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At least 433 records · Page 24

Computational model development and validation of fuel dispersal phenomena

As the US nuclear industry is renewing efforts to extend the rod average burnup limits in existing pressurized water reactors, it becomes imperative to investigate the safety implications associated with high burnup fuel fragmentation and the potential dispersal of fragmented fuel into the reactor coolant system. Additionally, as there is a growing interest in transitioning to new nuclear fuel designs to accommodate higher fuel burnup levels, it becomes essential to conduct studies on fuel behavior during design basis accident scenarios, particularly during a postulated loss-of-coolant accident, in order to ensure the safe operation of light-water reactors. The present study develops a computational model to simulate the complex three-phase flow of the fuel dispersal phenomena that occur following a breach in the cladding, when the stream of solid fuel particles and carrier fission gas stream may interact with the surrounding liquid medium. Furthermore, the proposed framework considers the solid phase as discrete Lagrangian particles while the interpenetrating gas–liquid continuum is modeled using a Eulerian framework. The developed simulation framework is validated with experimental results performed in a separate-effect test facility by comparing predicted particle settlement locations with the experimental observations reported in the literature. Key highlights of this study include the interpretation of the high-pressure boundary conditions and the three-phase flow coupling strategy. Also discussed are the gas–liquid hydrodynamics and the behavior of particles as they are transported through the gas and liquid phases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.

Sun, Kaichao (ORCID:0000000285871918)↗

Polymer Precursor Derived Li x PON Electrolytes: Toward Li–S Batteries

Efforts to develop polymer precursor electrolytes that offer properties anticipated to be similar or superior to (lithium phosphorus oxynitride, LiPON) glasses are reported. Such precursors offer the potential to be used to process LiPON-like thin glass/ceramic coatings for use in all solid state batteries, ASBs. Here, LiPON glasses provide a design basis for the synthesis of sets of oligomers/polymers by lithiation of OP(NH 2 ) 3–x (NH) x [from OP(NH) 3 ],OP(NH 2 ) 3-x (NHSiMe 3 ) x and [P=N] 3 (NHSiMe 3 ) 6–x (NH) x . Furthermore, the resulting systems have degrees of polymerization of 5–20. Treatment with selected amounts of LiNH 2 provides varying degrees of lithiation and Li + conducting properties commensurate with Li + content. Polymer electrolytes impregnated in/on Celgard exhibit Li + conductivities up to ~1 × 10 –5 S cm –1 at room temperature and are thermally stable to ~150 °C. A Li–S battery assembled using a Li 6 SiPON composition polymer electrolyte exhibits an initial reversible capacity of 1500 mAh g sulfur –1 and excellent cycle performance at 0.25 and 0.5 C rate over 120 cycles at room temperature.

25 ENERGY STORAGE↗

A general mechanism for stabilizing the small sizes of precious metal nanoparticles on oxide supports

We recently discovered that MgAl2O4 spinel {111} nano-facets optimally stabilize the small sizes of platinum nanoparticles even after severe high temperature aging treatments. Here we report the thermal stabilities of other precious metals with various physical and chemical properties on the MgAl2O4 spinel {111} facets, providing important new insights into the stabilization mechanisms. Besides Pt, Rh and Ir can also be successfully stabilized as small (1-3 nm) nanoparticles and even as single atomic species after extremely severe (800 °C, 1 week) oxidative aging. However, other metals either aggregate (Ru, Pd, Ag, and Au) or sublimate (Os) even during initial catalyst synthesis. On the basis of ab initio theoretical calculations and experimental observations, we rationalize that the exceptional stabilization originates from lattice matching, and the correspondingly strong attractive interactions at interfaces between the spinel {111} surface oxygens and epitaxial metals\metal oxides. On this basis, design principles for catalyst support oxide materials that are capable in stabilizing precious metals are proposed.

anti-sintering↗

Chapter 17: Adding Value to the Biorefinery with Lignin: An Engineer's Perspective

There is a long-standing belief that 'you can make anything out of lignin...except money.' This chapter serves to highlight that opportunities for making money from biomass-derived lignin exist both with current technology in the production of steam and power to new emerging areas of R&D focused on value-added chemical and material coproducts from lignin. To understand and quantify the economic potential for lignin valorization, the techno-economic analysis methodology is first described in detail. As demonstrated in the provided case study, these types of economic evaluations serve not only to estimate the economic impacts that lignin conversion could have for an integrated biorefinery and outline drivers for further cost reduction but also identify data gaps and R&D needs for improving the design basis and reducing the risk for process scale-up.

09 BIOMASS FUELS↗

Ideal hydrodynamic scaling relations for a stagnated imploding spherical plasma liner formed by an array of merging plasma jets

This work presents scaling relations for the peak thermal pressure and stagnation time (over which peak pressure is sustained) for an imploding spherical plasma liner formed by an array of merging plasma jets. Results were derived from three-dimensional (3D) ideal hydrodynamic simulation results obtained using the smoothed particle hydrodynamics code SPHC. The 3D results were compared to equivalent one-dimensional (1D) simulation results. It is found that peak thermal pressure scales linearly with the number of jets and initial jet density and Mach number, quadratically with initial jet radius and velocity, and inversely with the initial jet length and the square of the chamber wall radius. The stagnation time scales approximately as the initial jet length divided by the initial jet velocity. Differences between the 3D and 1D results are attributed to the inclusion of thermal transport, ionization, and perfect symmetry in the 1D simulations. A subset of the results reported here formed the initial design basis for the Plasma Liner Experiment [S. C. Hsu et al., Phys. Plasmas 19, 123514 (2012)].

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Assessment of NDE for Key Indicators of Aging Cables in Nuclear Power Plants – Interim Status

Degradation of the cable jacket, electrical insulation, and other cable components of installed cables within nuclear power plants (NPPs) is known to occur as a function of age, temperature, radiation, and other environmental factors. System tests verify cable function under normal loads; however, the concern is over cable performance under exceptional loads associated with design-basis events (DBEs). The cable’s ability to perform safely over the initial 40 year planned and licensed life has generally been demonstrated and there have been very few age-related cable failures. With greater than 1000 km of power, control, instrumentation, and other cables typically found in an NPP, replacing all the cables would be a severe cost burden. Justification for life extension to 60 and 80 years requires a cable aging management program to justify cable performance under normal operation as well as accident conditions. Currently the gold standard for determining cable insulation degradation is the elongation-at-break (EAB). This, however, is an ex-situ measurement and requires removal of a sample for laboratory investigation. A reliable nondestructive examination (NDE) in-situ approach is desirable to objectively determine the suitability of the cable for service. A variety of tests are available to assess various aspects of electrical and mechanical cable performance, but none of these tests are suitable for all cable configurations nor does any single test confirm all features of interest. Nevertheless, the complete collection of test possibilities offers a powerful range of tools to assure the integrity of critical cables. Licensees and regulators have settled on a practical program to justify continued operation based on condition monitoring of a lead sample set of cables where test data is tracked in a database and the required test data are continually adjusted based on plant and fleet-wide experience. As part of the Light Water Reactor Sustainability program sponsored by the U.S. Nuclear Regulatory Commission, the U.S. Department of Energy, and industry (represented by the Electric Power Research Institute), an assessment of cable NDE methods was commissioned. Technologies include both bulk electrical measurements (Tan d, time domain reflectometry, frequency domain reflectometry (FDR), partial discharge, and other techniques) and local insulation measurement (indenter, dynamic mechanical analysis interdigital capacitance, infrared spectral measurement, etc.). This aging cable NDE program update reviews the full range of techniques but focuses on the most interesting test approaches that have a chance to be deployed in-situ, particularly including Tan d, FDR, and ultrasound methods that have been reviewed most completely in this progress period.

cable aging inspection↗

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). Here, to further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Versatile Test Reactor Conceptual Core Design

The VTR is a 300-MW(thermal) sodium-cooled fast reactor (SFR) designed for the specific purpose of delivering unique testing capabilities to enable the advancement of all reactor technologies. With its flux level, irradiation volume, and operational flexibility, the VTR will enable accelerated testing of materials, fuels, and various components needing irradiation testing. Proven SFR technologies and design approaches have been leveraged in designing the VTR core, ensuring the highest possible readiness level. This resulted in the VTR using ternary metallic fuel and delivering fast flux levels in excess of 4 x 10 15 n/cm 2 ∙ s over large useful volumes, corresponding to about 60 dpa/year in steel. As part of the design efforts, the VTR core performance has been determined for a representative configuration, ensuring that the reactivity control systems offer sufficient shutdown margins, that the core can be safely cooled in all situations, and that reactivity feedback coefficients are conducive to a favorable safety behavior. Furthermore, the incorporation of features such as fuel assembly storage in the shield region supports the flexible and reliable operation of the VTR. Additional design work has been ongoing as well. This includes thorough shielding performance evaluations to ensure safe operation of the VTR, verification and validation of the design tools used to achieve compliance with Nuclear Quality Assurance (NQA-1) requirements, early assessment of the impact of irradiation experiments on the core performance envelope and associated margins, and in-depth uncertainty quantification efforts to quantify the anticipated range of performance characteristics. An experimental program supporting the VTR core design has been set up, with the current focus being on thermal-hydraulic experiments. The purpose of this experimental program is to obtain confirmatory measurements to serve directly as part of the core design basis or as part of the validation cases supporting the simulation tools used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Investigating the Role of Accident Tolerant Cladding on Source Term Reduction for High-Burnup PWRs Using MELCOR

The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.

Accident Tolerant Fuel↗

Intelligent Modeling for Nuclear Power Plant Accident Management

This study explores the viability of using counterfactual reasoning for impact analyses when understanding and responding to “beyond-design-basis” nuclear power plant accidents. Currently, when a severe nuclear power plant accident occurs, plant operators rely on Severe Accident Management Guidelines. However, the current guidelines are limited in scope and depth: for certain types of accidents, plant operators would have to work to mitigate the damage with limited experience and guidance for the particular situation. We aim to fill the need for comprehensive accident support by using a dynamic Bayesian network to aid in the diagnosis of a nuclear reactor’s state and to analyze the impact of possible response measures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Performing k eff Validation of As-Loaded Criticality Safety Calculations Using UNF-ST&DARDS: Sensitivity Calculations

The general method for performing validation of as loaded criticality safety calculations using UNF ST&DARDS is presented in a paper by Clarity, which includes a description of the UNF-ST&DARDS system. Proof-of-principle analyses were performed in the summer of 2019 for MPC-32 dual purpose canisters (DPCs) containing pressurized water reactor (PWR) fuel assemblies. Summaries of these results are presented in this and a companion paper for this conference. The current paper describes the TSUNAMI-3D calculations performed to generate sensitivity data, and the companion paper discusses the selection of critical experiments applicable for validation of the 11 MPC-32 DPCs considered. The generation of sensitivity data for as-loaded spent nuclear fuel (SNF) DPCs is a challenge given the detailed model of the fuel compositions generated by UNF ST&DARDS. Each fuel assembly is modeled with its own irradiation history in 18 axial nodes, unless the fuel assembly is damaged and thus considered as fresh by design basis. This results in a set of 576 fuel compositions, each of which must be processed separately in a multigroup (MG) calculation. Therefore, a continuous-energy (CE) TSUNAMI-3D method was chosen to alleviate this challenge. Two CE TSUNAMI-3D methods are available in SCALE 6.2.3: the iterated fission probability (IFP) and contribution-linked eigenvalue sensitivity/uncertainty estimation via track-length importance characterization (CLUTCH). Since the IFP method is not feasible because of memory requirements associated with its implementation in SCALE, the CLUTCH method was selected for these calculations. CLUTCH has been implemented in SCALE in parallel, allowing long calculations to be performed in reasonable timeframes. The two primary user inputs necessary for CLUTCH calculations are the F*(r) mesh and the number of latent generations used in determining the F*(r) function. This F*(r) function is used as the importance function for fission chains originating in a given volume element (voxel), and it is calculated using the IFP method in the skipped generations. A large number of skipped generations is thus required to ensure accurate calculation of this importance function. In these calculations, 500 generations were used to calculate the F*(r) function. For more information regarding the calculation of F*(r), see Jones [4]. The remainder of this paper is focused on the selection of the F*(r) mesh and the number of latent generations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

GLYCOLIC - FORMIC ACID FLOWSHEET DEVELOPMENT

Flowsheet testing was performed to further develop the nitric/glycolic/formic acid flowsheet as an alternative to the nitric/formic flowsheet currently being processed at the DWPF. This new flowsheet has shown that mercury can be removed in the Sludge Receipt and Adjustment Tank (SRAT) with minimal hydrogen generation. All other processing objectives were also met, including greatly reducing the Slurry Mix Evaporator (SME) product yield stress as compared to the baseline nitric/formic flowsheet. Eight runs were performed in total, including the baseline run. The baseline nitric/formic flowsheet run was extremely difficult to process under existing DWPF acceptance criteria with this simulant at the HM levels of noble metals. While nitrite was destroyed and mercury was removed to near the DWPF limit, the rheology of the SRAT and SME products were well above design basis and hydrogen generation far exceeded the DWPF limit. In addition, mixing during the SME cycle was very poor. In this sense, the nitric/glycolic/formic acid flowsheet represents a significant upgrade over the current flowsheet. In the nitric/glycolic/formic flowsheet runs, mercury was successfully removed with almost no hydrogen generation and the SRAT and SME products yield stresses were within process limits or previously processed ranges. It is recommended that DWPF continue to support development of the nitric/glycolic/formic flowsheet. Although experience is limited at this time, this flowsheet meets or outperforms the current flowsheet in many regards, including off-gas generation, mercury removal, product rheology and general ease of processing. Additional flowsheet testing will allow for a more thorough understanding of the chemistry and effectiveness of the flowsheet over a range of sludge compositions and formic/glycolic ratios. This testing will also show whether the REDOX and metal solubility concerns with this change in the flowsheet can be addressed by just adjusting the volumes of glycolic acid added. The outstanding issues regarding the glycolic/formic flowsheet include increasing understanding of the impact on glass REDOX control and increased metal solubility, particularly iron, during processing. Additionally, evaluations of the utility of the flowsheet over varying sludge compositions should be completed to ensure flowsheet robustness. Work has already been initiated to further understand the REDOX and iron solubility areas.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Integrated Energy System with Beneficial Carbon Dioxide (CO2) Use - Final Scientific/Technical Report

This report presents an integrated energy system that combines the production of substitute natural gas through coal hydrogasification with an algae process for beneficial carbon dioxide (CO2) use and biofuel production (funded under Department of Energy (DOE) contract DE-FE0001099). The project planned to develop, test, operate and evaluate a 2 ton-per-day coal hydrogasification plant and 25-acre algae farm at the Arizona Public Service (APS) 1000 Megawatt (MW) Cholla coal-fired power plant in Joseph City, Arizona. Conceptual design of the integrated system was undertaken with APS partners Air Liquide (AL) and Parsons. The process engineering was separated into five major areas: flue gas preparation and CO2 delivery, algae farming, water management, hydrogasification, and biofuel production. The process flow diagrams, energy and material balances, and preliminary major equipment needs for each major area were prepared to reflect integrated process considerations and site infrastructure design basis. The total project also included research and development on a bench-scale hydrogasifier, one-dimensional (1-D) kinetic-model simulation, extensive algae stressing, oil extraction, lipid analysis and a half-acre algae farm demonstration at APS?s Redhawk testing facility. During the project, a two-acre algae testing facility with a half-acre algae cultivation area was built at the APS Redhawk 1000 MW natural gas combined cycle power plant located 55 miles west of Phoenix. The test site integrated flue gas delivery, CO2 capture and distribution, algae cultivation, algae nursery, algae harvesting, dewatering and onsite storage as well as water treatment. The site environmental, engineering, and biological parameters for the cultivators were monitored remotely. Direct biodiesel production from biomass through an acid-catalyzed transesterification reaction and a supercritical methanol transesterification reaction were evaluated. The highest oil-to-biodiesel conversion of 79.9% was achieved with a stressed algae sample containing 40% algae oil. The effort concluded that producing biodiesel directly from the algae biomass could be an efficient, cost-effective and readily scalable way to produce biodiesel by eliminating the oil extraction process.

09 BIOMASS FUELS↗

GLYCOLIC-FORMIC ACID FLOWSHEET FINAL REPORT FOR DOWNSELECTION DECISION

Flowsheet testing was performed to develop the nitric-glycolic-formic acid flowsheet (referred to as the glycolic-formic flowsheet throughout the rest of the report) as an alternative to the nitric/formic flowsheet currently being processed at the DWPF. This new flowsheet has shown that mercury can be removed in the Sludge Receipt and Adjustment Tank (SRAT) with minimal hydrogen generation. All processing objectives were also met, including greatly reducing the Slurry Mix Evaporator (SME) product yield stress as compared to the baseline nitric/formic flowsheet. Forty-six runs were performed in total, including the baseline run and the melter feed preparation runs. Significant results are summarized. The baseline nitric/formic flowsheet run, using the SB6 simulant produced by Harrell was extremely difficult to process successfully under existing DWPF acceptance criteria with this simulant at the HM levels of noble metals. While nitrite was destroyed and mercury was removed to near the DWPF limit, the rheology of the SRAT and SME products were well above design basis and hydrogen generation far exceeded the DWPF SRAT limit. In addition, mixing during the SME cycle was very poor. In this sense, the nitric/glycolic/formic acid flowsheet represents a significant upgrade over the current flowsheet. Mercury was successfully removed with almost no hydrogen generation and the SRAT and SME products yield stresses were within process limits or previously processed ranges. The glycolic-formic flowsheet has a very wide processing window. Testing was completed from 100% to 200% of acid stoichiometry and using a glycolic-formic mixture from 40% to 100% glycolic acid. The testing met all processing requirements throughout these processing windows. This should allow processing at an acid stoichiometry of 100% and a glycolic-formic mixture of 80% glycolic acid with minimal hydrogen generation. It should also allow processing endpoints in the SRAT and SME at significantly higher total solids content and may be effective at acid stoichiometries below 100%, although no testing was performed below 100% acid stoichiometry. There are several issues related to the development of the glycolic-formic flowsheet. First, the measurement of anions using the new glycolate anion procedure likely needs to be optimized to improve the accuracy of the anions important to DWPF processing and REDOX prediction. Second, the existing REDOX equation with an added term for glycolate did not accurately predict the glass REDOX for the glycolic-formic flowsheet. Improvement of the anion measurement or modification of the REDOX methodology or equation may be necessary to improve the REDOX prediction. Last, the glycolic-formic flowsheet dissolves a number of metals, including iron. This leads to a thinner slurries but also dissolves up a portion of the iron, which is currently used for criticality control. It is recommended that DWPF continue to support development of the glycolic-formic flowsheet. This flowsheet meets or outperforms the baseline flowsheet in off-gas generation, mercury removal, product rheology and general ease of processing. Additional testing is in progress to demonstrate the effectiveness of the nitric-glycolic-formic flowsheet in processing a wide sludge processing window using the matrix sludge simulants.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Characterization of Settler Tank, KW Container and KE Container Sludge Simulants

The Sludge Treatment Project (STP), managed by CH2M Hill Plateau Remediation Company (CHPRC) has specified base formulations for non-radioactive sludge simulants for use in the development and testing of equipment for sludge sampling, retrieval, transport, and processing. In general, the simulant formulations are based on the average or design-basis physical and chemical properties obtained by characterizing sludge samples. The simulants include surrogates for uranium metal, uranium oxides (agglomerates and fine particulate), and the predominant chemical phases (iron and aluminum hydroxides, sand). Specific surrogate components were selected to match the nominal particle-size distribution and particle-density data obtained from sludge sample analysis. Under contract to CHPRC, Pacific Northwest National Laboratory (PNNL) has performed physical and rheological characterization of simulants, and the results are reported here. Two base simulant types (dry) were prepared by STP staff at the Maintenance and Storage Facility and received by PNNL in February 2009: Settler Tank Simulant and KW Container Sludge Simulant. A third simulant, KE Container Sludge Simulant was received by PNNL in December 2010. The objectives of this simulant characterization effort were to provide baseline characterization data on simulants being used by STP for process development and equipment testing and provide a high-level comparison of the simulant characteristics to the targets used to formulate the simulants.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Flood Hazards for the Materials and Fuels Complex at the Idaho National Laboratory Site

Researchers at Pacific Northwest National Laboratory conducted a flood hazard analysis for the Materials and Fuels Complex (MFC) site located at the Idaho National Laboratory (INL) site in southeastern Idaho. The general approach for the analysis was to determine the maximum water elevation levels associated with the design-basis flood (DBFL) and compare them to the floor elevations at critical building locations. Two DBFLs for the MFC site were developed using different precipitation inputs: probable maximum precipitation (PMP) and 10,000 year recurrence interval precipitation. Both precipitation inputs were used to drive a watershed runoff model for the surrounding upland basins and the MFC site. Outflows modeled with the Hydrologic Engineering Centers Hydrologic Modeling System were input to the Hydrologic Engineering Centers River Analysis System hydrodynamic flood routing model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

REPORTABLE RADIONUCLIDES IN DWPF SLUDGE BATCH 7A (MACROBATCH 8)

The Waste Acceptance Product Specifications (WAPS) 1.2 require that the waste producer 'shall report the curie inventory of radionuclides that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115.' As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type all radionuclides that have half-lives longer than 10 years and contribute greater than 0.01 percent of the total curie inventory from the time of production through the 1100 year period from 2015 through 3115. The initial list of radionuclides to be reported is based on the design-basis glass identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report. However, it is required that the list be expanded if other radionuclides with half-lives greater than 10 years are identified that meet the 'greater than 0.01% of the curie inventory' criterion. Specification 1.6 of the WAPS, International Atomic Energy Agency Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, and U-238; and Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete list of reportable radionuclides must also include these sets of U and Pu isotopes - and the U and Pu isotopic mass distributions must be identified. The DWPF receives HLW sludge slurry from Savannah River Site (SRS) Tank 40. For Sludge Batch 7a (SB7a), the waste in Tank 40 contained a blend of the heel from Sludge Batch 6 (SB6) and the Sludge Batch 7 (SB7) material transferred to Tank 40 from Tank 51. This sludge blend is also referred to as Macrobatch 8. Laboratory analyses of a Tank 40 sludge sample were performed to quantify the concentrations of pertinent radionuclides in the SB7a waste. Subsequently, radiological decay and in-growth were calculated over the time period from 2015 to 3115. This provided a basis for characterizing the radionuclide content of SB7a over time and for identifying the 'reportable radionuclides.' Details of the characterization methodology and the analytical results are the focus of this report. This work was performed at the request of the Waste Solidification Engineering Department of Savannah River Remediation, initiated via Technical Task Request (TTR) HLW-DWPF-TTR-2010-0031. A minor revision in the reporting requirements was requested via a subsequent email communication. The work was conducted in accordance with the protocols identified in Task Technical and Quality Assurance Plan SRNL-RP-2010-01218 and Analytical Study Plan SRNL-RP-2010-01219. All of the raw data related to this scope have been recorded in laboratory notebook SRNL-NB-2011-00061. The overall goal of this task was to characterize the radionuclide content of the SB7a waste sufficiently to meet the WAPS and DWPF reporting requirements. The specific objectives were: (1) Quantify the current concentrations of all radionuclides impacting (or potentially-impacting) the total curie content between calendar years 2011 and 3115. Also quantify the current concentrations of other radionuclides specifically requested in the TTR or required by the WAPS. (2) Calculate future concentrations of decayed and in-grown radionuclides impacting the total curie content between calendar years 2015 and 3115; (3) Identify as 'reportable' all radionuclides contributing {ge} 0.01% of the total curie content from 2015 to 3115 and having half-lives {ge} 10 years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗