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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Preliminary Study of Iodine Gas Removal in Sodium Pools

Potential iodine gas release from failed fuel pins is a critical factor in the source term analysis of oxide fuel-loaded sodium fast reactors (SFRs). The accumulated iodine-containing gas mixtures inside pin plenums are expected to be ejected during pin failures and rise through sodium pool, with potential release of gaseous iodine to the cover gas region. Due to its potential radiological impacts, a proper assessment of iodine behavior is necessary for an accurate source term assessment. Throughout the bubble rise trajectory in the sodium pool, iodine gas is continuously removed or transformed at the bubble interface by diffusion, as the combining reaction between the iodine and sodium to form sodium iodide (NaI) is a chemically preferred process. As the final amount of iodine released from the facility is strongly influenced by the removal phenomenon inside the sodium pool, experiments were previously performed by PNC (Power Reactor and Nuclear Fuel Development Corporation) to provide insight into this phenomenon. To assess the accuracy of present methods for predicting iodine gas removal within sodium pools, several candidate approaches, available in source term analysis codes, have been summarized and evaluated in this study. Spherical cap bubbles and spherical bubbles are considered in accordance with the methods adopted in each approach, and different forms of correlations for major parameters have been implemented in accordance with the original adoptions. Based on the summarized results, important aspects to be considered have been derived.

Decontamination↗
DOE OSTI2022

Electron beam treatment for the removal of 1,4-dioxane in water and wastewater

Electron beam (e-beam) treatment uses accelerated electrons to form oxidizing and reducing radicals when applied to water without the use of external chemicals. In this study, electron beam treatment was used to degrade 1,4-dioxane in several water matrices. Removal improved in the progressively cleaner water matrices and removals as high as 94% to 99% were observed at a dose of 2.3 kGy in secondary effluent. 1,4-dioxane removal was confirmed to be primarily through hydroxyl radical oxidation. The calculated electrical energy per order was found to be 0.53, 0.26, and 0.08 kWh/m 3 /order for secondary effluent (Avg. total organic carbon (TOC) 9.25 mg/L), granular activated carbon effluent (TOC 3.46 mg/L), and ultrapure water, respectively, with a 70% generation and transfer efficiency applied.

1,4-dioxane↗
DOE OSTI2024

Factorial experiment to identify two-way interactions between temperature, harvesting period, hydraulic retention time, and light intensity that influence the biomass productivity and phosphorus removal efficiency of a microalgae–bacteria biofilm

Rotating algae biofilm reactors (RABRs) can reduce energy requirements for wastewater reclamation but require further optimization for implementation at water resource recovery facilities (WRRF). Optimizing RABR operation is challenging because conditions at WRRF change frequently, and disregarding interaction terms related to these changes can produce incorrect conclusions about RABR behavior. This study evaluated the two-way interaction and main effects of four factors on the biomass productivity and phosphorus removal efficiency of a microalgae-bacteria biofilm grown in municipal anaerobic digester centrate, with factor levels and operating conditions selected to mimic a pilot RABR at a WRRF in Utah. Two-way interactions harvesting period*light intensity (LI), harvesting period*temperature, and LI*hydraulic retention time (HRT) had significant effects on biomass productivity: at high temperature and low LI, highest biomass productivity was achieved with a 14-day harvesting period, but at medium temperature and high LI, highest biomass productivity was achieved with a 7-day harvesting period. At high HRT, highest biomass productivity occurred at low LI, but at low HRT, highest biomass productivity occurred at high LI. Phosphorus removal was strongly influenced by LI and occurred most rapidly during the first 2 days HRT, which suggests precipitation contributed significantly to phosphorus removal. These observations provide insight for further RABR optimization.

09 BIOMASS FUELS↗
DOE OSTI2020

Cesium Removal Performance Comparisons of Crystalline Silicotitanate Media Batches with Savannah River Site Waste Simulant

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs- 137. The TCCR project is currently focused on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium salt cake solids) within the tank followed by at-tank ion exchange column treatment. Plans are underway to prepare and install a second TCCR unit at SRS. Capacity and particle size differences exist between archived (IE-911) and more recently prepared CST media batches (9120-B and 9140-B). Side-by-side comparison testing was performed to evaluate the cesium removal performance of each batch to aid in selecting the preferred CST batch and media characteristics to load into the second TCCR unit. Batch contact equilibrium and flow-through column tests have been conducted with three CST batches using an SRS Average Simulant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Summary of Expedited Results from Samples Supporting Tank Closure Cesium Removal (TCCR) Batch 3

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. As part of that process, Savannah River National Laboratory (SRNL) receives and analyzes samples in support of the qualification of each batch to be processed. SRNL recently received supernate samples retrieved from Tank 10H as well as in-tank batch contact samples for characterization in support of qualifying Batch 3 for processing through the TCCR unit. SRNL received and characterized a set of dip samples collected from Tank 10H (one surface and one variable depth sample). No solids were observed in either sample and therefore the samples were combined, and a suite of analyses were performed on the composite sample. The density of the combined sample was 1.174 g/mL (0.18 %RSD), which was consistent with the measured sodium concentration of 3.54 M (0.08 %RSD). The 137Cs activity in the sample was 4.75E+07 dpm/mL (4.28 %RSD) and the total Cs concentration was calculated to be 1.54 mg/L (1.16E-05 M). The in-tank batch contact samples consist of 0.1 g of crystalline silicotitanate (CST) contained within a teabag device. Duplicate samples were submerged in Tank 10H supernate for a period of about 13 days, after which time they were retrieved and transferred to SRNL for analysis. At SRNL the CST was rinsed to remove excess salt solution and the CST was then air dried before being digested for analysis. Results of the analysis indicated a 137 Cs loading of 3.66E+10 ± 1.40E+09 dpm/g or 16.5 ± 0.63 Ci/kg CST . This value represents a bounding upper limit as it includes the addition of two sigma uncertainty from replicate analysis of the individual teabag samples as well as the addition of the small amount of 137 Cs activity measured in the rinse solutions. The above values are based on the air-dried mass of CST. Correcting to the true dry mass using a F-factor of 0.8191 results in a maximum loading of 20.2 ± 0.77 Ci/kg CST .

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Epoxy / Silicone Potting Material Removal for Greater Recovery of Circuit Boards

This report documents an exploratory research project, “Epoxy/Silicone Potting Material Removal for Greater Recovery of Circuit Boards,” that investigated micro abrasion and laser ablation as potential cost effective methods for removal of epoxy and silicone potting material from printed circuit boards (PCBs) to expose components for testing and repair. The ability to cost effectively remove such PCB potting materials will improve PCB repair, remanufacture, and reuse opportunities with potential for sizable reductions in material use, E-waste, and energy. Laser ablation was identified as a method potentially warranting further investigation, in particular for epoxy based potting materials.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2022

Reduced Temperature Cesium Removal from AP-101 Using Crystalline Silicotitanate

The Tank Side Cesium Removal (TSCR) system, currently operational under Washington River Protection Solutions LLC (WRPS), sends initial low-activity Hanford waste tank supernate feed to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at the WTP. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIV™ R9140-B), was selected as the ion exchange media at TSCR. Laboratory-scale ion exchange processing using TSCR prototypic unit operations continues to contribute toward WRPS establishing accurate process flowsheets for the individual feed campaigns planned for TSCR. This report describes the small-scale ion exchange testing with 14.0 L of diluted and filtered supernate from tank 241-AP-101 (AP-101DF) at 16 °C (62 °F) to demonstrate processing at temperature conditions that are more prototypic of what the TSCR system could experience during colder seasons of the year. Since CST Cs capacity increases with decreasing contact temperature, testing at the lower operating temperature will help to predict the maximum 137 Cs loading onto the CST in the TSCR system. One of the waste acceptance criteria (WAC) for the WTP Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the AP-101DF tank waste to meet this criterion, only 0.144% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 694. Testing with AP-101DF matched TSCR prototypic operations where a lead-lag configuration was used until the lag column reached the WAC limit, then a polish column was brought online for continued processing in a lead-lag-polish column configuration. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. The Cs-decontaminated product was retained for vitrification testing (to be reported separately). The lead column reached 62% Cs breakthrough after processing ~1400 BVs of feed; the 50% Cs breakthrough occurred at 1250 BVs. Testing compared to previous AP-107 testing at 16 °C showed ~80 BV increases in volume processed to reach the WAC limit for both lead and lag columns. A similar slope in breakthrough curves for both tests indicates similar kinetic behavior, with variations in feed matrices (Na and Cs concentrations) likely responsible for the deviations in reaching the WAC limit. The Cs effluent from the lag column reached the WAC limit after processing 875 BVs. Anticipating this breakthrough point, the polish column was preemptively installed at 770 BVs. Cs breakthrough from the lag column began at 300 BVs, reaching 5.32×10 0 µCi/mL, or 5.6 % Cs breakthrough, after processing all 1400 BVs of feed. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Continuous Particle Monitoring and Removal for Molten Chloride CSP Systems

This report describes the technologies that were developed as part of the Continuous Particle Monitoring and Removal for Molten Chloride CSP Systems project that was funded by the EERE Solar Energy Technologies Office during FY2021. This project consisted of two main tasks: (1) the development electrical resistance tomography (ERT) sensors to provide particle sensing capabilities for CSP-relevant molten chloride salts, and (2) the development of vortex separation technologies to enable efficient removal of particles from the salt. Both of these tasks are crucial because, although the Gen3 CSP chloride heat transfer fluid is inexpensive and can achieve high temperatures, major issues with erosion corrosion and clogging can arise from the presence of MgO particles within the salt. The particle monitoring and removal technologies were both demonstrated over the course of this project. The ERT sensors were able to achieve their performance targets by providing accurate particle concentration measurements over a range from 0.0 to 8.0 wt%. The vortex separators were able to demonstrate separation of particles from molten salt streams, but the separation efficiencies were lower than the targeted level of 90%. With further development and deployment, these technologies will ultimately help to enable successful long-term operations of CSP systems that make use of high-temperature chloride salts.

14 SOLAR ENERGY↗
DOE OSTI2023

Cesium Removal from 5.5 and 7.0 M Na AP-105 Using Crystalline Silicotitanate

The Tank Side Cesium Removal (TSCR) system, currently operational by Washington River Protection Solutions LLC (WRPS), prepares initial low-activity Hanford waste tank supernate feeds for the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at the WTP as governed by a waste acceptance criterion (WAC). Specific to 137 Cs, this requirement is < 3.18E-5 Ci 137 Cs/mole of Na. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B), has been selected as the ion exchange media for TSCR. CST is a non-elutable inorganic material that has demonstrated robust chemical, physical, and radiation tolerance while maintaining functionality. However, testing to date on actual tank waste samples has been limited to Na concentrations between 5 and 6 M Na (Fiskum et al. 2019a, 2021b and Westesen et al. 2021a, 2021b, 2022) while actual tank conditions can reach upwards of 9 M Na. Testing with feed from Hanford tank AP-105 incorporated testing at both 5.5 and 7 M Na in order to evaluate the impact of Na concentration on volume of waste processed before reaching the WAC. A 6-L volume of 5.5 M Na AP-105 and an 8-volume of 7 M Na AP-105 was processed through the Radioactive Waste Test Platform system, established at Pacific Northwest National Laboratory to support small-scale waste qualification efforts. The columns consisted of 6- to 9-mL CST beds (CST Lot 2002009604, sieved to screen out >30-mesh particles) placed in 1.5-cm-inner-diameter columns. Feed was processed at 1.9 bed volumes (BV) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. Table S.1 and Figure S.1 summarize the measured AP-105 Cs load performance for each feed condition.

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

Cesium Removal from SY-101 Tank Waste Using Crystalline Silicotitanate

The newly operational Tank Side Cesium Removal (TSCR) system removes radioactive cesium-137 ( 137 Cs) and solids from tank waste supernate in the 200 east area of the Hanford site. Efforts to expand the removal capabilities to the 200 west area are underway by a system anticipated to be called the West Area Risk Management (WARM) system. Laboratory-scale ion exchange processing using expected WARM unit operations were conducted to contribute toward Washington River Protection Solutions (WRPS) establishing accurate process flowsheets for the individual feed campaigns planned for the west area supernate pretreatment. This report describes the small-scale ion exchange testing with 8.0 L of filtered supernate from tank 241-SY-101 (referred to as SY-101) at 16 °C (62 °F) to demonstrate processing conditions that would be prototypic of what the WARM system may experience. One of the waste acceptance criteria (WAC) for the Waste Treatment Plant (WTP) Low-Activity Waste Facility is that the waste must contain less than 3.18×10 -5 Ci 137 Cs per mole of Na. For the SY-101 tank waste to meet this criterion, only 0.44% of the influent 137 Cs concentration may be delivered to the WTP; this requires a Cs decontamination factor of 227. Testing with SY-101 matched current TSCR prototypic operations where a lead-lag configuration is used until the lag column reached the WAC limit. Neither the lead nor lag columns reached the WAC, so a polish column was never utilized during this test. Feed was processed at 1.9 bed volumes (BVs) per hour; the flowrate, in terms of contact time with the crystalline silicotitate (CST) bed, matched the current flowrate at TSCR (an expected flowrate to be used at WARM). The lead column only reached 0.09% Cs breakthrough after processing ~1363 BVs of feed. Cesium breakthrough from the lag column was not observed during the entire processing. Table S.1 and Figure S.1 summarize the observed column performance and relevant Cs loading characteristics. Table S.1 and Figure S.1 also summarize the measured SY-101 Cs load performance.

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

Finding Of Adverse Effect and Proposed Mitigation for the Removal of Joshua Trees at Building 23-210, Mercury, Area 23, Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) recently removed two fallen Joshua trees that were part of an Accessory Resource (AR) to the former Department of Defense (DOD) Motor Pool Maintenance Compound in Mercury (Nevada State Historic Preservation Office [SHPO] Resource No. C307) on the Nevada National Security Site (NNSS) in Nye County, Nevada (Figure 1). The Joshua trees were part of AR3, a landscaped feature at the entrance to the compound. There are two remaining Joshua trees at risk of falling which may require removal in the future. The NNSA/NFO is proactively consulting with the SHPO regarding the potential removal of the remaining two trees. The purpose of the undertaking is to ensure that dead or dying Joshua trees do not cause damage to the remaining landscape elements, nor injury to NNSS employees. The NNSA/NFO will implement this undertaking in accordance with the Programmatic Agreement between the National Nuclear Security Administration Nevada Field Office and the Nevada State Historic Preservation Officer Regarding Modernization and Operational Maintenance of the Nevada National Security Site, at Mercury in Nye County, Nevada, hereafter referred to as the Mercury PA. Building 23-210 was the DOD’s motor vehicle maintenance compound constructed in 1951. The compound included its own repair shop and fuel station. The town of Mercury and the immediate surrounding area have been formally determined eligible for listing in the National Register of Historic Places (National Register, NRHP) as the Mercury Historic District (MHD, SHPO Resource No. D230) under Criteria A and C for their importance in supporting nuclear testing and scientific research from 1951 through 1992. The former DOD Motor Pool Compound was identified as a contributing element to the MHD in a 2018 architectural survey of the district (Reno et al. 2018) and recorded on a Nevada Architectural Resource Assessment (ARA) form (Reno et al. 2017). The foundation for Building 23-210 was also identified in Appendix C of the Mercury PA as a Category III contributing element with three contributing ARs. Category III properties are those that may include elements for which there are several representatives in the MHD, such as foundations, and those elements which possess characteristics that are not unique to the MHD and are commonly found in other non-NNSS facilities. The former DOD compound, which is identified by the foundation for the Building 23-210 and its ARs, is a historic property for the purposes of compliance with Section 106 of the National Historic Preservation Act (NHPA) and subject to the stipulations of the Mercury PA.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Analysis of Second Target Station Target Segment Removal Dose Rates

This report documents the analysis of the dose rate fields in two configurations of the target system of the Second Target Station (STS) at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) in support of the remote handling target segment removal operation. The two configurations detailed in this report are both assuming 8 hours since beam was incident on the target. The first configuration assumes all shielding is in place just as it was during operation and allows for confirmation of hands-on maintenance capability of the target drive components. The second configuration assumes that a hatch in the target drive room ceiling has been opened to the high bay, the target segment removable core vessel shield block has been removed, and there is a direct line-of-sight to the target segment some 5 m below the target drive room elevation. Dose rate fields in both configurations are documented in this report along with details of the methods of computation. The dose rate fields included in this report are intended to support the engineering of remote handling special equipment and the planning of remote handing operations. The models and analysis detailed in this report are in support of the preliminary design of the STS and future analysis is needed to support the final design of the STS.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Towards Marine Carbon Dioxide Removal (mCDR) Centers of Excellence

A diverse portfolio of carbon dioxide removal (CDR) technologies will be essential to meeting climate goals while supporting sustainable development. In particular, marine carbon dioxide removal (mCDR) methods can help diversify the existing portfolio, which currently relies heavily on engineered direct air capture systems for large-scale atmospheric removals. To fulfill their potential in the CDR ecosystem, these early-stage mCDR technologies require additional research and development. Here, we propose regional mCDR Technology Centers of Excellence that will provide inventors and developers with access to right-scale facilities and engineering expertise that will meet their needs wherever they are on the journey from bench- to pilot-scale development. To support innovation across the technological readiness level spectrum, these Centers will provide collaborative access to ocean-based mCDR testing sites; other opportunities for partnering with multidisciplinary experts in technology development and commercialization; provide meaningful pathways for integrated engagement with local industrial and regulatory systems; and help in developing tailored, impactful community benefit models. Ideal sites are co-located in areas with a favorable natural environment for testing mCDR, well studied baselines, accessible infrastructure for multidisciplinary marine research, and technology commercialization support. As a case study, we offer the Pacific Northwest as particularly suitable for a regional mCDR Technology Center of Excellence given existing and potential growth of all the characteristics of ideal sites listed above, and which could especially benefit from the region’s growing marine climate technology sector. Last, we offer a brief whole-of-government perspective for supporting and regulating these Centers of Excellence, including identification of mCDR Science Centers of Excellence to ensure that the development of mCDR technologies dovetails with much-needed advancements in oceanographic observation and simulation infrastructure.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Cesium Removal from AP-106 Tank Waste Using Crystalline Silicotitanate

The Tank Side Cesium Removal (TSCR) system, currently operated by Washington River Protection Solutions, LLC, prepares Hanford waste tank supernate feeds for the Low-Activity Waste Facility at the Hanford Waste Treatment and Immobilization Plant (WTP). In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange (IX), allowing contact handling of the liquid effluent product at the WTP as governed by a waste acceptance criterion (WAC). Specific to 137 Cs, this requirement is <3.18E-5 Ci 137 Cs/mole of Na. Crystalline silicotitanate (CST), manufactured by Honeywell UOP, LLC (product IONSIV R9140-B), was selected as the ion exchange media for TSCR.

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

Mitigation for Removal of Building 01-103 in the Area 1 Subdock, Nevada National Security Site, Nye County, Nevada

Building 01-103, the Drill Bit Repair Building was originally recorded as SHPO Resource No. B18848 in 2020 during an architectural survey of the Area 1 Subdock (Collins et al. 2022). It was recommended individually eligible for listing in the NRHP under Criteria A and C and contributing to the significance of the Area 1 Subdock Historic District. The SHPO concurred with these determinations in a letter dated April 29, 2022 (Reed 2022). The Drill Bit Repair Building retains all seven aspects of integrity. The location, materials, workmanship, feeling, and association have not changed since its original construction. The design was modified slightly in 1996 with the addition of a removable and non-destructive chain-link fence. Nearby buildings 01-101 and 01-102 were removed (Collins and Menocal 2020) but the other buildings and storage yards remain. Therefore, the integrity of setting has been changed, but not to the degree that the building no longer conveys its significance (see Figures 4 and 5 for comparison). Building 01-103 was constructed in 1985 to coincide with the relocation of the Area 3 Subdock to Area 1. The design for the new Subdock provided a centralized cluster of buildings surrounded by two large storage yards. The buildings included administrative and engineering office space, warehouses, a radiograph building, and Building 01-103 as the drill bit maintenance and repair building, among others. The Drill Bit Repair Building was critical to the success of underground nuclear testing and is eligible under Criterion A for its contribution to broad patterns of history by maintaining specially designed equipment for underground nuclear testing at the NNSS. It is also eligible under Criterion C as it is the only recorded example of a facility built specifically for the maintenance of big-hole drill bits on the NNSS. As it retains both significance and integrity, it remains individually eligible under Criteria A and C. In addition to its individual significance, the Drill Bit Repair Building also contributes to the significance of the Area 1 Subdock Historic District as a representative of a big-hole drilling support facility necessary for underground nuclear testing. The building and the equipment therein were designed specifically for testing at the NNSS, where much of the early technology and methods for big-hole drilling were developed. Its removal will result in not only the demolition of a contributing element but also a change in the physical features of the district’s setting that contribute to its historic significance. The change in setting will diminish the district’s integrity. However, the district still contains numerous examples of other buildings constructed for the purpose of supporting big-hole drilling and remains eligible under Criteria A and C.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Improvements to Passive Heat Removal Systems in SAS4A/SASSYS-1

Safe operation of a sodium cooled fast reactor (SFR) can be significantly aided by heat removal systems that do not rely on active power, but function using passive means. Passive heat removal from such reactors can employ systems that extract heat directly from sodium coolant by utilizing heat exchangers submerged in the coolant and natural circulation of a metal fluid sodium (or sodium potassium eutectic) to heat exchangers (generally air cooled) externally located. These types of systems are generally called Direct Reactor Auxiliary Cooling Systems (DRACS). Another type of passive heat removal system particularly suitable for pool-type SFRs relies on indirect extraction of heat by utilizing air flow in an annular region surrounding the reactor guard vessel which discharges the heat via the chimney effects to the atmosphere. These systems are generally called Reactor Vessel Auxiliary Cooling Systems (RVACS).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Nuclear Power Plant Infrastructure Evaluations for Removal of Spent Nuclear Fuel

This report provides evaluations of the NPP site infrastructure and near-site transportation infrastructure for removing SNF from 19 NPP sites and the Morris Independent Spent Fuel Storage Installation (ISFSI). The material to be removed from the NPP sites includes both the SNF and the greater-than-Class C low-level radioactive waste (GTCC waste)3 that is stored, or will be stored, at the sites. This report is an update of the report Nuclear Power Plant Infrastructure Evaluations for Removal of Spent Nuclear Fuel (Maheras et al. 2021) and includes expansion of the site evaluations to include operating nuclear power plant (NPP) sites and to incorporate updated site inventory data. Figures that include the number of spent nuclear fuel (SNF) assemblies and metric tons heavy metal (MTHM) in a single figure have also been added to the report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗