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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Gas Retention and Release from Nuclear Legacy Waste - 16449

The retention of gas within corroded magnox sludge waste at Sellafield, UK and secondary reprocessing waste at Hanford, USA has significant economic and safety implications for decommissioning various nuclear legacy buildings, including the magnox swarf storage silos and first generation magnox storage pond. A series of laboratory scale gas retention tests within magnesium hydroxide soft sediments have revealed a 4-51 Pa yield stress range where consolidated beds could retain sufficient gas to become buoyant with respect to a water supernatant. Density inversion could lead to a Rayleigh-Taylor style instability which could result in an upward transfer of radioactive material from the consolidated bed. The applicable yield stress range suggests that such rollover events may occur in weaker sediment than previously hypothesized, based on current understanding of the fluidization and stable channel mechanisms for gas release from weak and very strong sediments respectively. X-ray computed tomography images of gas retained by 7 Pa yield stress soft sediment reveal both a stable foam layer at the top of the bed and regions dense with micro-bubbles which could provide pathways for gas transport through the bed. Extension of these pathways, hidden below the surface of the sediment, to the container walls and the foam layer could represent a novel mechanism for gas release from intermediate strength sediments of <100 Pa yield stress. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Long-Term Prediction of Radionuclide Emission from Submerged Objects in the Kara Sea

Analysis of the distribution of radionuclides in the water and bottom deposits in the bays of the Kara Sea attests that the radionuclide emission from the sources of radioactive materials dumped into the bays decrease over time. The equations of the transport of radionuclides in submerged objects into the marine environment beyond their limits were constructed on the basis of new results. Evaluations of the consequences of the long-term presence of submerged radiation-hazardous objects at the sea bottom depend on the temporal reduction of the time constant for the emission of radionuclides. The calculations show that taking account of this parameter with the formation of corrosion openings in closed submerged objects the emission of {sup 239}Pu from them into the ambient environment will almost cease after several hundreds of years (700 or more years). Neglecting this factor, the emission of {sup 239}Pu will continue to increase for 6000–12000 years, and the contamination of the bottom deposits by this isotope will be significant.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Occurrence of {sup 222}Rn in irrigation water from Wadi Al-Rummah Qassim province, Saudi Arabia

Naturally accruing radioactive materials in the environment have received attention since they may be present in high level and pose risk to human health. The present work deals with measuring of {sup 222}Rn in irrigation water samples from Wadi Al-Rummah, Qassim province, in central of Saudi Arabia. {sup 222}Rn concentrations were measured by RAD7. It was found that the concentration of {sup 222}Rn ranged from 2.1 ± 1.2 to 7.2 ± 1.5 BqL{sup −1}. These values are below 11.1 BqL{sup −1} the maximum contamination level recommended from the U.S. Environmental Protection Agency. The calculated annual effective dose (AED) ranging from 7.5 to 26.1 µSv/y. It was evident that the total annual effective dose resulting from radon in irrigation groundwater in Wadi Al-Rummah in Qassim area were significantly lower than the recommended limit 1 mSv/y for the public.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2012

BWR ex-vessel steam explosion analysis with MC3D code

A steam explosion may occur, during a severe reactor accident, when the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. To resolve the open issues in steam explosion understanding and modeling, the OECD program SERENA phase 2 was launched at the end of year 2007, focusing on reactor applications. To verify the progress made in the understanding and modeling of fuel coolant interaction key phenomena for reactor applications a reactor exercise has been performed. In this paper the BWR ex-vessel steam explosion study, which was carried out with the MC3D code in conditions of the SERENA reactor exercise for the BWR case, is presented and discussed. The premixing simulations were performed with two different jet breakup modeling approaches and the explosion was triggered also at the expected most challenging time. For the most challenging case, at the cavity wall the highest calculated pressure was {approx}20 MPa and the highest pressure impulse was {approx}90 kPa.s. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Annual Report for Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility – Fiscal Year 2011

As a condition to the Disposal Authorization Statement issued to Los Alamos National Laboratory (LANL or the Laboratory) on March 17, 2010, a comprehensive performance assessment and composite analysis maintenance program must be implemented for the Technical Area 54, Area G disposal facility. Annual determinations of the adequacy of the performance assessment and composite analysis are to be conducted under the maintenance program to ensure that the conclusions reached by those analyses continue to be valid. This report summarizes the results of the fiscal year 2011 annual review for Area G. Revision 4 of the Area G performance assessment and composite analysis was issued in 2008 and formally approved in 2009. These analyses are expected to provide reasonable estimates of the long-term performance of Area G and, hence, the disposal facility's ability to comply with Department of Energy (DOE) performance objectives. Annual disposal receipt reviews indicate that smaller volumes of waste will require disposal in the pits and shafts at Area G relative to what was projected for the performance assessment and composite analysis. The future inventories are projected to decrease modestly for the pits but increase substantially for the shafts due to an increase in the amount of tritium that is projected to require disposal. Overall, however, changes in the projected future inventories of waste are not expected to compromise the ability of Area G to satisfy DOE performance objectives. The Area G composite analysis addresses potential impacts from all waste disposed of at the facility, as well as other sources of radioactive material that may interact with releases from Area G. The level of knowledge about the other sources included in the composite analysis has not changed sufficiently to call into question the validity of that analysis. Ongoing environmental surveillance activities are conducted at, and in the vicinity of, Area G. However, the information generated by many of these activities cannot be used to evaluate the validity of the performance assessment and composite analysis models because the monitoring data collected are specific to operational releases or address receptors that are outside the domain of the performance assessment and composite analysis. In general, applicable monitoring data are supportive of some aspects of the performance assessment and composite analysis. Several research and development (R and D) efforts have been initiated under the performance assessment and composite analysis maintenance program. These investigations are designed to improve the current understanding of the disposal facility and site, thereby reducing the uncertainty associated with the projections of the long-term performance of Area G. The status and results of R and D activities that were undertaken in fiscal year 2011 are discussed in this report. Special analyses have been conducted to determine the feasibility of disposing of specific waste streams, to address proposed changes in disposal operations, and to consider the impacts of changes to the models used to conduct the performance assessment and composite analysis. These analyses are described and the results of the evaluations are summarized in this report. The Area G disposal facility consists of Material Disposal Area (MDA) G and the Zone 4 expansion area. To date, all disposal operations at Area G have been confined to MDA G. Material Disposal Area G is scheduled to undergo final closure in 2015; disposal of waste in the pits and shafts is scheduled to end in 2013. In anticipation of the closure of MDA G, plans are being made to ship the majority of the waste generated at LANL to off-site locations for disposal. It is not clear at this time if waste that will be disposed of at LANL will be placed in Zone 4 or if disposal operations will move to a new location at the Laboratory. Separately, efforts to optimize the final cover used in the closure of MDA G are underway; a final cover design different than that adopted for the performance assessment and composite analysis will likely emerge from that investigation. All of these changes will require re-examination of the assumptions upon which the performance assessment and composite analysis are based and, in all likelihood, revision of those analyses.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Vertical and horizontal distribution of radionuclides ({sup 232}Th, {sup 238}U and {sup 40}K) in sediment from Manjung coastal water area Perak, Malaysia

Distribution of radionuclides from anthropogenic activities has been widely studied in marine coastal area. Due to rapid population growth and socio-economic development in Manjung area such as coal fired power plant, iron foundries, port development, waste discharged from factories and agriculture runoff may contribute to increase in pollution rate. The radioactive materials from anthropogenic activities could deteriorate the quality of the marine ecosystem and thus lead to possible radiological health risk to the population. Radionuclides ({sup 232}Th, {sup 238}U and {sup 40}K) content in surface and profile sediment from Manjung coastal area was determined in this study. Radionuclides in sediment from seven locations were collected using sediment core sampling and measurements were carried out using Energy Dispersive X-ray Fluorescence (EDXRF) spectroscopy. The results show that the concentration of radionuclides in surface sediment and distribution trend of depth vertical profile sediment generally varies depending on locations. Enrichment factors (EF), geo-accumulation index (I{sub geo}) and pollution index (PI) were applied to determine level of pollution of this study area. The radiological risks related to human exposure were evaluated based on external hazard index (H{sub ex})

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Closure Device For a Containment Vessel in a Small Gram Quantity Shipping Package

SRNL has been chartered by DOE Environmental Management (DOE-EM) to develop a new Type B shipping package for Small Gram Quantities (SGQ) of radioactive material. Design a containment vessel (CV) closure device that will reduce the operation time and improve the ease-of- use for Operations Personnel. The Containment Vessel (CV) includes a closure lid cap with spring shaft and locking tabs, multiple disc springs, and vessel with integral locking ring. The CV is assembled by aligning the tabs on the lid cap with the openings in the locking ring. Next, the lid cap is pushed down to compress the disc springs. Finally, the lid is rotated 45-degrees to locking location. When released, the tabs are secured against the locking ring. Disc springs (e.g. Belleville washers, coned-disc springs, Belleville spring, etc.) exert axial loads on the mating parts when compressed. The orientation of the disc springs can be configured to change the deflection and axial loads for specific designs. In addition to securing the lid cap to the vessel, the compressed disc springs can be used to apply a load to a containment seal (e.g. o-ring). A prototype of the CV closure was 3D printed to evaluate the locking mechanism design and functionality. A test plan will be developed to evaluate the mechanical properties (i.e. compression forces) and functionality of the CV closure design.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Results from Evaluation of Proposed ASME AG-1 Section FI Metal Media Filters - 13063

High efficiency particulate air (HEPA) filtration technology is commonly used in Department of Energy (DOE) facilities that require control of radioactive particulate matter (PM) emissions due to treatment or management of radioactive materials. Although HEPA technology typically makes use of glass fiber media, metal and ceramic media filters are also capable of filtering efficiencies beyond the required 99.97%. Sintered metal fiber filters are good candidates for use in DOE facilities due to their resistance to corrosive environments and resilience at high temperature and elevated levels of relative humidity. Their strength can protect them from high differential pressure or pressure spikes and allow for back pulse cleaning, extending filter lifetime. Use of these filters has the potential to reduce the cost of filtration in DOE facilities due to life cycle cost savings. ASME AG-1 section FI has not been approved due to a lack of protocols and performance criteria for qualifying section FI filters. The Institute for Clean Energy Technology (ICET) with the aid of the FI project team has developed a Section FI test stand and test plan capable of assisting in the qualification ASME AG-1 section FI filters. Testing done at ICET using the FI test stand evaluates resistance to rated air flow, test aerosol penetration and resistance to heated air of the section FI filters. Data collected during this testing consists of temperature, relative humidity, differential pressure, flow rate, upstream particle concentration, and downstream particle concentration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A compact high-resolution X-ray ion mobility spectrometer

For the ionization of gaseous samples, most ion mobility spectrometers employ radioactive ionization sources, e.g., containing {sup 63}Ni or {sup 3}H. Besides legal restrictions, radioactive materials have the disadvantage of a constant radiation with predetermined intensity. In this work, we replaced the {sup 3}H source of our previously described high-resolution ion mobility spectrometer with 75 mm drift tube length with a commercially available X-ray source. It is shown that the current configuration maintains the resolving power of R = 100 which was reported for the original setup containing a {sup 3}H source. The main advantage of an X-ray source is that the intensity of the radiation can be adjusted by varying its operating parameters, i.e., filament current and acceleration voltage. At the expense of reduced resolving power, the sensitivity of the setup can be increased by increasing the activity of the source. Therefore, the performance of the setup can be adjusted to the specific requirements of any application. To investigate the relation between operating parameters of the X-Ray source and the performance of the ion mobility spectrometer, parametric studies of filament current and acceleration voltage are performed and the influence on resolving power, peak height, and noise is analyzed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Development of CFD Flow Analysis for Nuclear Facility Ventilation Systems

Understanding and modeling of particulate transport has importance related to safety in ventilated environments that contain hot cells, gloveboxes, and fume hoods. In particular, good modeling of the airflow in these spaces is required to understand areas of stagnation where particles drop out of the free stream and accumulate. This accumulation of particulate can cause either contamination and/or higher than acceptable local dose levels. It is known that multiple sites across the DOE complex take credit for hot cells, gloveboxes, and/or hoods in their safety basis for providing a defense-in-depth benefit for both onsite and offsite releases. By providing confinement of radioactive materials, such features serve to reduce direct doses to facility workers and mitigate the consequences to the environment due to an uncontrolled release. Each of these features has access points that interface with the personnel space. Understanding how air flow behaves at these access points is of great interest to those performing hazard analysis. Computational Fluid Dynamics (CFD) can be used to analyze and model the flows in hot cells and glove boxes as a way to confirm operation. While CFD capability continues to advance, there are still important modeling assumptions that are left to the analyst's discretion. These are most notably the choice of the turbulence models, mesh structure, and wall boundary condition assumptions used in the model. In this work, we apply CFD in modeling airflow through an occupiable work area of the Alpha Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory (ANL). This modeling is then compared to field test data to both inform and verify the modeling assumptions. Based on the results of field tests, modeling assumptions and boundary conditions are refined and the process is repeated until the results are found to be reliable with a high level of confidence.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Low-Altitude Unmanned Aerial Vehicles as a Tool for the Remediation of Radiologically Contaminated Environments - 16257

Following the events at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011, significant quantities of radioactive material were released into the local Japanese as well as the wider global environment. At the fifth anniversary of the incident, much work and expense is still being dedicated to the remediation of a large area of eastern Japan, contaminated primarily with radio-caesium. Due to the complex topography of the geographical area of Japan effected, it is important to understand contaminant evolution/migration at the greatest possible resolution - previous methods have lacked this. An increase in the spatial resolution of radiation mapping over other previous methods has been achieved through the deployment of an unmanned aerial vehicle (UAV). The platform used during this work combined a low-altitude multi-rotor UAV with a lightweight radiation mapping system to achieve sub-meter resolution. Using this system it was possible to measure the distribution of radionuclide contamination at a number of sites within the Fukushima Prefecture region of Japan. Unlike ground-based surveys conducted on foot by humans that attain a comparable resolution, such a system eliminates the potentially significant dose that would otherwise be received, as well as the influence of shielding on results. In addition to providing a rapid and high resolution response to radiological incidents, such monitoring technology has the potential to be a powerful tool in post-operational clean-out (POCO) at nuclear sites undergoing decommissioning worldwide. (authors)

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

Verification of severe accident management guideline (SAMG) entry condition for OPR1000 - 367

The current severe accident prevention and mitigation strategies for OPR 1000 have emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs), respectively. The main objective of SAMG is to prevent a release of radioactive materials from releasing into environment and mitigate severe accident phenomena during severe accident. The current SAMG entry condition is met when the core exit temperature (CET) reaches 650 deg. C in OPR 1000. Peak cladding temperature (PCT), which represents core integrity more precisely than CET does, cannot be used since it is not measurable in the main control room (MCR). Current SAMG entry condition for every kind of accident scenarios is not reasonable, since entry condition is determined without considering operator action time. Therefore, SAMG entry condition need to be verified with consider operator action time. Thus, an analysis methodology for SAMG entry condition was developed to verify the SAMG entry condition. A severe accident DB were also developed in order to analyze the SAMG entry condition. The analysis results of DB show that available operator action time from the SAMG entrance to RV fail is shorter than the required time in most cases. As a result, the proper SAMG entry condition was suggested through the results of DB analysis. When the SAMG entry condition is changed from CET to suggested condition, the available action time is increased about 350 to 4300 sec. Also, the suggested entry condition was verified through the simulation which considered operation action time and reflecting the suggested entry condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

The development and initial testing of the Ice Pig cleaning method for nuclear reprocessing plants - 15363

There are numerous situations throughout the nuclear industry where it would be advantageous to clean out enclosed systems, e.g. ducts and pipes, to remove sediments and radioactive contamination. However, conventional flushing methods result in the production of significant volumes of contaminated liquor. The wall Shear achieved with water flushing is insufficient to remove and carry dense deposits often found in nuclear facilities. Conventional mechanical pigs can be propelled through the duct by application of a driving medium such as compressed air or water. These are able to achieve very high wall shears, but can also get physically stuck where there is variation in the pipe / duct diameter or small radius bends. The field of application is best suited to straight constant diameter systems with no discontinuities such as Tee pieces, valves and in-pipe instrumentation. Driven by the desire to clean and displace radioactive materials from complex topology ducts and pipes, without the risk of having conventional pigs becoming lodged in a plant and coupled with the need to minimize effluent volumes, the UK Nuclear Industry led by Sellafield Limited (on behalf of the UK Nuclear Decommissioning Authority), has worked to demonstrate and underpin the use of the ice pigging technology for the nuclear industry. It is found that high ice fraction slurries flow as viscous liquids that exhibit a number of mechanical properties that offer much higher shear rates. The 'Ice Pig' exerts shear rates on pipe walls which are typically 1000 times greater than those achieved with water flowing at the same velocity. The 'Ice Pig' is able to negotiate almost any topology, flow through pipe networks of vastly differing diameter, through heat exchangers and in-line instrumentation etc. Experimental data and operational experience from other industries demonstrates that whilst there is high shear, the 'Ice Pig' does not damage the surface of the pipe wall. A further attractive characteristic of the 'Ice Pig' is that it eventually melts rendering it a simple effluent for processing by downstream processes. Due to the special nature of ice slurry, its unique rheological and physical characteristics, together with the complementary underpinning experimental work undertaken by Sellafield Limited enables the Ice Pigging process to move from an 'interesting' technology to one suitable for use by the nuclear industry for operations, Post Operational Clean Out (POCO) and decommissioning. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Update of Erosion Conditions and Assessment of Mitigation Options for the L-Bar Site

The L-Bar former uranium milling site was remediated and closed per regulations and is now under the long-term stewardship of the U.S. Department of Energy, Office of Legacy Management (DOE/LM). Since assuming stewardship in 2004, DOE/LM has observed erosion and sedimentation processes at the L-Bar Site that will challenge meeting the design closure criteria in the long term, and therefore erosion control structures were installed in several areas. Portions of these structures were subsequently damaged by summer convective storm events and they are being redesigned. At the request of DOE/LM, Desert Research Institute (DRI) reviewed the proposed redesign plans of the erosion control structures. The purpose of this review is to support DOE/LM decisions regarding mitigation of ongoing erosion processes, including assessing possible repair of the existing erosion control structures, as well as plan for effective long-term site management. This design review is performed in reference to the Uranium Mill Tailings Radiation Control Act (UMTRCA) standards contained in 40 CFR 192.02 that require containment of residual radioactive material be effective for up to 1,000 years (if achievable, but certainly for 200 years), limit radon release, protect groundwater, and minimize future maintenance. The erosion control structures under review are not subject to the UMTRCA standard directly, but they contribute to the performance of the overall containment system by virtue of their relationship with the large diversion channels bounding the disposal cell. The 11 design drawings included in “Erosion Control Alternatives Analysis, L-Bar, New Mexico, Disposal Site, LTS-111-0043-08-002,” dated January 11, 2019, and listed in Appendix A, and supporting calculations in Weston (2019) are the focus of the review. Additionally, DOE/LM provided supporting data such as a 2018 photogrammetry-based digital elevation model of the site, precipitation records collected from an on-site weather station, and sedimentation records collected at an on-site basin. This assessment includes review of the proposed design to mitigate current site conditions but also addresses site changes between the initial erosion assessment report (Miller et al., 2012) and current conditions. The analysis is at a predesign engineering level and does not include development of hydraulic or sediment transport models. Site precipitation data are included in Miller et al. (2012). The natural conditions of the site, the design of site diversion systems, observed erosion at the site, and predesign engineering analysis are used to develop a holistic view of the challenges to L-Bar Site stability and possible responsive actions.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Remediation of Occupied Commercial Property Subject to Widespread Radium-226 Contamination: Update from WM2012 Paper - 16652

Amec Foster Wheeler was contacted by a company that managed commercial office space. High Rn-222 measurements had been observed throughout the facility and the landlord had been advised to commission a radiological survey of the site. The site had been purchased by the client in the 1990's. Initial desk studies found that the building had operated for around 50 years as a compass factory. Non-intrusive investigation identified widespread Ra-226 contamination. Ra-226 was found in the fabric of the building, in attic spaces, buried under floor boards and underlying car parks. Intrusive investigation was undertaken to estimate volume(s) of waste, waste categories, activity concentrations and the total inventory of radioactive materials on site. This work identified the presence of 180 GBq of Ra-226 on site. A programme of work is currently underway to remediate the site tackling areas posing the greatest risk to site occupants as a priority. We have worked closely with Regulators, our client, and tenants, to decontaminate the fabric of the building whilst areas of the building remain occupied. Early decontamination works identified that radon levels within the building were stubbornly high despite the removal of radium contamination. An independent radon source was suspected. A contaminated drain was discovered leading to a blocked-off river race running under the old mill building. A significant quantity of radium contaminated sludge was identified within the river race. This was believed to be the source of the excessive radon concentrations within the building. The paper will present work undertaken to investigate and remediate the inventory of sludge from beneath the building. It will describe the operational difficulties of working within a dynamic environment impacted by high dose rates, high radon concentrations, water ingress and waste retrieval challenges. The paper will present the radiological and environmental impacts of the work and detail the achieved end points. This was a challenging project working in confined spaces impacted by physical and radiological constraints. The remedial action undertaken significantly reduced radon concentrations in the building metres above the river race. It also significantly reduced the likelihood of radium contamination being mobilised into the river and the wider environment. The achieved end point permitted the building to return to uncontrolled commercial use. The end point achieved within the river race will ensure a safe working environment for future regular structural survey work. (authors)

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

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

ASME AG-1 Section FC Qualified HEPA Filters; a Particle Loading Comparison - 13435

High Efficiency Particulate Air (HEPA) Filters used to protect personnel, the public and the environment from airborne radioactive materials are designed, manufactured and qualified in accordance with ASME AG-1 Code section FC (HEPA Filters) [1]. The qualification process requires that filters manufactured in accordance with this ASME AG-1 code section must meet several performance requirements. These requirements include performance specifications for resistance to airflow, aerosol penetration, resistance to rough handling, resistance to pressure (includes high humidity and water droplet exposure), resistance to heated air, spot flame resistance and a visual/dimensional inspection. None of these requirements evaluate the particle loading capacity of a HEPA filter design. Concerns, over the particle loading capacity, of the different designs included within the ASME AG-1 section FC code[1], have been voiced in the recent past. Additionally, the ability of a filter to maintain its integrity, if subjected to severe operating conditions such as elevated relative humidity, fog conditions or elevated temperature, after loading in use over long service intervals is also a major concern. Although currently qualified HEPA filter media are likely to have similar loading characteristics when evaluated independently, filter pleat geometry can have a significant impact on the in-situ particle loading capacity of filter packs. Aerosol particle characteristics, such as size and composition, may also have a significant impact on filter loading capacity. Test results comparing filter loading capacities for three different aerosol particles and three different filter pack configurations are reviewed. The information presented represents an empirical performance comparison among the filter designs tested. The results may serve as a basis for further discussion toward the possible development of a particle loading test to be included in the qualification requirements of ASME AG-1 Code sections FC and FK[1]. (authors)

42 ENGINEERING↗
DOE OSTI2023

The influence of temperature history and flow mixing on the vapor-phase speciation of uranium oxide nanoparticles

The transport of radioactive material following a nuclear event depends heavily on fallout characteristics (e.g., composition, shape, and size), which in turn are controlled by the conditions under which the fallout forms. To this end, we use a novel plasma flow reactor (PFR) to explore how conditions including temperature history and gas mixing influence the chemical speciation and physical characteristics of uranium oxide nanoparticles. The PFR consists of a glass tube attached to an inductively coupled plasma (ICP) torch via an adaptor piece (ring flow injector) located downstream of the plasma-generating RF coil. We increase the gas flow rates from normal ICP operating conditions to higher values both upstream at the RF coil and downstream at the ring flow injector to create four distinct temperature profiles and two distinct gas flow mixing conditions (normal and higher Ar atomic ratios). Ex situ transmission electron microscopy (TEM) analysis is used to determine the crystal structure, morphology, and size distribution of the synthesized particles as a function of temperature and mixing. The particles display a striking dependence on these conditions in terms of the resulting speciation of uranium oxide: lower temperature (higher upstream) and slower cooling (normal downstream with normal Ar atomic ratio mixing) favor the formation of α-UO 3 particles, while faster cooling (higher downstream with higher Ar atomic ratio mixing) and higher temperature (normal upstream) favor the formation of UO 2 particles. The extent of gas mixing dictates particle aggregation: better mixing encourages the particles to aggregate more extensively. Here these results demonstrate the sensitivity of uranium to initial local conditions and can be used to improve our understanding of the kinetics that drive uranium speciation and ultimately inform fallout transport models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗