POTENTIAL DISPOSITION OPTIONS FOR SURPLUS US DOE NUCLEAR MATERIALS IN SUPPORT OF DISPOSITION PLANNING
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ALMOND, PHILIP M.↗
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The relationship between certain characteristics of industrial research proposals and their subsequent disposition.
The ideas produced by three idea generation groups operating within a large industrial R & D laboratory were studied over a period of two years. Data were collected in order to relate idea dispositions by management to subjective evaluations provided by respondents representing the technical staff and to study the roles of urgency, predictability, and expected time horizon in the disposition decision. The findings are (1) a significant positive correlation was found between idea dispositions and associated subjective evaluations despite the fact that management did not know the subjective evaluations at the time the disposition decisions were made; (2) urgency was the dominant consideration in the disposition decisions, but predictability and expected time horizon were dominant in the subjective evaluations; (3) predictability and time horizon are covariants; and (4) management hedges by shelving marginal ideas. Implications for R & D management are drawn.
Baker, N. R.↗
Alternative Analysis and Prioritization of Department of Energy “TBD” Materials with No Identified Disposition Pathway
The Department of Energy (DOE) complex manages a significant inventory of excess nuclear materials for which disposition pathways have not been identified, commonly referred to as "To Be Determined" (TBD) items. The Disposition Pathways Program, initiated in FY2018, provides a standardized framework for identifying viable disposition pathways for these materials. In 2023, the program undertook a comprehensive review and systematic analysis of the remaining TBD material groups, updating the assessment from the 2020 TBD Study using the 2022 fiscal year-end Nuclear Material Inventory Assessment (NMIA) as the primary data source. This effort aimed to utilize quantitative analysis to down-select from a wide range of potential disposition options and prioritize the remaining pathways to facilitate informed, risk-based decision-making for future programmatic funding and execution. This paper details the alternative analysis methodology used for screening and prioritization, highlighting the key criteria, ranking process, and resultant recommendations. The study successfully narrowed fifty-two potential disposition options down to nineteen, providing a focused path forward for addressing a longstanding challenge within the DOE complex.
Ramsey, Catherine [Savannah River National Laborat↗
Dispositional logic
The applicability of conventional mathematical analysis (based on the combination of two-valued logic and probability theory) to problems in which human judgment, perception, or emotions play significant roles is considered theoretically. It is shown that dispositional logic, a branch of fuzzy logic, has particular relevance to the common-sense reasoning typical of human decision-making. The concepts of dispositionality and usuality are defined analytically, and a dispositional conjunctive rule and dispositional modus ponens are derived.
Le Balleur, J. C.↗
FY24 Task 5: Leachate Disposition
Directly feeding sludge solids to the high-level waste (HLW) Waste Treatment Plant represents an alternative flowsheet seeking to initiate sludge processing as soon as possible. Key processing functions previously captured during baseline pretreatment operations include leaching and washing prior to solids concentration. These operations should be considered in the potential direct feed flowsheets to maximize waste feed loading, minimize HLW volume, and mitigate corrosion challenges associated with vitrification of high phosphate and fluoride concentrations. Additionally, single-shell tank (SST) retrievals and waste transfers to double-shell tanks (DSTs) in a direct feed flowsheet would likely also benefit from some level of leaching, washing, and solids concentration in order to reduce DST space and mission duration. These operations could occur in a new facility or potentially in available DSTs. If washing and leaching are utilized, an effective disposition pathway for the wash water and leachate solutions are needed. Three target species that benefit significantly from leaching and washing are phosphate, fluoride and aluminum. Phosphate (PO 4 3- ) and fluoride (F - ) can contribute substantially to the amount of carrier fluid needed for dissolution, and the resulting volume of liquid generated. Disposition of this retrieval solution should be evaluated in order to prevent crystallization of these anions throughout system processing. Since there is a high probability that any retrieval solutions will be at or near their PO 4 3- and F - solubility limits, evaporation or blending with a high Na supernate (>3.5 M) is not recommended for the wash water streams without a method to remove precipitants prior to solution disposal. Additionally, aluminum present in the southeast quadrant of the Hanford site represents roughly 60% of the waste solids in the initial processing tanks. These aluminum solids are in the form of gibbsite (Al(OH) 3 ) and can pose significant challenges for processing due to the fast-settling times and high solids loading associated with these materials. Easily remediated by caustic addition to the solids, these wash solutions could be processed through crystalline silicotitanate (CST) ion exchange columns to prepare the supernate solutions for disposition. The current target for feed conditions to the Low Activity Waste (LAW) melter are waste streams that contain nominally 5-6 M Na. Fractions within the tanks contain upwards of 0.2 M phosphate and fluoride in solution at 3.5 M Na. Concentrating these solutions above 5 M Na would result in an exceedance of the solubility limits, and potential for uncontrolled precipitation of the phosphate and fluoride crystal material. The resulting crystalline salt material is typically sodium fluoride phosphate, also referred to as natrophosphate (Na 7 FPO 4 ·19H 2 O). Salt phases are of importance in tank waste due to their chemical reactivity, which can result in precipitation, dissolution, or transformation, impacting any downstream processes (Bolling et al. 2020, Russell, Snow, and Peterson 2010). Salt generation and precipitation could pose challenges by causing system plugging and melter corrosion if left in the supernate stream, or limit sodium molarity of the supernate that would be accepted without incident in waste operations. To understand the impact of this salt generation, the crystallization of natrophosphate in multiple simulant feed matrices was studied to understand the implications of various tank waste supernate chemistries. Three matrices were examined: high PO 4 3- /low F - , low PO 4 3- /high F - , and an average matrix. Subsequent testing was performed with the average matrix with the inclusion of CsNO 3 , and a final run with the average matrix including CsNO 3 and a 137 Cs spike for tracer purposes.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DCS: A Case Study of Identification of Knowledge and Disposition Gaps Using Principles of Continuous Risk Management
The Human Research Program (HRP) is formulated around the program architecture of Evidence-Risk-Gap-Task-Deliverable. Review of accumulated evidence forms the basis for identification of high priority risks to human health and performance in space exploration. Gaps in knowledge or disposition are identified for each risk, and a portfolio of research tasks is developed to fill them. Deliverables from the tasks inform the evidence base with the ultimate goal of defining the level of risk and reducing it to an acceptable level. A comprehensive framework for gap identification, focus, and metrics has been developed based on principles of continuous risk management and clinical care. Research towards knowledge gaps improves understanding of the likelihood, consequence or timeframe of the risk. Disposition gaps include development of standards or requirements for risk acceptance, development of countermeasures or technology to mitigate the risk, and yearly technology assessment related to watching developments related to the risk. Standard concepts from clinical care: prevention, diagnosis, treatment, monitoring, rehabilitation, and surveillance, can be used to focus gaps dealing with risk mitigation. The research plan for the new HRP Risk of Decompression Sickness (DCS) used the framework to identify one disposition gap related to establishment of a DCS standard for acceptable risk, two knowledge gaps related to DCS phenomenon and mission attributes, and three mitigation gaps focused on prediction, prevention, and new technology watch. These gaps were organized in this manner primarily based on target for closure and ease of organizing interim metrics so that gap status could be quantified. Additional considerations for the knowledge gaps were that one was highly design reference mission specific and the other gap was focused on DCS phenomenon.
Norcross, Jason↗
DEVELOPMENT OF AN APPROACH FOR DOWNBLENDING AND DISPOSITION OF IRRADIATED GRAPHITE FUEL
The IGR reactor, a pulse-type test experimental reactor, reached full power in 1961. The reactor contains a center axial experimental ampoule that is subjected to a large pulse of neutrons (neutron flux maximum density of 7×1016 n/cm2·s, peak power 10GW) when the reactor is operated. The fuel form for the IGR reactor is graphite cubes and rods impregnated with highly enriched (90%) uranyl-dinitrate (UO2(NO3)2·6H2O). This fuel is eligible for return to Russia for reprocessing under the Russian Research Reactor Fuel Return program. However due to some technical and licensing issues it was decided to immobilize this fuel in Kazakhstan. The disposition approach thus selected for the spent HEU fuel involves down-blending of the fuel followed by cementation. Development of the process for down-blending and subsequent cementation requires a process that satisfies many criteria, including fissile content, compressive strength, permeability, and containment of radioactivity, among others. One key goal of the final product is to meet the criteria for termination of safeguards. This paper will describe the overall challenge associated with disposition of irradiated HEU graphite spent fuel, the identification of the objectives of the disposition process, and the constraints associated with development of the process and the waste form.
Gaines, Kris↗
A Simulation Modeling Approach to Optimizing Nuclear Waste Dispositioning
The dispositioning of nuclear waste generated at facilities across the country is an ongoing battle that affects us all. National laboratories and research centers dealing in medical research, clean energy, and other nuclear activities such as the Department of Energy (DOE) facilities face the need to properly manage and dispose of nuclear waste. A dynamic modeling solution would enable the DOE and others to make decisions on waste disposal and technological options. In doing so, this research explores modeling techniques using available data to address these situations. The focus being on developing an initial robust and adaptable discrete event model using the ExtendSim tool. This modeling effort will target the dispositioning of transuranic waste at the Savannah River National Laboratory (SRNL) which can be expanded to represent the current state of disposition process for waste generated at other DOE facilities. The model aims to assess resource allocation and waste processing options to stabilize productivity and cut the backlog of nuclear waste. By assessing the results of different scenarios, this research aims to provide actionable insights for the DOE. This approach has the potential to significantly improve the management of radioactive waste, offering the capability of evaluating options for optimizing the process for nuclear waste disposal. The findings of this study can serve as a valuable resource for decision-makers and other national laboratories, or research entities engaged in nuclear operations by enabling them to make more informed choices.
Andaverde, Alexis↗
A Phase-Disposition PWM Enabled Model Predictive Control for a Nine-Level Inner-Interleaved Hybrid Multilevel Converter
This article brings forward a phase-disposition pulse width modulation enabled model predictive control (PDPWM-MPC) for a nine-level inner-interleaved hybrid multilevel converter (9L-IHMC). Firstly, three layers of virtual space vector diagrams (VSVD) are established based on the sign patterns of the original and virtual reference vectors in the abcframe, respectively, to achieve the phase-disposition pulse width modulation (PD-PWM) in a non-independent three-phase way. Then, three adjacent virtual vectors in the third-layer VSVD, together with their optimized duty cycles, are applied to guarantee the optimal current tracking. Finally, through the use of the duty cycle alternation approach, dc-link and floating capacitors voltages are balanced and circulating currents are mitigated as well. The proposed PDPWM-MPC can not only enable the decoupling of the low-and high-frequency stages in the 9L-IHMC, but also reduce both output current ripples and computational burden. In addition, it can achieve a constant equivalent switching frequency and address the disproportion of power losses associated with the PD-PWM. Here, both simulations and experiments on a silicon carbide device-based prototype substantiate the effectiveness of the proposed control strategy.
42 ENGINEERING↗
Disposition of Solids from Hydrothermal Liquefaction of Biomass: Current Understanding, Research Gaps, and Opportunities
Characteristics of solid residuals generated during hydrothermal liquefaction (HTL) of biomass are reported to establish disposition options for these residuals to reduce liability or barriers to commercialization of HTL. HTL solids from pilot scale are reported and compared to existing HTL solids literature for a variety of biomass feedstocks and HTL configurations, noting differences among literature HTL solids due to solids collection methods. Reported analyses include CHNOS, moisture, ash, inorganics by Inductively Coupled Plasma (ICP), and High Heating Value (HHV). HTL solids are determined to be substantially compatible with U.S. Environmental Protection Agency (EPA) disposal regulations within existing data. Disposition and valorization routes are identified, yet combustion as valorization is determined to be not economically beneficial. Further, research gaps are identified, which include the need for future research to be performed on HTL solids that are representative of a continuous, commercializable flowsheet; verification that specific HTL solids meet EPA disposal requirements; exploration of valorization options; and the need for invention of continuous HTL solids separations/management.
09 BIOMASS FUELS↗
Post-closure Nuclear Criticality Safety Evaluations for Disposition of Criticality Control Overpacks at the Waste Isolation Pilot Plant
The Waste Isolation Pilot Plant (WIPP) is a geological repository in southern New Mexico that provides for disposal of transuranic (TRU) wastes from atomic energy defense activities. The Sandia National Laboratories (Sandia) Report, Consideration of Nuclear Criticality When Disposing of Transuranic Waste at the Waste Isolation Pilot Plant, addresses nuclear criticality safety based on the projected inventory characteristics for the initial compliance certification application of WIPP in 1996. As the inventory, waste forms, and disposal package designs change, revised or new analyses are necessary to demonstrate acceptability for these configurations within the WIPP safety basis and compliance with 10,000-year post-closure standards of the US Environmental Protection Agency (EPA). Saylor and Scaglione evaluated criticality control overpacks (CCOs) in 2017 based on conservative assumptions for post-closure repository structural conditions with resulting effects on containers and container spacing, The Saylor and Scaglione evaluation of CCOs addressed a single waste configuration that represents the Surplus Plutonium Disposition Program’s dilute and dispose waste form and composition. This initial CCO study demonstrated that 50 grams of boron carbide (B 4 C) per CCO is sufficient to ensure post-closure criticality safety based on a well-mixed waste composition, and Oak Ridge National Laboratory (ORNL) subsequently determined that this amount of B 4 C does not require constraints on moisture or plastic present as moderator. The Saylor and Scaglione analysis conservatively assumes repository room closure that eliminates all space between fissile gram equivalent (FGE) 239 Pu masses. The close-packed array was selected based on limited availability of repository salt creep modeling results at that time. In 2019, Brickner provided additional evaluations for pipe overpack containers (POCs), building on the conservative basis provided by Saylor and Scaglione. Brickner’s 2019 analysis made use of new geomechanical data for post-closure spacing that rely on advances in repository modeling as documented in the work by Reedlunn and Bean. This current CCO evaluation for generic waste materials expands on earlier work performed at ORNL and includes evaluation of CCOs across a much broader range of possible waste compositions and geometries. This evaluation is intended to provide input for the required feature, event and process (FEP) screening to determine if post-closure criticality must be included as an event in the 10,000-year regulatory evaluation. As such, the approach to modeling post-closure criticality presented in this report has been coordinated with the Sandia team responsible for FEP screening. The resulting analysis supports disposition of fissile materials in the CCO containing up to 380 FGE 239 Pu and expands conditions acceptable for disposal of fissile material in CCOs. This evaluation builds on the methodology of Saylor and Scaglione and Brickner, using the most recently available geomechanical data for CCO spacing under salt creep compaction scenarios provided by Reedlunn and Bean. The broad range of fissile material configurations analyzed in this report are intended to account for configurations that may occur during the post-closure disposal time period, and it also includes waste configurations that are not physically possible to support analysis of conditions that influence neutron fluence.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Syllogistic reasoning in fuzzy logic and its application to usuality and reasoning with dispositions
A fuzzy syllogism in fuzzy logic is defined to be an inference schema in which the major premise, the minor premise and the conclusion are propositions containing fuzzy quantifiers. A basic fuzzy syllogism in fuzzy logic is the intersection/product syllogism. Several other basic syllogisms are developed that may be employed as rules of combination of evidence in expert systems. Among these is the consequent conjunction syllogism. Furthermore, it is shown that syllogistic reasoning in fuzzy logic provides a basis for reasoning with dispositions; that is, with propositions that are preponderantly but not necessarily always true. It is also shown that the concept of dispositionality is closely related to the notion of usuality and serves as a basis for what might be called a theory of usuality - a theory which may eventually provide a computational framework for commonsense reasoning.
Zadeh, L. A.↗
Disposition of recommended modifications of JSC 30426
On May 11, 1988 changes and additions to the Space Station External Contamination Control Document JSC 30426 were addressed at length as part of the charter of this workshop. The modifications and disposition thereof are given below in a concise form in order that a clear understanding of the recommendations and current status be presented. The format is that each paragraph under question is given along with the proposed modified paragraph followed by the workshop's disposition. In some cases, a brief explanation of the issue is given prior to the paragraph in question.
Spann, James F.↗
Nasa Records Disposition Handbook: Procedures Governing the Retention, Retirement, and Destruction of Agency Records
This handbook sets forth the disposition of official records of the National Aeronautics and Space Administration. Its provisions are applicable to NASA Headquarters and all field installations. This revised edition has been enlarged in scope and re-titled to provide guidance in all aspects of records retirements, transfers, destruction, and retrievals from Federal Records Centers. New records control schedules have been added and others revised. Also included are procedures for making recommendations for improved coverage of records categories by additions or revisions. The NASA Records Control Schedules are issued under authority of the NASA Records Management Officer in accordance with Section 101-11.406, Federal Property Management Regulations. They were approved for NASA use by the National Archives and Records Service, the General Accounting Office, and by the Joint Committee on the Disposition of Executive Papers, U.S. Congress.
Source record↗
Surplus Plutonium Disposition Sphincter Seal Development & Testing
The Surplus Plutonium Disposition Project (SPD, Project Y744) requested the Savannah River National Laboratory (SRNL) to perform a series of leak and durability tests on sphincter seals that were designed to process blend cans and shield cans into an SPD glovebox. Each sphincter seal assembly consisted of a stainless-steel outer sleeve weldment that was bolted onto the test enclosure and an aluminum inner sleeve that housed seven to nine Neoprene seals. All tests were executed to determine long term effects from processing approximately a year’s worth of blend and shield cans through the sphincter seals.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Alloying of Pu-Al with Stainless Steel for Material Disposition
The Savannah River National Laboratory (SRNL) has evaluated several options for the disposition of stainless-steel clad plutonium metal alloy. One of the technologies under consideration is alloying of the material with stainless steel (SS). The resulting SS-Pu alloy would be a non-proliferable waste form consisting of a secondary Pu composition region microencapsulated in the refractory stainless steel. Two 8-kg ingots were made at SS-1.8Zr-0.4Pu alloys (wt %); 8 kg was determined in a previous study to be the maximum mass of SS ingot at the maximum target Pu loading of 350-g that would result in a SS-4.4Pu alloy (wt %). Two smaller 500-g ingots were also produced at SS-1.6Zr-1.4Pu and SS-1.4Pu (wt %). The 500-g ingots evaluated alloying at a higher Pu concentration than in the 8 kg ingots, and they evaluated the necessity of adding Zr metal to incorporate the Pu and control Pu oxidation. Zr addition was found to be unnecessary to incorporate the Pu and control Pu oxidation. Drill turnings were collected from the large and small ingots and metallographic samples were directly cut from the small ingots. Both were analyzed to validate the structure and composition region formation. Chemical analyses of turnings proved that the Pu was dispersed within the SS ingots.
36 MATERIALS SCIENCE↗
(U) Meeting Notes and Process Flow Diagrams from 3/31/21 Pu/Z/ICE Disposition Meeting [Slides]
This slide set captures the content presented and discussion points covered in the 3/31/21 Pu/Z/ICE disposition working group meeting.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Demonstration of the On-the-Fly Shielding Analysis Method: Spent Fuel and Waste Disposition
This report documents work performed supporting the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management activities under work breakdown structure element 1.08.02.04.01, “Data and Tools Development, Validation, and Maintenance.” In particular, this report fulfills milestone M3SF-21OR020401016, “Implement on-the-fly dose analysis methodology in UNF-ST&DARDS” within work package SF-21OR02040101, “Commercial SNF Characterization - ORNL.” The Used Nuclear Fuel - Storage, Transportation & Disposal Analysis Resource and Data System (UNFST& DARDS) enables automated dose rate calculations for spent nuclear fuel (SNF) transportation packages and storage casks using a Monte Carlo radiation transport code. The explicit method uses a detailed model of the SNF system and its contents. Therefore, a dose rate calculation is required for each as-loaded transportation package or storage cask because the SNF assemblies within a canister typically have unique irradiation characteristics. An alternate method, referred to as the “on-the-fly” shielding analysis method, has been proposed that requires only a set of Monte Carlo dose rate calculations for each transportation packaging/storage cask design. The results of the Monte Carlo dose rate calculations are independent of the SNF assembly irradiation and decay characteristics. The dose rate values may then be combined with the radiation source strength of the SNF assemblies associated with a particular transportation packaging/storage cask design to determine actual dose rates. This report presents on-the-fly dose rate calculations for a representative SNF storage cask and verification of the on-the-fly dose rate calculation results by comparison with reference dose rate calculations using the explicit Monte Carlo dose rate calculation. The on-the-fly shielding analysis method was implemented in UNF-ST&DARDS. A Python program was developed to process the MAVRIC dose rate results obtained by source particle type, energy group, and fuel geometry region. A Python processor created binary files, which were saved as a special UNF-ST&DARDS library for on-the- fly shielding analyses. UNF-ST&DARDS uses the precalculated on-the-fly binary libraries generated by the Python data processor and directly executes the Python code for on-the-fly dose analysis. This Python code unzips the pre-generated binary files mentioned above, reads the data, and combines them with user-specified sources for dose and uncertainty calculations. The Python programs were verified using Excel calculations and by comparison with the values obtained with the MAVRIC post-processing utilities applied to the 3dmap files. This method can currently be used to determine dose rates for as-loaded HI-STORM FW storage casks. The UNF-ST&DARDS analysis wizard for on-the-fly shielding analysis is described in this report.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗