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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Criticality Safety Aspects of Higher Fissile Gram Equivalent at the AMWTF - 19296

Operating since 2003, the Advanced Mixed Waste Treatment Facility (AMWTF) has used high-tech characterization equipment, robotics, and automated treatment processes to accomplish its mission. The facility was designed to treat contact handled waste up to 325 Fissile Gram Equivalent (FGE). As the lower fissile mass drums and boxes were treated and shipped for disposal, other high fissile mass containers needed to be treated and shipped. The facility's mission evolved to include treating containers with up to 3,000 FGE. The nuclear criticality safety group has worked diligently with operations to increase AMWTF's fissile mass limits. As operations change across the Department of Energy facility, the nuclear criticality safety groups must work with operations to meet the mission. How Fluor Idaho LLC is accomplishing this task with respect to criticality safety is the subject of this paper. The signature piece of equipment at AMWTF is the Supercompactor, a giant hydraulic ram capable of exerting 4 million pounds of force that can reduce a 55-gal drum to a 5-inch puck. By compacting waste, the Department of Energy can put more waste into the limited storage space of the Waste Isolation Pilot Plant. Compaction also reduces the number of trips required to transport waste to the Waste Isolation Pilot Plant. There are two box-lines with hydraulic manipulators, nondestructive assay systems for drums and boxes, along with real-time-radiography. The plant's operations are highly automated. As time passed, AMWTF was tasked to treat drums that were not treatable with current operations. Based on analysis and experience of the nuclear criticality safety personnel, we have changed the fissile limits while maintaining a critically safe environment for the facility. Credit was taken for form, geometry, and lack of moderation in the material being treated. As the need arose to process higher FGE containers, more controls were identified for the treatment process. Criticality safety evaluations were expanded to 1,000 FGE for operational flexibility. Those evaluations included the box-line, the hot maintenance cell, and conveyors. Later, criticality safety evaluations were written to allow tanks to be processed in a box-line, as well as the use of a port in the box-line. Criticality safety evaluations for the box-line, hot maintenance cell, and conveyor were expanded to 3,000 FGE. The criticality safety officer is onsite and actively involved in operations. By expanding the nuclear criticality safety operating parameters, the AMWTF will meet the 2008 Record of Decision in which the Department of Energy directed 8,700 cubic meters of waste to be treated at the facility. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Reactor Neutronics: Impact of Fissile Material

Here, given a wide variety of reactor designs and fuel types, it can be difficult to identify the underlying cause of basic performance differences such as flux level and enrichment requirement. In this paper, using solely the definitions of the core multiplication factor and core power, simple relations have been derived allowing estimates of the flux ratio and fissile material concentration ratio for any reactor concept when 235 U is replaced with 239 Pu or vice-versa. These relations are functions of the neutron non-leakage probability, and one only needs to know number of neutrons emitted per fission, and the fission cross-section ratio between the 235 U system and the 239 Pu system. It is found that for a reactor concept having significant leakage, the achievable flux level when using 239 Pu as fissile material can be up to 45% larger than when using 235 U as fissile material, and the required fissile concentration of 239 Pu is up to 48% lower than that of 235 U to achieve criticality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

The Future of Nuclear Archaeology: Reducing Legacy Risks of Weapons Fissile Material

This report describes the value proposition for a "nuclear archeological" technical capability and applications program, targeted at resolving uncertainties regarding fissile materials production and use. At its heart, this proposition is that we can never be sure that all fissile material is adequately secure without a clear idea of what "all" means, and that uncertainty in this matter carries risk. We argue that this proposition is as valid today, under emerging state and possible non-state nuclear threats, as it was in an immediate post-Cold-War context, and describe how nuclear archeological methods can be used to verify fissile materials declarations, or estimate and characterize historical fissile materials production independently of declarations.

Fissile material, plutonium, Highly Enriched Urani↗
DOE OSTI2020

Design Optimization of A generic Fissile Solution for Mo99 production using Electron Beam-based Neutron Generator using MCNP+CFD

Molybdenum-99 ( 99 Mo) is a medical isotope used in 80% of all medical imaging procedures today. However, 99 Mo is not directly administered to patients for imaging; its decay product, technetium-99m ( 99 mTc), is a metastable isotope with a half-life of ~6 hours. Such a short half-life does not allow for production, separation, and application to a patient before decaying further. Therefore, the solution is to produce the precursor to 99 mTc, 99 Mo. 99 Mo has a half-life of 66 hours, long enough to produce the material and send it to the desired location before it fully decays. Under National Nuclear Security Administration’s (NNSA) highly enriched uranium (HEU) minimization mission, the Material Management and Minimization (M3) office leads the Molybdenum-99 (Mo-99) program. A major goal of the Mo-99 program is to develop methods of producing 99 Mo using low enriched uranium (LEU) opposed to the traditionally used HEU domestically in the U.S. Internationally, LEU is the new standard for use in research reactors and isotope production facility. In this study, we introduce a generic fissile solution system for 99 Mo production facility meeting the design requirement of 1) LEU usage as fissile material and 2) subcritical fissile system coupled with a horizontal electron beam (E-beam) accelerator. The focus of this research is to optimize a fissile solution configuration for maximum 99 Mo production yield using MCNP model, and ultimately implement the optimized design with computational fluid dynamics (CFD) model to understand thermal hydraulic behavior and solution convection characteristic. The report consists of two parts: 1) isotope yield calculation and criticality analysis with a wide range of parametric test matrix, and 2) thermal hydraulic analysis for solution mixing and cooling assessment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

New Approach in Fissile Assay Using LSDS - 17076

A quantitative assay of isotopic fissile materials (U{sup 235}, Pu{sup 239}, Pu{sup 241}) was conducted at the Korea Atomic Energy Research Institute (KAERI), using a lead slowing down spectrometer (LSDS). Using a uranium rod and plutonium, the actual measurement of induced fission neutrons was performed in the RPI (Rensselaer Polytechnic Institute) LSDS facility. In the simulation, a content of U{sup 235}, Pu{sup 239}, and Pu{sup 241} was changed. The detected signal has a direct relationship with the mass of the existing fissile isotopes. The mathematical model for fissile assay was applied to the measured and simulated detection data. The assay results show that the isotopic content was consistent with 2 to 3% uncertainty for Pu{sup 239}. The simulation indicated that the assay result was affected by the self-shielding correction. Usually, in highly enriched plutonium, the correction plays an important role and good understanding of it is required in order to increase assay accuracy. The combination of LSDS technique measurement and simulation is a very powerful and direct way to analyze the isotopic fissile content. The information provided by such analysis is applicable to nuclear fuel cycle development and spent fuel management for safety and economics. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Content Addition and Testing for 2017 NRC Amendment for the Versa-Pac Type A Fissile Packaging - 18120

The Versa-Pac is a certified Type A, fissile packaging (USA NRC Certificate no. USA/9342/AF-96), often used as a waste package for one time use. Continuous content additions have improved the versatility of the package resulting in expanded uses for the nuclear industry. Principal design of the Versa-Pac packaging maintains the use of an inner liner positioned inside a 55-gallon (VP-55) or 110-gallon drum (VP-110), as used for decades in the transport and disposal of Type A fissile materials. The Versa-Pac is used directly or in conjunction with pails, pipe containers, poly bottles, and a variety of smaller containers, inserts and vessels. The Versa-Pac content developments assist the industry with site shutdown, decommissioning, and cleanup operations, allowing for removal, transport, and storage options for the residual waste materials. A 2017 NRC license amendment application adds new capabilities to the Versa-Pac that include a low 1.25 wt% {sup 235}U enriched uranium loading, small quantities of UF{sub 6} in the form of 1S and 2S cylinders, and a payload mass increase from 250 to 350 pounds for the Versa-Pac 55-gallon version (VP-55). The lower uranium enrichment loading of 1.25 wt% {sup 235}U allows for a higher uranium mass limit. The inclusion of 1S and 2S UF{sub 6} cylinders fills an industry need to remove and dispose of long-stored material during site cleanup activities. The content payload mass limit is increased for low contaminated or radioactive heavy metal contents, such as cut plant components and ingots. Regulatory certification of new contents is demonstrated by test, analysis, and a combination of these techniques to show that the design function of the Versa-Pac as a Type A fissile package remains unchanged. Specific UF{sub 6} criticality and thermal analysis upgrades are added to the design documentation to accommodate 1S and 2S UF{sub 6} cylinders. The parameters for criticality analyses are simplified, ensuring no credit for cylinder containment is required for HAC thus reducing the reliance on a single specific shoring device. An upgrade of the Versa-Pac thermal analysis demonstrates compliance with ANSI N14.1 UF{sub 6} cylinder thermal limits. A new drop test sequence for the VP-55 was performed, documenting the addition of the 350-pound payload and demonstrating the VP-55 maintains structural integrity during the NCT and HAC drop, crush and puncture events as defined by 10 CFR 71. The testing and analyses show no loss of the material occurs and the package maintains the material in a subcritical state. This paper traces the engineering and licensing efforts involved in expanding the Versa-Pac's versatility for the 2017 NRC amendment application; summarizing the new testing and results for the payload increase, the content addition criticality and thermal evaluations, and content description for the Versa-Pac Safety Analysis Report. The NRC license amendment is scheduled for Winter 2017. Testing and supporting analyses reported in the Versa-Pac Safety Analysis Report demonstrates the content developments met the performance requirements of 10 CFR 71 for a Type A fissile package. A revision of the NRC Certificate of Compliance is expected in Spring 2018. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Fissile material found in electrical junction box at nuclear fuel services, Inc. - 14569

On October 29, 2014 at Nuclear Fuel Services, Inc., while troubleshooting a failed heater control circuit, fissile material was found in a small electrical junction box that contained a Resistance Temperature Detector (RTD). The RTD was mounted inside a stainless steel thermowell located in a uranium recovery boildown loop. Over time, the end of the thermowell had become pitted allowing fissile solution to penetrate the thermowell and collect in the electrical junction box. Conduit connected the small electrical junction box to an unfavorable geometry electrical panel, providing a credible scenario where an unsafe volume of fissile liquid could potentially collect in an unfavorable geometry. This postulated upset scenario had not been analyzed by Nuclear Criticality Safety (NCS). Prior to restart of the operation, a corrective action was taken to replace the damaged thermowell. Additionally, a review of other fissile systems was completed to identify all unfavorable geometry electrical panels where the potential for a similar accident scenario existed. Drain holes were drilled in the bottom of all those identified. Analysis of the damaged thermowell is ongoing. Expanded details of the incident, the short term and planned long term corrective actions, and lessons learned are discussed. (authors)

42 ENGINEERING↗
DOE OSTI2019

Exploiting fission chain reaction dynamics to image fissile materials

Radiation imaging is one potential method to verify nuclear weapons dismantlement. In this work, we present a method to discriminate between non-multiplying and multiplying neutron sources using a neutron coded aperture imaging system. This method applies time-correlated pulse-height (TCPH) analysis to identify neutrons in fission chain reactions and recreate images using only these events. This analysis was applied to measurements of weapons-grade plutonium (WGPu) metal and 252 Cf performed at the Nevada National Security Site (NNSS) Device Assembly Facility in July 2015. The results demonstrate it is possible to eliminate the non-fissile 252 Cf source from the image while preserving the fissile WGPu source. TCPH analysis was also applied to additional scenes in which the WGPu and 252 Cf sources were measured individually. The results of these separate measurements further demonstrate the ability to remove the non-fissile 252 Cf source and retain the fissile WGPu source. Simulations performed using MCNPX-PoliMi indicate that in a one hour measurement, hollow WGPu spheres are retained at a 1σ level for neutron multiplications M ≃ 2:7 and above. This work has potential application to dismantlement verification where it may be valuable to verify that all neutron sources in an image are multiplying.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Initial Assessment for a Generic Horizontal Fissile Solution Vessel with an Electron Beam-based Neutron Generator

The LANL NA-23 program office has provided an extensive design study for a medical isotope (i.e. Molybdenum-99) production facility based on a fissile solution vessel with various accelerator technologies. These concept designs and associated R&D have provided affordable pathways for system optimization for industry counter partner such as SHINE and Global Medial Isotope Supply (GMIS). A horizontal fissile solution vessel design with an electron beam (E-beam) accelerator is proposed and studied with modeling capabilities developed for previous concepts (i.e. a generic vertical fissile solution vessel with a deuterium-tritium (DT) accelerator). An initial system assessment is performed by estimating theoretical Molybdenum-99 production yield with MCNP6.2. A baseline system configuration is constructed based on early design discussions between LANL and GMIS. Since only an initial design concept is available at this stage, all associated system design parameters can be adjusted to fit the best system performance characteristics. In this work, unlike previous vertical system studies, the cooling configuration of the horizontal fissile solution vessel is not considered, and only fission reactions are used to quantify the theoretical product yield. Five parametric design studies are proposed, preliminary results are briefly reported, and extended design studies beyond what is proposed here are also discussed. It is concluded that a comprehensive system analysis via multiphysics coupled calculation developed by LANL should be further conducted for the final design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Fissile Material Stockpile Declarations and Cooperative Nuclear Archaeology

With the beginning of the Cold War, the United States and the Soviet Union quickly began to produce fissile material (plutonium and highly enriched uranium, HEU) for military purposes on an industrial scale. By the mid-1950s, both countries were already making ton-quantities of fissile material per year to supply their growing nuclear arsenals. They were soon joined by the United Kingdom (1951), France (1955), China (1964), and Israel (1965)—and later by India, Pakistan, and finally North Korea. By the time most of these production efforts ended in the early 1990s, the global stockpile of military fissile material had reached fantastic levels. While the combined global nuclear weapon inventory had peaked at about 64,000 in 1986,1 the military fissile-material stockpile ultimately reached almost 2200 tons (240 tons of plutonium and 1960 tons of highly enriched uranium), which is sufficient to make about 200,000 simple nuclear weapons.2 Since the end of the Cold War, almost 700 tons of highly enriched uranium have been down-blended and other material is now reserved for naval fuel. Overall, the military material available for weapon purposes today is sufficient for more than 150,000 weapons (Figure 1).

Glaser, Alexander↗
DOE OSTI2020

Accelerating fissile material detection with a neutron source

A system discriminating fissile material from nonfissile material wherein a digital data acquisition unit collects data at high rate, and processes large volumes of data directly to count neutrons from the unknown source and detect excess grouped neutrons to identify fission. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and inducing neutron emission therefrom, and a DC power supply that exhibits electrical ripple of less than one part per million. A neutron count histogram and Poisson count distribution are overlaid to provide a visual indication of the difference in correlation of natural and induced emitted neutrons from the radiation source to characterize the neutron source as fissile material or non-fissile material.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

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

Optimizing a detection system for fissile material in nuclear disarmament verification

In arms control treaties, verification plays a crucial role in detecting non-compliance, deterring future violations, and building trust between state parties. Neutron interrogation that induces fission reactions in fissile isotopes and measures the resulting fission neutrons, could be employed for this purpose. This study aims to develop a system which can determine the presence of fissile material while intrinsically protecting information. In this paper, we focus on optimizing the system for discriminating between an enriched uranium block from a depleted uranium (DU) block. The system was built and we report on benchmark measurements with DU and 16% enriched uranium blocks. Furthermore, the Excalibur (Experiment for Calibration with Uranium) neutron source, a neutron spectrometer (redBubble Technology Industries (BTI) N-Probe), and superheated droplet detectors were used for these measurements. MCNP simulations provided insights into detector responses to fissile materials with varying isotopic compositions, confirming that the system functioned as designed.

Active neutron interrogation↗
DOE OSTI2011

Fusion-fission hybrid for fissile fuel production without processing

Two scenarios are typically envisioned for thorium fuel cycles: “open” cycles based on irradiation of 232 Th and fission of 233 U in situ without reprocessing or “closed” cycles based on irradiation of 232 Th followed by reprocessing, and recycling of 233 U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in the critical reactors. This combination consumes about 20% of the thorium initially loaded in the hybrid reactor (~200 GWd/tHM), partially during hybrid operation, but mostly during operation in the critical reactor. The plant support ratio is low compared to the one attainable using continuous fuel chemical reprocessing, which can yield a plant support ratio of about 20, but the resulting fuel cycle offers better proliferation resistance as fissile material is never separated from the other fuel components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument’s LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Demonstration of Defense Waste Processing Facility (DWPF) Higher Fissile Content Glass

The Accelerated Basin De-inventory (ABD) Program has been proposed as an alternative for future spent nuclear fuel (SNF) and nuclear material processing at the Savannah River Site (SRS). This approach would change the baseline H-Canyon (HCAN), Concentrate, Storage, and Transfer Facility (CSTF), and Defense Waste Processing Facility (DWPF) operations. The ABD Program would require that all domestic and foreign research reactor SNF currently at SRS be dissolved, stored, and then transferred to CSTF without the recovery of uranium. Preliminary assessments in the ABD Program plan have shown that ~5000 extra SRS high-level waste (HLW) canisters would be produced if the fissile mass loading remains at the current 897 g/m 3 limit; however, increasing the limit to 2500 g/m 3 would result in ~520 extra canisters. Thus, the ABD Program plan requires an increase of the DWPF fissile mass loading limit to 2500 g/m 3 to minimize canister production. DWPF considers the following isotopes in the calculation of fissile mass loading: U-233, U-235, Pu-239 and Pu-241.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Evaluation Of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

Per a directive from the Department of Energy Savannah River Operations Office (DOE-SR) in 2008, the fissile mass loading concentration must remain below 897 g/m 3 in each high-level waste (HLW) glass canister produced by the Defense Waste Processing Facility (DWPF). To support Sludge Batch 5 (SB5) processing, the Savannah River National Laboratory (SRNL) developed a technical basis that facilitates the evaluation of fissile mass loading of the glass product. The calculation is based on the iron (Fe) concentration in the glass as determined by measurements from the Slurry Mix Evaporator acceptability analysis as well as the glass density. In April 2022, a subsequent DOE-SR directive increased the fissile mass loading limit to 2500 g/m 3 beginning with Sludge Batch 11. Thus, the 897 g/m 3 limit still applies to Sludge Batch 10 (SB10) processing. For SB5 through initial Sludge Batch 9 (SB9) processing prior to coupled operation with the Salt Waste Processing Facility (SWPF), SRNL provided DWPF a bounding glass density value that was based on a statistical evaluation of density measurements. To eliminate the need for experimental work, a composition-based density model for HLW glasses was developed at SRNL in 2019. The objective of this report is to present the bounding glass density determined with the composition-based density model for SB10 sludge-only (SO) and coupled processing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Evaluation of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

Per a directive from the Department of Energy Savannah River Operations Office (DOE-SR) in 2008, the fissile mass loading concentration must remain below 897 g/m 3 in each high-level waste (HLW) glass canister produced by the Defense Waste Processing Facility (DWPF). To support Sludge Batch 5 (SB5) processing, the Savannah River National Laboratory (SRNL) developed a technical basis that facilitates the evaluation of fissile mass loading of the glass product. The calculation is based on the iron (Fe) concentration in the glass as determined by measurements from the Slurry Mix Evaporator acceptability analysis as well as the glass density. In April 2022, a subsequent DOE-SR directive increased the fissile mass loading limit to 2500 g/m 3 beginning with Sludge Batch 11. Thus, the 897 g/m 3 limit still applies to Sludge Batch 10 (SB10) processing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗