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At least 1,387 records · Page 77

Influence of Nanoconfinement and Elevated Temperatures on Geocolloid-Facilitated Transport of Energy-related Contaminants

The main objective of this project is to obtain a better understanding of the transport behavior and interactions of geocolloids in the presence of energy- related contaminants under bulk and nanoconfined conditions. In that, we first established solid protocols to synthesize and fabricate numerous types of geocolloids with a. various sizes ranging from 40 – 800 nm and b. different hydrophile-lipophile balance (HLB) ratios in the range of 25:75 to 75:25. Direct force measurements with geocolloids having different degrees of surface coverages (i.e. HLB to realistically mimic adsorption of energy related contaminants) were conducted using the Surface Forces Apparatus (SFA) over a distance regime starting from 8μm all the way down to molecular contact. Repulsive forces were observed on approach starting from > 3μm, followed by an exponential increase of which magnitude appears to be larger than a decay length obtained from Derjaguin–Landau–Verwey–Overbeek (DLVO) theory in pure water. When the geocolloids were confined in salted water, the magnitude of onset of repulsion was varied as a function of salinity in solution, which can significantly alter a purely repulsive screened electrostatic (coulombic) interaction arising from, among geocolloids as well as between geocolloids and geosurfaces. The viscosity and flow characteristics of geocolloidal suspensions at different degrees of confinement were also investigated where we identified highly discontinuous rheological behaviors below a critical nanoconfinement level. We anticipate that the knowledge gained through this study will enable the scientists and researchers to better assess transport and fate behaviors of geocolloidal dispersions that can carry energy-related contaminants under realistically emulated geosystem.

03 NATURAL GAS↗

Disposition of Radioisotope Thermoelectric Generators Currently Located at the Oak Ridge National Laboratory - 12232

Under the American Recovery and Reinvestment Act (ARRA), the U.S. Department of Energy (DOE) awarded SEC Federal Services Corporation (SEC) a 34-building demolition and disposal (D and D) project at the Oak Ridge National Laboratory (ORNL) that included the disposition of six Strontium (Sr-90) powered Radioisotope Thermoelectric Generators (RTGs) stored outside of ORNL Building 3517. Disposition of the RTGs is very complex both in terms of complying with disposal facility waste acceptance criteria (WAC) and U.S. Department of Transportation (DOT) requirements for packaging and transportation in commerce. Two of the RTGs contain elemental mercury which requires them to be Land Disposal Restrictions (LDR) compliant prior to disposal. In addition, all of the RTGs exceed the Class C waste concentration limits under Nuclear Regulatory Commission (NRC) Waste Classification Guidelines. In order to meet the LDR requirements and Nevada National Security Site (NNSS) WAC, a site specific treatability variance for mercury was submitted to the U.S. Environmental Protection Agency (EPA) to allow macro-encapsulation to be an acceptable treatment standard for elemental mercury. By identifying and confirming the design configuration of the mercury containing RTGs, the SEC team proved that the current configuration met the macro-encapsulation standard of 40 Code of Federal Regulations (CFR) 268.45. The SEC Team also worked with NNSS to demonstrate that all radioisotope considerations are compliant with the NNSS low-level waste (LLW) disposal facility performance assessment and WAC. Lastly, the SEC team determined that the GE2000 Type B cask met the necessary size, weight, and thermal loading requirements for five of the six RTGs. The sixth RTG (BUP-500) required a one-time DOT shipment exemption request due to the RTG's large size. The DOT exemption justification for the BUP-500 relies on the inherent robust construction and material make-up of the BUP- 500 RTG. DOE-ORO, SEC, and the entire SEC RTG team are nearing the conclusion of the Sr-90 RTG disposition challenge - a legacy now 50 years in the making. Over 600,000 Ci of Sr-90 waste await disposal and its removal from ORNL will mark an historical moment in the clean-up of the cold-war legacy in the ORNL central industrial area. Elimination (i.e., removal) of the RTGs will reduce security risks at ORNL and disposal will permanently eliminate security risks. The RTGs will eventually decay to benign levels within a reasonable timeframe relative to radiological risks posed by long-lived isotopes. The safety authorization basis at ORNL Building 3517 will be reduced enabling greater operational flexibility in future clean-out and D and D campaigns. Upon disposition the Department of Energy will realize reduced direct and indirect surveillance and maintenance costs that can be reapplied to accelerated and enhanced clean-up of the Oak Ridge Reservation. At present, waste profiles for the RTGs are developed and under review by NNSS RWAP staff and approval authorities. Disposition schedule is driven by the availability of compliant shipping casks necessary to safely transport the RTGs from ORNL to NNSS. The first disposal of the RCA RTG is expected in April 2012 and the remaining RTGs disposed in 2012 and 2013. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Achieving the End State for the Pahute Mesa Corrective Action Units at the Nevada National Security Site - 20355

Underground nuclear testing at the Nevada National Security Site (NNSS) ended in 1992. To address the impact of radionuclide contamination from underground nuclear testing on the groundwater resources of Nevada, the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program created the Underground Test Area (UGTA) activity to characterize the radionuclides in groundwater, understand the nature and extent of radionuclide migration, and forecast the distribution of radionuclides in groundwater for 1,000 years. The UGTA activity is regulated by the Federal Facility Agreement and Consent Order (FFACO), an agreement negotiated between the Nevada Division of Environmental Protection (NDEP), DOE, and the U.S. Department of Defense (DoD). DOE is close to achieving the end state (closure in place with monitoring and institutional control) for three of the five UGTA corrective action units (CAUs) on the NNSS. A number of challenges remain to achieve the end state of the other two CAUs, both located on Pahute Mesa. Pahute Mesa was the location of 82 underground nuclear tests, less than 10% of the total number of tests on the NNSS. Yet, Pahute Mesa contains slightly more than 60% of the total radionuclides (in curies). Based on groundwater sampling in monitoring wells, two radionuclide plumes have migrated several kilometers (km) in groundwater from selected test cavities and have crossed the boundary of the NNSS (yet remain within the boundaries of Federally controlled land). The path to achieve the end state continues to evolve as more data are collected and a better understanding of radionuclide migration in groundwater is developed. DOE has invested in drilling and sampling more than 50 characterization and monitoring groundwater wells over the past 25 years. The data indicate that the primary radionuclide of concern is tritium as it is about 89% of the total radionuclide inventory (in curies). As well, only tritium has been measured in groundwater outside of cavities at concentrations exceeding the Safe Drinking Water Act (SDWA) standard. For one of the tritium plumes that has migrated across the NNSS boundary, the average rate of migration has been measured as about 45 meters (m) per year with the rate of migration at the leading edge of the plume of about 85 meters per year. At that rate of migration, the tritium plume will decay to safe levels and not reach the publicly accessible environment at concentrations above the SDWA standard. Other radionuclides, at concentrations below the SDWA standards, will be monitored to ensure they remain at safe levels. Consequently, the end state path forward for Pahute Mesa has evolved to take full advantage of the data from the monitoring network to constrain uncertainty in model forecasts and to reduce reliance on probabilistic simulations. The focus of the end state approach relies on developing a monitoring well network that is protective of human health by increasing confidence that no radionuclide plume in the groundwater will migrate undetected to the accessible environment, located about 22 km from the nearest up-gradient underground nuclear test. Using the measured data to remove uncertainty, the evaluation of radionuclide migration from Pahute Mesa is directed toward meeting the goals of the end state. (authors)

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Production of H{sub 2} by water radiolysis in cement paste under electron irradiation: A joint experimental and theoretical study

Long-term confinement of nuclear waste is one of the main challenges faced by the nuclear industry. Fission products such as {sup 90}Sr and {sup 137}Cs, both β{sup −} emitters known to induce serious health hazards, represent the largest fraction of nuclear waste. Cement is a good candidate to store them, provided it can resist the effects of irradiation over time. Here, we have investigated the effects of β{sup −} decay on cement by performing electron irradiation experiments on different samples. We show that H{sub 2} production in cement, the main effect of water radiolysis, depends strongly on composition and relative humidity. First-principles calculations indicate that the water-rich interlayer regions with Ca{sup 2+} ions act as electron traps that promote the formation of H{sub 2}. They also show that holes localize in water-rich regions in low Ca content samples and are then able to participate in H{sub 2} production. This work provides new understanding of radiolysis effects in cements. - Graphical abstract: H{sub 2} produced by radiolysis in cement paste is investigated by experimental and computational studies.

36 MATERIALS SCIENCE↗

Radiotoxicity Characterization of Multi-Recycled Thorium Fuel - 12394

As described in companion papers, Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the transuranic (TRU) contained in the used nuclear fuel. The potential of thorium as a TRU burner is described in another paper presented at this conference. This paper analyzes the long-term impact of thorium on the front-end and backend of the fuel cycle. This is accomplished by an assessment of the isotopic make-up of Th in a closed cycle and its impact on representative metrics, such as radiotoxicity, decay heat and gamma heat. The behavior in both thermal and fast neutron energy ranges has been investigated. Irradiation in a Th fuel PWR has been assumed as representative of the thermal range, while a Th fuel fast reactor (FR) has been employed to characterize the behavior in the high-energy range. A comparison with a U-fuel closed-cycle FR has been undertaken in an attempt of a more comprehensive evaluation of each cycle's long-term potential. As the Th fuel undergoes multiple cycles of irradiation, the isotopic composition of the recycled fuel changes. Minor Th isotopes are produced; U-232 and Pa-231 build up; the U vector gradually shifts towards increasing amounts of U-234, U-235 etc., eventually leading to the production of non negligible amounts of TRU isotopes, especially Pu-238. The impact of the recycled fuel isotopic makeup on the in-core behavior is mild, and for some aspects beneficial, i.e. the reactivity swing during irradiation is reduced as the fertile characteristics of the fuel increase. On the other hand, the front and the back-end of the fuel cycle are negatively affected due to the presence of Th-228 and U-232 and the build-up of higher actinides (Pu-238 etc.). The presence of U-232 can also be seen as advantageous as it represents an obstacle to potential proliferators. Notwithstanding the increase in the short-term radiotoxicity and decay heat in the multi-recycled fuel, the Th closed cycle has some potentially substantial advantages compared to the U cycle, such as the smaller actinide radiotoxicity and decay heat for up to 25,000 years after irradiation. In order for these benefits to materialize, the capability to reprocess and remotely manufacture industrial amounts of recycled fuel appears to be the key. Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the TRU contained in the current UNF. The general approach and the potential of thorium as TRU burner is described in other papers presented at this conference. The focus of this paper is to analyze the long-term potential of thorium, once the legacy TRU has been exhausted and the thorium reactor system will become self-sufficient. Therefore, a comparison of Th closed cycle, in fast and thermal neutron energy ranges, vs. U closed cycle, in the fast energy range, has been undertaken. The results presented focus on selected backend and front-end metrics: isotopic actinide composition and potential implications on ingested radiotoxicity, decay heat and gamma heat. The evaluation confirms potential substantial improvements in the backend of the fuel cycle by transitioning to a thorium closed cycle. These benefits are the result of a much lower TRU content, in particular Pu-241, Am-241 and Pu-240, characterizing the Th vs. U actinide inventories, and the ensuing process waste to be disposed. On the other hand, the larger gamma activity of Th recycled fuel, consisting predominantly of hard gammas from U-232's decay products, is a significant challenge for fuel handling, transportation and manufacturing but can be claimed as beneficial for the proliferation resistance of the fuel. It is worth remembering that in our perspective the Th closed cycle and the U closed cycle will follow a transmutation phase which will likely take place over several decades and dictate the technologies required. These will likely include remote fuel manufacturing, regardless of the specific system adopted for the transmutation, which could then be inherited for the ensuing closed cycles. Finally, specific data related to the fuel manufacturing and separation technologies and their performance in the prospected industrial scale deployment, are key for further quantification of the potential merits of the options explored. Further studies in this direction should be warranted before making definitive conclusion. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Key results from examinations of seven high burnup pressurized water reactor spent nuclear fuel rods

At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Impact of radon gas concentration in the aerosoles profile

Radon gases relased from building materials and from earth surface are the major responsibility of air ionization. Radon nuclear decay can produce an alpha particle with high energy and Radon progeny. This particle and gamma rays can deliver particles in the air and produce ions with different polarities. This ions, because of induced electric charge, can attach with air aerosols and charge them with their electric charge. The charged aerosols can interact with the other aerosols and ions. Because of this exchange, the air conductivity and the aerosol profiles will change dependently by Radon gas concentration and gamma radiation. Observations show an increase in concentration of Radon during the night, and a decrease during the daylight time. The Radon gas concentration changed hour by hour can induce aerosol profile to change. This dependency between the aerosol profiles and the Radon gas concentrations is discussed.

54 ENVIRONMENTAL SCIENCES↗

33 Shafts Category of Transuranic Waste Stored Below Ground within Area G

This report compiles information to support the evaluation of alternatives and analysis of regulatory paths forward for the 33 shafts. The historical information includes a form completed by waste generators for each waste package (Reference 6) that included a waste description, estimates of Pu-239 and uranium-235 (U-235) based on an accounting technique, and calculations of mixed fission products (MFP) based on radiation measurements. A 1979 letter and questionnaire (Reference 7) provides information on waste packaging of hot cell waste and the configuration of disposal shafts as storage in the 33 Shafts was initiated. Tables of data by waste package were developed during a review of historical documents that was performed in 2005 (Reference 8). Radiological data was coupled with material-type data to estimate the initial isotopic content of each waste package and an Oak Ridge National Laboratory computer code was used to calculate 2009 decay levels. Other sources of information include a waste disposal logbook for the 33 shafts (Reference 9), reports that summarize remote-handled waste generated at the CMR facility (Reference 10) and placement of waste in the 33 shafts (Reference 11), a report on decommissioning of the LAMPRE reactor (Reference 12), interviews with an employee and manager involved in placing waste in the 33 shafts (References 13 and 14), an interview with a long-time LANL employee involved in waste operations (Reference 15), a 2002 plan for disposition of remote-handled TRU waste (Reference 16), and photographs obtained during field surveys of several shafts in 2007. The WIPP Central Characterization Project (CCP) completed an Acceptable Knowledge (AK) summary report for 16 canisters of remote-handled waste from the CMR Facility that contains information relevant to the 33 Shafts on hot-cell operations and timeline (Reference 17).

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E-Area Low-Level Waste Facility Inadvertent Human Intruder Limits and Doses in Support of the PA2022

This report documents the inadvertent human intruder (IHI) analysis for the E-Area Low-Level Waste Facility (ELLWF) at the Savannah River Site (SRS), near Aiken, South Carolina. This analysis supports the revised ELLWF Performance Assessment (PA), complying with the Department of Energy standard for operation of low-level waste disposal facilities (USDOE, 2017). The ELLWF is an operating waste disposal facility and is scheduled to continue accepting waste to 2065. One task of the revised PA is to establish waste inventory limits for the various disposal units at ELLWF. This is done by modeling future contaminant release and transport through applicable pathways to human receptors, comparing predicted doses per disposed curie with applicable performance measures, to obtain inventory limits which will assure that doses to receptors do not exceed performance measures. This report documents results of modeling future doses to one class of receptor, the inadvertent human intruder. It is assumed that after site closure, public knowledge of the site is lost, and IHIs will engage in activities on the ELLWF that will disrupt the closure cap, causing dose to the IHI. Following USDOE (2017), six different stylized exposure scenarios are considered, simulating activities by an IHI which could result in a radiological dose. The six scenarios are: • Acute – Basement Construction: IHI constructs a basement and encounters waste during excavation which is inadvertently mixed with clean soil and diluted. • Acute – Well Drilling: IHI drills a water well through waste and is exposed to drill cuttings mixed with clean soil that are brought to the surface. • Acute – Discovery: IHI begins constructing a basement but stops when encountering the riprap in the final closure cap and is exposed to photon radiation from unexcavated material residing in the undisturbed waste zone. • Chronic – Agriculture: Resident IHI is exposed to waste that was excavated for basement construction and mixed with native soil in the intruder’s vegetable garden. • Chronic – Post-Drilling: Resident IHI is exposed to waste from drill cuttings mixed with native soil and scattered in the garden area. • Chronic – Residential: Resident IHI is exposed to external radiation while in home located above waste with shielding provided by the concrete basement floor and any soil or engineered material remaining between the basement and waste. Dose calculations are performed using the SRNL Dose Toolkit (Aleman, 2023), following the approach of Smith et al (2019). Calculations are performed separately for 27 of the 33 disposal units (DUs) at ELLWF and are radionuclide specific. The results of the IHI analysis include: • Dose Factors: mrem per disposed curie (acute) and mrem/yr per disposed curie (chronic) for each parent radionuclide, for each DU. • Inventory Limits: in curies, for each parent radionuclide, for each DU. • Estimated Dose to IHI: mrem (acute) and mrem/yr (chronic), for each DU, given its projected closure inventory without inventory biases applied. Most DU-specific IHI inventory limits are in the range of 10 3 to 10 7 curies per nuclide. The lowest inventory limits are associated with gamma-emitters such as Sn-126, Ra-226, Th-232, and Cm-248. Radionuclides with short half-lives such as Pu-241, and nuclides which are pure beta emitters or which decay by electron capture, such as Ni-59 and Ni-63, have the highest limits. For the 27 evaluated DUs, predicted IHI doses are shown in Table ES-1. The maximum acute dose is 1.18 mrem, at ST23, much less than the DOE performance measure of 500 mrem (USDOE, 2017). The highest chronic dose is 37.2 mrem/yr at ST02, below the DOE performance measure of 100 mrem/yr. Also shown are estimated inventory sums of fractions (SOFs) at closure in 2065, for groundwater (GW) and IHI pathways. For each DU, the inventory is constrained by the GW pathway. For most DUs, the IHI SOFs are approximately 1000 times lower than the GW SOF values, and the IHI pathway does not drive risk for any disposal unit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Remote Radiation Sensing Using Aerial and Ground Platforms

Remote sensing of ionizing radiation has a significant role in waste management, nuclear material management and nonproliferation, and radiation safety. Robotic platforms can surpass the number of tasks that are achieved by humans. With this technique, the operator's radiation exposure can be decreased. Remote sensing allows for the evaluation and monitoring of radiological contamination. Gamma-ray and neutron sensors were integrated onto the robotic platforms. This approach allows for the radiation sensor data to be dynamically tracked and mapped thus enabling further analysis of the radiation flux in temporal and spatial domains. The goal is to complete scheduled tasks while the robot is being irradiated. To achieve this, electronic components must be shielded and radiation hardened. CZT Detector: Cadmium Zinc Telluride (CZT) detector technology has been a promising solution for gamma-ray and x-ray measurements. Detector data is transferred to the Odroid minicomputer that controls and powers the module via the USB. Robot Operating System (ROS) was utilized for data acquisition and data fusion. The Mariscotti method was employed for the spectrum analysis. A function was programmed in ROS for the automatic identification of photopeaks. CLYC Detector: A Cs{sub 2}LiYCl{sub 6}:Ce{sup 3+} (CLYC) detector was used for simultaneous medium-resolution gamma-ray measurements and neutron counting. A 2.54 cm diameter photomultiplier tube (PMT) was equipped with a high voltage supply and a miniature digitizer. Gamma-ray excitation: fast core-to-valence luminescence (CVL) with 1 ns decay constant, and prompt Ce{sup 3+} emission with 50 ns decay constant. Neutron excitation: slow cerium self-trapped excitation (Ce{sup 3+} STE), 1000 ns decay constant. Radiation Source Localization: Maximum Likelihood Estimation (MLE) and gradient-based methods were used to locate the position of a radiation source based on measured radiation intensities. Multi-Particle Transport Code FLUKA: Estimation of radiation damage of the electronic components is important in order to optimize the robot's operational time while it is irradiated. Displacement per atom (DPA) represents the radiation damage in materials exposed to the ionizing radiation. Various shielding layers of different thickness t were analyzed (< 5% statistical error). The model of the controller of the UAS was designed in FLUKA. Conclusion: CZT and CLYC detectors were integrated onto the robotic platforms. Radiation source localization and contour mapping using robotic platforms were studied. Functions for data analysis and fusion were developed in ROS. FLUKA code was utilized to analyze DPA values. Layers of low-density and high-density materials were used to shield the UAS electronics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Advanced Fuel Cycle Cost Basis – 2017 Edition

This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Risk-Based Remediation Approach for Cs-137 Contaminated Sediment/Soils at the Savannah River Site (SRS) Lower Three Runs Tail (U) - 13348 - SRNS-RP-2012-00546

Lower Three Runs is a large blackwater stream that runs through the eastern and southern portion of the Savannah River Site. The Lower Three Runs watershed includes two SRS facility areas: P Area (P Reactor) and R Area (R Reactor) that provided effluent discharges to Lower Three Runs. During reactor operations, effluent discharges were well above natural (pre-industrial) or present day stream discharges. The watershed contains a 2,500-acre mainstream impoundment (PAR Pond), several smaller pre-cooler ponds, and a canal system that connects the pre-cooler ponds and discharges surface water to PAR Pond. From the PAR Pond dam, Lower Three Runs flows approximately 36 kilometers braiding through bottom-land/flood-plain forests before it enters the Savannah River. About eight kilometers downstream from the PAR Pond dam, the SRS boundary narrows (termed the Lower Three Runs tail) providing a limited buffer of DOE property for the Lower Three Runs stream and associated flood-plain. Previous screening characterization efforts revealed Cs-137 contamination in the sediment/soils of the flood-plain. As a part of the American Recovery and Reinvestment Act stimulus package, a comprehensive characterization effort was executed on the sediment/soils of the Lower Three Runs tail flood-plain providing a comprehensive look at the contaminant signature of the area. As a follow-up to that characterization, a regulatory decision Core Team, comprised of members of the South Carolina Department of Health and Environmental Control, Environmental Protection Agency - Region IV, and DOE, conducted negotiations on a risk-based approach to address the level of contamination found in the tail flood-plain as an early action that provided a long-term solution to exposure scenarios. For evaluation purposes, the adolescent trespasser was selected as the most likely human receptor for the Lower Three Runs tail portion because of the natural attractiveness of the area for recreational activities (i.e., hunting, fishing, hiking etc.) and access from public property. Exposure of the adolescent trespasser to Cs-137 contaminated sediment/soil at concentrations greater than 23.7 pico curies per gram have been calculated to result in an unacceptable cancer risk (> 1 x 10{sup -4}). Comparing the characterization sampling results conducted in 2009 with the benchmark concentration of 23.7 pCi/g, identified elevated risk levels along three sampling areas in the Lower Three Runs tail portion. On January 5, 2012, it was agreed by the core team that a Removal Action in the Lower Three Runs tail was to be conducted for the identified soil/sediment locations in the three identified areas that exceed the 1 x 10{sup -4} risk (23.7 pCi/g) for the adolescent trespasser receptor. The addition of Land Use Controls following the Removal Action was appropriate to protect human health and the environment. A systematic screening matrix was initiated at the identified hot spots (i.e., sampling points with Cs-137 activities greater than 23.7 pCi/g) to identify the limits of the excavation area. Sediment/soil within the defined removal areas would be excavated to the depth necessary to achieve the cleanup goal and disposed of in a CERCLA Off-Site Rule approved disposal facility. It was agreed that this removal action would adequately reduce the volume of available Cs-137 in the Lower Three Runs tail and consequently residual activities of the Cs-137 would decay over time reducing the amount of Cs-137 available in the tail which would curtail risk. The Land Use Controls consist of installation of an additional seven miles of fencing at major road crossings, utility easements, and at areas that showed a higher probability of access. In addition, signs were placed along the entire SRS perimeter of the Lower Three Runs tail approximately every 200 feet. Sign posts included both a No Trespassing sign and a Contaminant Warning sign. The project initiated a subcontract for both the removal action and the installation of fencing and signs on May 1, 2012. All field activities were completed by July 26, 2012. The project excavated and disposed of over 2700 cubic yards of contaminated sediment/soil, erected approximately seven miles of fence and placed over 2,000 signs demonstrating DOE's commitment to protect human health and act as a good neighbor to residents in the area. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Modeling and Analysis of Coupled Flow Through Packed Beds as a Diagnostic Tool for Packed Bed Reactor Health - 17172

The US Department of Energy, Office of River Protection, is building the Waste Treatment and Immobilization Plant (WTP) in Richland, Washington, to immobilize the hundreds of millions of liters of radioactive waste stored at the Hanford Site in Washington state. The offgas generated during vitrification of these wastes requires treatment to remove and/or destroy chemical and radioactive hazardous components before release to the atmosphere. In particular, volatile and semi-volatile hydrocarbon components are to be catalytically oxidized to carbon dioxide and water, whereas nitrogen oxides (NOx) are to be catalytically reduced to nitrogen and water. These tasks are accomplished in a series of packed bed reactors. Achieving the requisite Destruction/Removal Efficiency (DRE) for each of the tracked components is highly dependent on the flow characteristics of the offgas through the packed beds. The flow kinetics may be monitored and characterized during the initial warm-up of the system. The gases, initially at 76 deg. C (169 deg. F), (a) flow through one side of a countercurrent heat exchanger to recover heat from the exiting gases (which have already passed through the catalytic beds), (b) are then further heated electrically to approximately 400 deg. C (750 deg. F; the preferred operating temperature of the Thermo Catalytic Oxidizer (TCO)), (c) pass through the TCO where the reduced components of the offgas are oxidized over a supported Pt catalyst bed, (d) are partially cooled to approximately 315 deg. C (600 deg. F) by addition of approximately 25% (by weight) diluted ammonia/air mixture, (e) pass through the Selective Catalytic Reducer (SCR) where the NOx components are converted to N{sub 2} and H{sub 2}O over a supported V2O{sub 5}/TiO{sub 2} catalyst bed, and (f) exit through the hot side of the countercurrent heat exchanger. The electrical heater is not powerful enough to raise the temperature of the offgas to the desired 400 deg. C without the contribution of the heat recovered in the heat exchanger. Moreover, the thermal mass of the beds and housings of the TCO and SCR require a significant input of thermal energy to reach the desired operating temperatures; the better part of a day is required to bring the unit from ambient temperatures to operating temperatures. Therefore, the temperature of the stream during the initial warm-up of the system at each point in the system is dependent upon the thermal mass, flow dispersion, and permeability of the components, quantified as characteristic/representative volumes and residence times. By monitoring the temperature of the offgas as it enters and exits each portion of the system during this warm-up period one may model the system as a series of small Continuously Stirred Tank Reactors (CSTR)s (referred to as 'CSTRlets') to explain the flow kinetics through the catalyst beds. These component CSTRlets are adjustable in size and number, that one may model the flow profile of the gases through the catalyst beds. For example, a single large CSTR will present an effluent temperature profile with a simple exponential decay shape. Concatenating more CSTRlets generate more complex sigmoidal responses. Concatenating one CSTRlet significantly larger than the others in the train also alters the shape of the effluent temperature profile, in a fashion not replicable by a train of equal-sized CSTRlets. The ultimate goal is to accurately emulate the entire system as a representative CSTRlet train. Once accomplished, this characterization may be used in subsequent warm-ups (following shutdown for catalyst maintenance, for example) or by an intentional step change in inlet offgas temperature to detect/diagnose anomalies in flow through the beds (e.g., channeling or partial blockage.) This work presents development of this model through analytical, calculational, and empirical means. - The analytic model is accomplished by Laplace transform analysis of each component to recover component process transfer functions. - The calculational portion is performed via brute force finite element calculation. - Fitting empirical data from the warm-up of the TCO skid allows one to select the hypothetical CSTRlet trains which best fit the TCO and SCR beds for the analytical and finite element calculations. The first portion of this paper presents the development of the analytical (Laplace space) model of the coupled system. Estimated values for the initial temperature, mass, specific heat, electrical heat, and relative quantities of the offgas, catalyst beds, stainless steel housings, and ammonia dilution stream are used to generate expected temperature profiles at several selected junctures of the system. The number and size of the component hypothetical CSTRlets are calculated from the analytical model and are used to predict temperature profiles at the exit of the system and at intermediate junctions between system components (where thermocouples are present in the actual unit.) The finite element calculations employ a series of CSTRlet volumes to produce a similar suite of temperature profiles, for comparison with the analytical model results. Finally, empirical data from the Factory Acceptance Test system warm-up are used to select and compare the results from each modeling strategy to select the CSTRlet train(s) which best model the behavior of the actual system. This work presents a novel method of system component characterization, of value to the WTP plant and similar vitrification facilities as well as chemical process industries unrelated to waste management. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

PGSFR General Arrangement

The Nuclear Island of the PGSFR is constructed upon the seismic isolation structure bearing a common basemat of the Reactor Building and the Auxiliary Building. The Reactor Building is divided into two different functional structures: 1) containment structure comprised of a operating floor, head access area (HAA), annulus area and etc., 2) the other areas accessible to reactor vault area outside of the containment structure via the Auxiliary Building. The arrangement of the Auxiliary Building is composed of symmetric divisional areas where there are two steam generators, heat exchangers of the Decay Heat Removal System (DHRS), facilities of the Intermediate Sodium Purification System (ISPS), the Primary Inert Gas Processing System (PIGPS), a main control room (MCR), electric equipment rooms, and main steam valve houses (MSVH), and etc. The general arrangement of the Reactor Building and the Auxiliary Building of the PGSFR is currently in development. Not only NSSS but supporting systems have not been finalized yet. Moreover, the other facility buildings such as a Turbine Generator Building, a fuel cycle facility, radwaste treatment facilities, and etc. have to be designed with relevant interfaces to the Reactor Building and the Auxiliary Building. We can expect additional modifications to optimize the general arrangement. However, the basic principles such as sodium and non-sodium segregation, convenient accessibility and maintenance, economics of architectural volume would be kept in the design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Second Report of the Nuclear Data Subcommittee of the Nuclear Science Advisory Committee

The central importance of the nuclear data curated by the US Nuclear Data Program (USNDP) for clean energy generation, national security, nonproliferation, medical applications, and space exploration as well as basic science was described in a prior report issued by the DOE/NSF Nuclear Science Advisory Committee subcommittee on Nuclear Data (NSAC-ND) in September 2022. In this report, we present a set of fourteen (14) recommendations that will enhance and advance DOE-NP's stewardship of nuclear data. The first three recommendations focus on the existing core USNDP capabilities, namely: 1) Support the nuclear structure evaluation workforce to improve the currency, consistency, and accessibility of the Evaluated Nuclear Structure Data File (ENSDF); 2) Enhance nuclear reaction evaluation within the USNDP in support of the Evaluated Nuclear Data File (ENDF) through expansion of the workforce and integration of high-performance computing, automation, and machine learning and; 3) Continue atomic mass evaluation in support AME and NUBASE databases. This is followed by eight (8) recommendations representing new cross-cutting initiatives involving both measurement and evaluation to address outstanding nuclear data needs. These new initiatives require a highly trained, diverse workforce that includes personnel with expertise from both inside and outside the nuclear physics community from which evaluators have traditionally been recruited. As such, many of these initiatives are accomplished via a Topical Nuclear Data Collaborations (TNDC). A TNDC is made up of domestic and international stakeholders, subject matter and nuclear data experts, and nuclear data evaluators and features a workforce development plan to ensure that nuclear data evaluators maintain currency in the relevant applications and are seen as equity partners in the endeavor. These include: 1) Establish a coordinated effort to improve evaluation and modeling in nuclear astrophysics for stellar dynamics, multi-messenger astronomy and nucleosynthesis; 2) Initiate a TNDC to develop and maintain nuclear structure evaluation beyond discrete states, including nuclear level densities, photon strength functions and photonuclear data for improved reaction modeling, and exploring nuclear structure at finite temperature; 3) Create a TNDC to perform correlated fission data evaluation, including cross sections, fragment yields, v(A), v(E n ) for nuclear energy, national security, nonproliferation and basic science; 4) From a panel of subject matter experts to establish and annually update a roster of key decay data to nurture its accelerated dissemination including both measurement and evaluation for targeted high-value nuclides for national security, nonproliferation and medical applications; 5) Comprehensive, consistent neutron-induced structure and reaction data for nuclear energy, national security, nonproliferation and planetary nuclear spectroscopy; 6) Charged-particle stopping powers for detector design, space effects and ion beam therapy; 7) High-energy reactions for space exploration and medical nuclide production, and; 8) The creation of an infrastructure for open data and data preservation for use by the entire nuclear physics community. All told, these initiatives require approximately $6.5M increase in NP support of the USNDP in fiscal year 2023 dollars and would require at least 3-5 years to carry out due to the length of time needed to recruit and train new nuclear data researchers. This relatively modest investment would help ensure that the fruits of the nuclear data research carried out by DOE-NP and its collaborators would be brought to bear to address some of the most important needs of our nation and the world. To ensure effective execution of this plan, we present an overview of recruitment, training, and retention goals for the USNDP, the centerpiece of which is a mutually agreed upon code of conduct. Finally, we identify the facility and instrumentation needed to perform the recommended experimental activities. This includes a short review of target fabrication capabilities, reactors, neutron beam, light- and heavy-stable ion, gamma-ray, high-energy and radioactive ion beam facilities. Lastly, a more complete appendix of experimental facilities previously compiled is included with new input provided for 6 facilities.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Induction Heating for Tritium Storage Beds

Metal hydride beds used for storing hydrogen isotopes accumulate helium due to beta-decay of tritium. Helium alters the hydrogen capacity and plateau pressures of these storage beds and ultimately results in the beds being disqualified for use in tritium facilities. Removal of helium and hydrogen isotopes for disposal or storage bed regeneration can be achieved with sufficient heating; however, existing Gen 1 and Gen 2 storage beds are not designed to reach helium removal temperatures and face disposal challenges. In this report, we describe experiments and Multiphysics models that were used to develop a method to achieve induction heating of existing tritium storage beds using induction coils placed on the outside of the beds. The heating method involves low-frequency (< 100 Hz) induction heating through flexible solenoid coils that wrap tightly around the outer jacket of a tritium storage bed. This method of heating offers a pragmatic path to heat existing storage beds for metal hydride regeneration, helium removal for disposal, and possibly routine gas cycling operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Engineering Scale Drying of Aluminum-Clad Spent Nuclear Fuel: Experiment Report

Engineering-scale tests were performed to evaluate the relative effectiveness of forced gas and vacuum-drying processes on aluminum-clad spent nuclear fuel (ASNF) in preparation for extended dry storage. Drying models were developed to improve confidence in the likely range of conditions at the start of dry storage and to understand the factors most significant to each process. Testing was performed at the Holtec International Forced Helium Dehydration training facility in Camden, New Jersey, by students from the University of South Carolina in collaboration with Idaho National Laboratory. A full-size Type 1a basket, with a full-diameter vessel configuration (based on the Department of Energy standardized canister design), was used in combination with a 10 assembly surrogate ASNF arrangement to provide a proxy for geometry. A resistance heater incorporated in one assembly simulated the influence of 100 W decay heat. This instrumented chamber allowed testing of the removal of bulk water, physisorbed water, and chemisorbed water over a range of operating parameters with either drying process. Thick oxide films were grown on aluminum test plates to mimic the ASNF surface chemistry. These one-use test plates were housed within select assemblies during drying. Post-test analysis of these plates gave a measure of the moisture removed during each test, which was compared to a control. Instrumentation provided detailed thermal, humidity, and gas or pressure data relevant to each test. Model simulations generally compared favorably to instrument data for the tests. While both processes were able to remove bulk water, removal of chemisorbed water was shown to depend heavily on local temperature. To the disadvantage of vacuum drying, operations approaching or exceeding 220°C achieved consistently more dehydroxylation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗