclear materials against radiation exposures resulting from the leakage of stored materials. The SAVY-4000 is an innovative and creative design: it can be opened and closed in a few seconds without using torque wrenches and instead only a tire iron or equivalent prying tool; has a built-in, fire-rated filter that prevents the build-up of hydrogen gas yet retains 99.97% of plutonium particulates at 200 ACCM rated flow; and prevents the release of stored materials upon a 12- foot drop.

42 ENGINEERING↗
DOE OSTI2024

Effect of Anoxic Iron Corrosion on WIPP Brine Geochemistry FY23 Final Report (U)

A 280-day study was completed to evaluate the effect of zero-valent iron (Fe 0 ) on the Waste Isolation Pilot Plant (WIPP) brine geochemistry under anticipated reducing conditions. Hydrogen (H 2 ) gas is expected to be present in the repository after closure due to the anoxic corrosion of a vast quantity of iron contained in the waste forms disposed at WIPP; therefore, a background argon atmosphere containing H 2 was chosen for this study. WIPP groundwater brine pH and E h will impact the mobility and fate of plutonium within the repository. Modeling and laboratory results for Castile WIPP brine indicate that equilibrium fa values relative to the standard hydrogen electrode (SHE) are 40 mV more reducing (i.e., more negative) than those for Salado WIPP brine (-480 mV vs. -440 mV, respectively) because of the higher pH of the Castile brine (pH 9 .3 for Castile vs. pH 8.8 for Salado). The E h and pH data were corrected for the effects of high ionic strength. The experimental results for both brines are consistent with thermodynamic predictions using OLI Systems' Mixed Solvent Electrolyte chemical equilibrium model. The measured and corrected pH and E h data from this study are provided in Table ES-I and Table ES-2, respectively. The experimental study, with four test conditions in triplicate, was performed in a dual glovebox with a nominally 3 vol.% H 2 in argon atmosphere (target H 2 range: 3 ± I vol.%). Simulants containing MgO only ( experimental control) and MgO+Fe 0 (WIPP base case) were prepared for both the Salado and Castile brines. MgO was included in all simulants to account for the use of bulk magnesium oxide in the WIPP repository. Fe 0 was included in some simulants to incorporate the effects of the anoxic corrosion of iron and in-situ hydrogen generation in the study. The brine compositions were developed by Sandia National Laboratory (SNL; Xiong, 2008) and have been used in previous WIPP evaluations. The test method (agitation, etc.) is partially based on ASTM D3987-12. Twelve rounds of periodic measurements of pH and E h were performed over the course of the study. Chemical analysis results for liquids and solids (ICP-MS, ICP-ES, IC Anion, TIC, SEM-EDX) are consistent with the pH, E h , and thermodynamic modeling results. This study included the following conditions that deviate from anticipated post-closure conditions following brine intrusion, but were selected to facilitate bench-scale testing to validate modeling of pH and E h for the post-closure WIP P repository: an anoxic glove box atmosphere containing ≤ 4 vol. % H 2 vs. substantially higher H 2 gas concentrations assumed in the WIPP Performance Assessment (PA); a significantly higher liquid-to-solid test ratio compared to the much lower phase ratio anticipated in the WIP P repository; agitation of the simulant bottles to maximize mass transfer; and finally the use of Fe 0 reagents having a much greater surface area than expected in the WIP P repository. Non-representative conditions were chosen for various reasons such as: to provide bounding conservative results, to provide a margin of safety for testing, or to facilitate simulant sub-sampling and analysis. In a parallel effort, aqueous electrolyte thermodynamic models were developed for the synthetic Salado and Castile brines to inform the experimental design, facilitate laboratory data interpretation, and allow extension of evaluations beyond the parameters tested. Thermodynamic modeling simulations including the MgO and Fe 0 additives that are directly relevant to the experimental measurements (e.g., pH calibration curve, ORP corrections) are included in this report. The measured fa of the simulants was close to the OLI model predictions for both brines and was largely controlled by the background H 2 partial pressure in the vapor phase as well as H 2 generated in situ in the aqueous phase by the Fe 0 corrosion. The H 2 gas-phase concentration tested and thermodynamically evaluated was much lower than is assumed in the WIPP PA; however, H 2 (g) concentrations significantly below this level are still predicted to result in very reducing conditions. In conclusion: • The experimental results are consistent with thermodynamic model predictions for fa, pH, and the effects of high ionic strength. • Evidence to date suggests that the H2 concentration in the glovebox atmosphere ultimately determined the final E h values of the simulants and resulted in highly reducing conditions. As a result, little difference was observed between the control simulants containing only MgO and the WIPP base-case simulants that contained MgO and Fe 0 . • This test methodology is recommended for future studies evaluating WIPP repository conditions. The methodology includes: (1) background H 2 in argon with agitation ( or could alternatively include in-situ-generated H 2 in sealed bottles); (2) carefully measured and corrected ORP data ( with much effort focused on allowing the probes to fully stabilize); and (3) ionic-strength-corrected pH data. Other best practices, such as simulant sparging/handling, ORP probe replacement, etc., should also be considered. • The coupling of experimental studies and thermodynamic modeling is also highly recommended because these methods inform and direct one another leading to greater confidence in and understanding of the results.

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

Microscopic Measurement: Need for Nuclear Material Sample Specimen Characterization Capability

The Surveillance, Fabrication, and Off-Site Operations team at Lawrence Livermore National Laboratory is beginning an update to the nuclear material machining lab in Superblock. Plutonium and other nuclear materials tend to accelerate the aging process for equipment and machines that would otherwise be maintainable for long lives. They have their lifespans shortened from deteriorated seals, embrittled polymer components, accelerated corrosion, and damage to electronics from exposure to radioactive materials. The cost of installation of equipment into a glove box (GB) tends to outstrip the cost of the equipment itself. Due to the magnification of cost, the historical approach to equipment installs has been to purchase and install the highest-quality and most robust equipment available to maximize time between failures and updates. To make room for the installation of a new lathe, the inspection GB is being moved to a different location within the machining lab. As part of the overall effort, the equipment within the inspection GB is being revisited. Current measurement capabilities are antiquated and starting to become problematic due to age. Advances in metrology are sought for more comprehensive characterization of scientific samples. This document aims to identify several options for the replacement of current measurement equipment and identify a system for purchase and eventual installation. The intention is not to select the cheapest system, but rather to identify a system that best meets the needs of both the operators and principal investigators (PI). Following the convention of the SYSM-5620 Design Thinking and System Engineering, it was determined that a laser confocal measurement system from Keyence is the most suitable option for our programmatic need. Form factor, capability, and maintainability were considered. Along with stakeholder input, to ensure the best decision for the Nuclear Material Technology Program.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Full Submersion Water Testing of SAVY-4000 Nuclear Material Storage Containers

SAVY-4000 (SAVY) containers are the primary container used at Technical Area (TA) TA-55 plutonium facility for the prevention of water ingress to mitigate against a criticality event. The basis for water resistance of these containers has long been attributed to the Polytetrafluoroethylene (PTFE) membrane that is assembled on the outermost surface of the filter assembly. On August 22nd a test of a container used inside of a glovebox was performed that brought into question this long-standing-basis. This test was performed by inverting a SAVY-lid onto a specialized piece of equipment for evaluating the integrity of the PTFE membrane. During the test water was observed passing through the filter indicating that the filter membrane was no longer preventing the ingress of water through the filter. The assumption after making this observation was that alpha-particles were rapidly degrading the membrane creating a leak path through the underlying aluminosilicate media (Fiberfrax®). The apparatus used was designed to only test the lid rather than the entire SAVY assembly. A test plan, PA-PLAN-01921, was developed to investigate whether a fully assembled container with a fully degraded PTFE membrane would be capable of meeting the criteria defined in PA-RD-1009 of not allowing more than 200 ml of water to enter the container with a water column of 6-inches applied over a 2-hour period. This report provides the results of the testing performed against PA-PLAN-01921.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Performance Characterization of FB-Line Neutron Multiplicity Counter and Large Neutron Multiplicity Counter

Savanah River National Laboratory’s (SRNL) Nuclear Measurements group was tasked with characterizing the performance of two neutron multiplicity counters located at SRNL. Characterization measurements were made to determine the gate width, pre-delay, deadtime parameters, triples and doubles gate fractions, detector efficiency, and operating high voltage for the Large Neutron Multiplicity Counter (LNMC) and the FB Line Neutron Multiplicity Counter (FBLNMC). The parameters were determined, shown below, and were, as to be expected, slightly different than the previous calibrations, which were performed over 20 years ago. Several Pu samples were measured to validate the characterizations of the FBLNMC and LNMC. The measurements determined the sample Pu-240 mass within <2% deviation for the pure plutonium samples and ~8% for the mixed oxide sample. The pure Pu samples had significantly better accuracy compared with the impure mixed oxide sample due to the lack of induced fission or alpha,n neutrons from impurities. Overall, the characterization of the neutron multiplicity counters, and the determination of their operability has been completed successfully.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Trace Element Analyses of Micron-Size Particles and Statistical Determination of Minimum Detection Limits

Savannah River National Laboratory (SRNL) has developed expertise in producing homogeneous, ca. 1 m-diameter spherical particles of mixed-element components, wherein dopants can be varied from a trace constituent (ppm) to wt.% concentrations. The samples used for this work are nickel-doped cerium oxide microspheres produced by SRNL. They were initially selected as analogs for plutonium-doped uranium oxide particles and analyzed as part of a larger study to evaluate whether electron probe microanalyzers (EPMA) can be used to characterize nuclear materials as an alternative or complementary method to mass spectrometers. The five samples used in this study contained nominal compositions of 0, 0.004, 0.04, 0.4 and 4 wt.% Ni. They were analyzed by both an Agilent 7900 Q-ICP-MS at SRNL and the JEOL JXA&#x2;8530F Plus EPMA at the University of Minnesota. In addition to EPMA results (calibrated with high-precision Q-ICP-MS analyses) suggesting that the EPMA could address outstanding nuclear material characterization needs, these samples 1) showcase the ability of the EPMA to quantify not just trace concentrations, but trace concentrations in microparticles (1 m diameter, Fig. 1), and 2) offer a unique opportunity to evaluate the methodology for assessing the minimum detection limits of EPMA analyses.

McSwiggen, Peter [JEOL USA, 11 Dearborn Road, Peab↗
DOE OSTI2025

Foam and Unreacted Material Reduction in Magnesium Oxysulfate for SRPPF Waste Solidification

A magnesium oxysulfate (MOS) grout formulation has been proposed to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). This formulation uses a combination of light-burnt MgO, anhydrous MgSO 4 , and dead-burnt MgO resulting in a grout with good mixability and acceptable density. The setting time for the grout is within the operational limits of the SRPPF facility and the leachate pH is ~9.4, which is within the assumed pH range of the brine from the Waste Isolation Pilot Plant (WIPP). When first tested at the 1-gallon small-scale and 55-gallon full-scale, there was a significant foam layer that raised concerns of a nonhomogeneous final form. A nonhomogeneous mixture could have variable density throughout the waste form and therefore allow for unequal shielding. To reduce the potential for unequal shielding in the final form, a series of foam reduction tests were carried out and a final formulation of 28 wt% light burnt MgO, 10 wt% MgSO 4 , and 62 wt% dead burnt MgO was used to remove the foam in under two hours while still maintaining a mix that has an acceptable density, mix time, peak temperature, and set time.

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

Evaluating FRAM v.6.1 with LANL and Euratom data

The United States Department of Energy and the European Atomic Energy Community (EURATOM) collaborate to test, evaluate, and improve the performance of the FRAM 6.1 code with various nuclear materials, especially mixed oxide (MOX) containing an isotopic blend of uranium and very high burnup plutonium. This joint effort is intended to promote effective implementation of the FRAM 6.1 code into nuclear material verification methods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Effects of SRPPF ARS Effluent Variations on Magnesium Oxysulfate Solidification

A magnesium oxysulfate (MOS) grout formulation has been identified to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). The formulation uses reactive, light burnt MgO (28 wt%), anhydrous MgSO 4 (10 wt%), and dead burnt MgO (62 wt%), resulting in a grout with good mixability, acceptable density, acceptable setting times, and a leachate pH around 9.4, within the assumed Waste Isolation Pilot Plant (WIPP) brine pH range. A series of tests were then designed to observe how this formula reacted to variations in the liquid effluent. Liquid effluent temperature, the caustic pH level, concentration of NaNO 3 , and the presence of neutralization products, trace metals, and/or sodium sulfate (all potential residuals found in the effluent from upstream processing), were evaluated to see how each impacts the grout mixing time, peak temperature, setting time, presence of bleed water, leachate pH, and density. The standard mix has an average mixing time of 18 minutes, peak temperature of 100.5 °C, and set time of 60 minutes. Increasing the temperature of the liquid effluent increased the reaction rate of the mix and reduced mixing and setting times. Between 30 and 40°C the change was relatively small, raising no more than 6 °C compared to the standard mix maximum temperature. Increasing the pH of the liquid effluent had no significant impact on the mix. The addition of neutralization products and trace metals had an overall impact of decreasing the reactivity of the mix. The addition of more ions in the liquid effluent, such as an increase in the concentration of NaNO 3 and the addition of Na 2 SO 4 , generally decreased the reactivity of the mix. None of the variations to the liquid effluent hindered solidification as indicated by the lack of bleed water found on all the samples. The leachate pH values for all mixes tested did not significantly vary from the expected pH of 9.4 and none of the density values fell below 1.8 g/cm 3 . Overall, changes to the liquid effluent were found to have a minimal impact on the formulation suggesting the grout's ability to properly form despite increased temperatures and the presence of residuals typically found in the liquid effluent from the ARS.

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

NCERC 2024 Highlights

The National Nuclear Security Administration (NNSA) is entrusted with ensuring the safety, security, and reliability of the nation’s nuclear weapons stockpile while advancing programs aimed at reducing global nuclear proliferation. These critical mission objectives are achieved through the expertise of a highly skilled team of professionals. The operations at the National Criticality Experiments Research Center (NCERC) play a vital role in developing and enhancing knowledge and expertise in advanced nuclear technologies. NCERC supports a wide range of mission areas, including nuclear criticality safety, nuclear emergency response, and nuclear nonproliferation, safeguards, and arms control. It also provides support to the Department of Homeland Security, advances stockpile stewardship science, and delivers scientific expertise to other government agencies, such as NASA and the Defense Threat Reduction Agency. NCERC conducts experiments utilizing diverse nuclear materials, from small neutron-emitting sources for testing radiation detection equipment to larger quantities of uranium and plutonium for criticality experiments. A cornerstone of NCERC's mission portfolio includes the operation of four critical mass assembly machines—Planet, Comet, Flattop, and Godiva-IV—which are instrumental in advancing nuclear science and ensuring the nation’s nuclear security objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

"Isotopics by nuclides" tool performance in InterSpec

This report provides a performance summary of the “Isotopics by Nuclides” tool in InterSpec. The primary objective of the NA-241 FY25 project was the detection and characterization of non-homogeneous uranium samples. However, the tool also demonstrates effectiveness in determining the enrichment of homogeneous uranium or plutonium from single gamma spectra. This paper presents a subset of evaluated data to avoid distribution limitations while offering potential users’ insight into the tool's performance.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Parametric Analysis of Holdup in HEPA Filters

This technical document (TECH) contains an analysis of several conditions deemed credible for plutonium oxide buildup in high-efficiency air particulate (HEPA) filters to determine the effect of holdup inside the filter on reactivity. This document is performed in accordance with NCS-AP-005, R3 [Ref. (1)].

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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↗