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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

High-ambition climate action in all sectors can achieve a 65% greenhouse gas emissions reduction in the United States by 2035

Under the next cycle of target setting under the Paris Agreement, countries will be updating and submitting new nationally determined contributions (NDCs) over the coming year. To this end, there is a growing need for the United States to assess potential pathways toward a new, maximally ambitious 2035 NDC. In this study, we use an integrated assessment model with state-level detail to model existing policies from both federal and non-federal actors, including the Inflation Reduction Act, Bipartisan Infrastructure Law, and key state policies, across all sectors and gases. Additionally, we develop a high-ambition scenario, which includes new and enhanced policies from these actors. We find that existing policies can reduce net greenhouse gas (GHG) emissions by 44% (with a range of 37% to 52%) by 2035, relative to 2005 levels. The high-ambition scenario can deliver net GHG reductions up to 65% (with a range of 59% to 71%) by 2035 under accelerated implementation of federal regulations and investments, as well as state policies such as renewable portfolio standards, EV sales targets, and zero-emission appliance standards. This level of reductions would provide a basis for continued progress toward the country’s 2050 net-zero emissions goal.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Determination of Seven Organic Impurities in FD&C Yellow No. 6 by Ultra-High-Performance Liquid Chromatography

The U.S. Food and Drug Administration (FDA) batch certifies FD&C Yellow No. 6 (Y6) to ensure that the color additive meets requirements published in the Code of Federal Regulations (CFR), including specifications for seven organic manufacturing impurities consisting of two intermediates, a reaction by-product, an impurity originating from an intermediate, and three subsidiary colors. An ultra-high-performance liquid chromatography (UHPLC) method was developed and validated for determining seven organic impurities in Y6. For comparison, the currently used high-performance liquid chromatography (HPLC) method was also validated. The new UHPLC method uses a 1.7 µm particle size biphenyl column with aqueous ammonium formate and methanol as eluants. Analytes are identified by comparing their retention times and UV-visible spectra to those of reference standards. Calibration is performed in the presence of the dye matrix and analyte levels are determined from their peak areas. UHPLC and HPLC validation studies obtained linear calibration curves and excellent values for limits of detection, limits of quantitation, recovery, repeatability, and intermediate precision for all analytes. Survey analyses of 30 samples using the UHPLC and HPLC methods yielded results that were consistent within experimental error; however, the UHPLC method provided more accurate results for two analytes that coelute using the HPLC method. The UHPLC method satisfies the accuracy and precision requisites for routine certification of Y6. The UHPLC method is faster and generates less waste than the HPLC method while delivering better sensitivity, separation, and accuracy.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Internal Radiation Dose Evaluation For An Unruptured Post Release Tristructural Istropic Fuel Particle for Advanced and Micro-reactor Applications

There are unique benefits from advanced/micro reactor designs and fuel types that offer safety features in the case of an accident that may reduce environmental consequences compared to conventional reactors and fuels. Several micro reactor designs, for both civilian and military applications, are proposed using tristructural-isotropic (TRISO) fuel. TRISO particles are a robust advanced nuclear fuel type that leads to the unique question of how unruptured, activated TRISO particles will interact with humans. Human exposure to nuclear fuel, or radionuclides, includes all pathways of exposure. However, TRISO particles are 0.9 mm in size, and that particle size restricts internal dose assessment to the ingestion pathway; eliminating the inhalation and dermal absorption pathways. Furthermore, activity of the TRISO particle was established by High Temperature Engineering Test Reactor simulations assuming the TRISO particle was one day post-release from the reactor. International Commission of Radiological Protection publications 100 and 133 were used to develop the internal dosimetry model for an unruptured TRISO particle for the International Commission of Radiological Protection reference male and female dose assessment. The rectosigmoid colon had the highest internal organ dose for both reference male (2.1 Sv) and reference female (2.3 Sv). The internal dose from ingestion of the scenario-defined TRISO particle was 0.25 Sv for the reference male and 0.29 Sv for the refence female, which exceeds the annual occupational effective dose limit of 0.05 Sv in the code of federal regulations, 10.CFR.20 Subpart C.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Modification in Applying Appendix D of 40 CFR Part 61 to Heated Solid Radionuclide Materials With High Melting and Boiling Points

Appendix D of Title 40 Part 61 of the US Code of Federal Regulations (CFR) provides a procedure that US Department of Energy (US DOE) facility owners and operators can use to estimate radionuclide emissions to the atmosphere for dose calculations instead of measuring emissions for minor sources under the 40 CFR Part 61, Subpart H, National Emission Standards for Emissions of Radionuclides Other Than Radon From Department of Energy Facilities, regulation. The procedure assumes that any radioactive material heated above 100 °C is completely vaporized and emitted to the atmosphere. In 1991, the US DOE Oak Ridge Reservation (ORR) requested approval to use different release fractions (RFs) for uranium because of its high melting and boiling points. In response to the request, the US Environmental Protection Agency (US EPA) Region IV approved the use of modified RFs for elemental uranium provided no reaction had taken place to alter its chemical form. In 2015, the ORR requested approval to use different RFs for tungsten, again because of its high melting and boiling points. EPA Region IV approved the use of modified RFs for heated radioactive tungsten metal. In accordance with the two precedents set for heating uranium and radioactive tungsten metals, in 2016, the ORR requested approval to use modified RFs in a similar fashion for other radioactive solid metals and compounds with melting and boiling points above 500 °C that might be heated above 100 °C in future research projects and experiments. EPA Region IV again granted approval to use modified RFs for the list of compounds. This note discusses the proposed modified RFs and their development.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Structural Analysis Approach for the Defense Programs Package 3 (DPP-3)

The Pacific Northwest National Laboratory (PNNL) is the design authority for a new Type B hazardous materials transportation package designated as the Defense Programs Package 3 (DPP-3) for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA). The DPP-3 has been developed using similar materials and fabrication methods employed in previous U.S. Nuclear Regulatory Commission (NRC), DOE, and NNSA certified packages. The DPP-3 design criteria are derived from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), NNSA guidance and NRC regulatory guides in order to safely and securely transport a variety of payloads. Final regulatory approval by the NNSA will require physical testing to demonstrate that the containment vessel (CV) remains leaktight after enduring the entire regulatory testing sequence prescribed in Title 10 of the Code of Federal Regulations Part 71 (10 CFR 71). In order to gain confidence that the DPP-3 will remain leaktight after testing, the DPP-3 has been structurally analyzed using the Finite Element Analysis (FEA) software LS-DYNA. The FEA analyses serve two general purposes: first, they aid in design and development of the package, and second, they advise as to which drop orientations are expected to cause the most damage during regulatory testing. This paper will discuss how the design criteria are incorporated into analytical techniques needed to evaluate the FEA structural simulation results for 10 CFR 71 conditions to give confidence the DPP-3 testing campaign will be successful.

Sakalaukus Jr., Peter J.↗
DOE OSTI2022

Large-scale parametric modeling of spent nuclear fuel dynamics in the 30 cm package drop scenario

Packages used to transport spent nuclear fuel (SNF) are required by the U.S. Code of Federal Regulations 10 CFR 71.71 to demonstrate satisfactory performance during a drop scenario. While the CFR is meant to ensure safe package function, it does not evaluate survival of the SNF within. The U.S. Department of Energy Spent Fuel and Waste Science and Technology program is working on closing the knowledge gap related to the response of SNF to external mechanical loads, including the hypothetical 30 cm package drop scenario in the CFR. In support of this effort, LS-DYNA finite element simulations were developed by Pacific Northwest National Laboratory (PNNL) to model generic drop scenarios at both the package and fuel assembly level. The models were validated against one-third scale package and full scale fuel assembly drop test data and were exercised to predict fuel cladding strains in a narrow range of model configurations. This work describes a large-scale parametric study conducted by PNNL using the previously developed and validated PWR finite element model, with the addition of a new generic BWR assembly model. The motivation for the parametric study was to characterize the broad range of SNF responses in the 30 cm package drop scenario. This was accomplished by varying the drop orientation, fuel assembly type (17x17 PWR and 10x10 BWR), burnup, cladding temperature, spacer grid buckling load, package mass, impact limiter stiffness, and mechanical gap conditions within the basket. A MATLAB framework was developed to automate LS-DYNA model generation and execution on PNNL institutional computing resources. In total, over 2000 simulations were performed. For each simulation, the SNF response was quantified in terms of permanent grid deformation, fuel rod contact pressure, and strains within the fuel rods, guide tubes, and water rods. The results provide valuable insight into the range of responses that could be reasonably expected from SNF in the hypothetical drop scenario, as well as the sensitivity to each input parameter. The results of this parametric study are a key component of the testing and modeling strategy the Spent Fuel and Waste Science and Technology program is using to close the external loads knowledge gap.

Kadooka, Kevin↗
DOE OSTI2023

Supplemental Structural Analyses Used in Support of Certification of the Defense Programs Package 3

The Defense Programs Package (DPP)-3 is to be certified by the National Nuclear Security Administration (NNSA) Packaging and Transportation Division (PTD). Certification is based on successful physical testing of the package, and six certification test units (CTUs) have been subjected to the normal conditions of transport (NCT) and hypothetical accident conditions (HAC) outlined in Title 10 of the Code of Federal Regulations Part 71 (10 CFR 71). During regulatory testing of one of the CTUs, the drum lid closure bolt closest to the impact surface failed during the HAC 30-ft side drop. One of the testing goals for this CTU was to minimize the thermal pathway from the drum exterior shell to the containment vessel (CV) flange and O-rings, leading to the maximum damage scenario in preparation for thermal testing at 1475 °F. Therefore, the puncture bar impact proceeded as originally planned by striking the side wall of the package exterior closest to the location of the CV flange. This puncture bar impact did not attempt to exploit the region of the failed drum lid closure bolt to potentially cause more damage to the package. After the conclusion of regulatory testing, the DPP-3 package finite element model was used to numerically evaluate the effects of puncture bar impacts to the drum exterior in the vicinity of the failed drum lid closure bolt to further demonstrate the robustness of the DPP-3 design. Three alternate puncture bar impact locations and orientations were evaluated to demonstrate that these puncture bar impacts would not substantially reduce the effectiveness of the DPP-3 packaging. This paper describes the supplemental finite element models performed and structural analysis results that demonstrate the alternate HAC 40-in. puncture bar impact tests would have no deleterious effect on the performance of the DPP-3.

Sakalaukus Jr., Peter J.↗
DOE OSTI2024

The Completion of Surrogate Testing for Low-Temperature TEX and a Look Towards the Future

To address the mounting need for below room temperature nuclear data validation, the Low-Temperature Thermal Epithermal eXperiments (LT-TEX) have been designed. Validation of low-temperature neutron cross sections is necessary to verify any operation at temperatures below room temperature which is typically observed in environments far from the equator. For example, a fissile material transportation truck may routinely observe ambient temperatures down to -40°C, which is the lower temperature bound of the normal conditions of transportation defined in the United States Title 10 Code of Federal Regulations §71.71c2. Additionally, sub-room temperature benchmarks can validate newly produced cross sections, that include novel thermal scattering laws, from North Carolina State University.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Chemical Waste Landfill Annual Post-Closure Care Report, Calendar Year 2019

The Chemical Waste Landfill (CWL) at Sandia National Laboratories/New Mexico (SNL/NM) is a remediated hazardous waste landfill that underwent closure in accordance with Title 20, Chapter 4, Part 1 of the New Mexico Administrative Code (20.4.1.600 NMAC), incorporating Title 40, Code of Federal Regulations (CFR), Part 265, (40 CFR § 265) Subpart G, and the CWL Final Closure Plan (SNL/NM December 1992 and subsequent revisions). The CWL Post- Closure Care Permit (PCCP) (NMED October 2009), which became effective June 2, 2011 (Kieling June 2011) and as modified, defines all post-closure requirements. This ninth CWL Annual Post-Closure Care Report documents all activities and results as required by the PCCP Attachment 1, Section 1.12.

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

NESHAPs CY19 Report: Radionuclide Air Emission Report for 2019

Lawrence Livermore National Security, LLC operates facilities at Lawrence Livermore National Laboratory (LLNL) in which radionuclides are handled and stored. These facilities are subject to the U.S. Environmental Protection Agency (EPA) National Emission Standards for Hazardous Air Pollutants (NESHAPs) in Code of Federal Regulations (CFR) Title 40, Part 61, Subpart H, which regulates radionuclide emissions to air from Department of Energy (DOE) facilities. Specifically, NESHAPs limits the emission of radionuclides to the ambient air to levels resulting in an annual effective dose equivalent of 10 mrem (100 μSv) to any member of the public. Using measured and calculated emissions, and building-specific and common parameters, LLNL personnel applied the EPA-approved computer code, CAP88-PC, Version 4.0.1.17, to calculate the dose to the maximally exposed individual member of the public for the Livermore Site and Site 300. In 2019, LLNL maintained its compliance with 40 CFR 61, Subpart H. All radioactive air emissions resulted in calculated doses far below the annual 10 mrem (100 μSv) sitewide standard. The annual doses to the site-wide maximally exposed individual member of the public at the Livermore Site and Site 300 from planned and unplanned operations in 2019 are: Livermore Site: 4.3 x 10⁻³ mrem (4.3 x 10⁻² μSv); Site 300: 9.5 x 10⁻⁸ mrem (9.5 x 10⁻⁷ μSv).

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

Proposed guidance for preparing and reviewing a molten salt non-power production or utilization facility application

Development of non-power molten salt reactors (MSRs) are under consideration to further establish an MSR experience base, support the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.43(e), and provide any additional analyses needed for development of a full-scale MSR. Guidance provided in this report is based on MSRs operating with liquid fuel (i.e., fuel dissolved within a molten salt). These reactors, unless owned by the DOE or DOD, will require licensing by the US Nuclear Regulatory Commission (NRC) staff. Standard review plan (SRP) guidance for large light water reactors (LWRs) is available in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants; Light Water Reactor (LWR) Edition. However, NRC staff observed that NUREG-0800 is very cumbersome to apply to non-power reactors “because of the great differences in complexity and hazards between non-power reactors and nuclear power plants.” Therefore, a program to develop performance-based guidance applicable to non-power reactors was initiated. In 1996, NUREG-1537, Parts 1 and 2, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, was published. Part 1, the format and content guide, suggests a uniform format for presenting information in non-power reactor applications that is acceptable to the NRC staff, but conformance with the format and content is not required. Part 2, the SRP, ensures the quality and uniformity of the staff review of an application. Unfortunately, the application guidelines and SRP do not provide adequate guidance for all advanced non-LWR technologies and applications. This discrepancy eventually led to the 2012 development of interim staff guidance (ISG) for NUREG-1537, which includes criteria for describing and reviewing aqueous homogeneous reactors (AHRs). Specifically, NUREG-1537 ISG, 2012 expanded the original document to address three areas: 1. updated criteria for heterogeneous non-power reactors, 2. criteria for licensing AHRs, and 3. criteria for licensing a Part 50-licensed isotope production facility. In 2015, the US Department of Energy (DOE) opted to build on the AHR NUREG-1537 ISG experience by performing a gap analysis of the guidance that would be used to license a non-power MSR. MSRs represent one of the advanced non-LWR technologies selected by DOE for development through a multiyear cost share award with Southern Company Services. Under this Advanced Reactor Concepts 2015 (DOE Advanced Reactor Concepts [ARC] 15) award program, the DOE tasked Oak Ridge National Laboratory (ORNL) to evaluate the guidance changes that the NRC may need to consider when licensing an MSR non-power reactor. ORNL staff, with support from Boston Government Services, LLC, focused on five system-related chapters in NUREG-1537 that were considered most relevant to inform the effort that would be required for a non-power MSR applicant. ORNL documented this review in a technical report, ORNL/TM-2018/834, Proposed Guidance for Preparing and Reviewing Molten Salt Non-Power Reactor License Applications (NUREG-1537). The report was subsequently shared with industry and the NRC. The 2018 review was limited in scope, focusing on key system chapters based on the expected significance of each chapter relative to expected differences in addressing advanced non-LWR technologies, specifically non-power MSRs, compared with heterogeneously fueled non-power reactors. In the ORNL report, proposed generic adaptations were suggested for the following NUREG-1537 chapters: Chapter 4, “Reactor Description”; Chapter 5, “Reactor Coolant Systems”; Chapter 6, “Engineered Safety Features”; Chapter 9, “Auxiliary Systems”; Chapter 11, “Radiation Protection Program and Waste Management” The inclusion of Chapter 11 in the previous review effort was intended to provide guidance for categorizing the waste-handling process for an MSR operating with homogenous fuel. The introductions from Parts 1 and 2 of the 2012 NUREG-1537 ISG provide guidance for the application and review of production facilities. After a period of operation, non-power MSRs with homogenous fuel will include gaseous and soluble fission products. The gaseous fission products will be collected and held for decay in an off-gas system. There might also be an initiative to polish or filter the soluble fission products in the fuel salt by some mechanical or chemical means. The treatment and handling of fission products in the non-power MSR fuel salt and the description of this process in the safety analysis report (SAR) must be very precise to avoid the waste treatment facility being construed as a co-located special nuclear material (SNM) fuel cycle facility (see Section 2.3 of this report). Subsequent to the release of ORNL/TM-2018/834, NRC staff expressed a desire to continue the regulatory gap analysis that was begun in that report. This would provide additional clarity and information addressed in certain sections of the original report, while also providing new guidance on certain topics not addressed in the original report. This revision would benefit the NRC staff reviewing applications involving non-power MSR designs and would help developers understand how the NRC staff might approach the review of such applications. The focus of this report is to provide infrastructure support to the NRC staff for the regulatory review of non-power MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Los Alamos National Laboratory Hazardous Waste Facility Permit Community Relations Plan

The requirements of the Hazardous Waste Facility Permit are fulfilled through execution of LANL's fourth strategic goal: “Enabling mission delivery through next-generation facility, infrastructure, and operational excellence.” With input from our stakeholders, we will continue to fulfill those requirements and to improve our waste operations. In 2015, the Laboratory purpose statement focused waste management efforts by directing delivery of mission success through operational effectiveness and scientific excellence. This Plan advances that purpose through community involvement. In the process of achieving our national security mission, Los Alamos National Laboratory generates some hazardous and mixed waste. Hazardous waste is solid waste that is dangerous or potentially harmful to human health or the environment. Hazardous wastes can be liquids, solids, gases, or sludges. They can be discarded as commercial products, such as cleaning fluids or pesticides, or as the byproducts of operations. Specific substances are listed in 40 Code of Federal Regulations (CFR), Part 261: 40 CFR Part 261. Hazardous waste management activities are regulated by the U. S. Environmental Protection Agency and the New Mexico Environment Department (NMED) pursuant to New Mexico Hazardous Waste Act (HWA; Chapter 74, Article 4 NMSA 1978) and regulations under the Act. In 1989, NMED issued the Hazardous Waste Facility Permit (EPA ID Number NM0890010515-1) that established standards for the Laboratory to manage, store, and treat hazardous wastes on-site and to undertake the closure, post closure care, and cleanup as necessary, of permitted waste management units. On November 30, 2010, NMED renewed the Hazardous Waste Facility Permit (the Permit). In June 2020 the Permittees submitted a permit renewal application to the NMED for their review and approval. The Hazardous Waste Facility Permit Community Relations Plan (CRP) describes the scope of public involvement in the activities of the Permit. The CRP is specifically designed to facilitate the community outreach, engagement, and relations activities concerning the Laboratory’s Hazardous Waste Facility Permit and coordinates with but does not include public involvement for other Laboratory programs, initiatives, or environmental activities.

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

Summary Report for the NEPA Impact Analysis. Revision 1

The Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico, has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of defense-related transuranic (TRU) waste. Containment of TRU waste at the WIPP facility is derived from standards set forth in Title 40 of the Code of Federal Regulations (CFR), Part 191. The DOE assesses compliance with the containment standards according to the Certification Criteria in Title 40 CFR Part 194 by means of Performance Assessment (PA) calculations performed by Sandia National Laboratories (SNL). WIPP PA calculations estimate the probability of radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. The DOE Carlsbad Field Office (CBFO) has initiated a National Environmental Policy Act (NEPA) action for a proposal to excavate and use additional transuranic (TRU) waste disposal panels at the WIPP facility. This report documents an analysis undertaken as part of an effort to evaluate the potential environmental consequences of the proposed action. Although not explicitly required for a NEPA analysis, evaluations of a dose indicator to hypothetical members of the public after final facility closure are presented in this report. The analysis is carried out in two stages: first, Performance Assessment (PA) calculations quantify the potential releases to the accessible environment over a 10,000-year post-closure period. Second, dose was evaluated for three hypothetical exposure pathways using the conservative radionuclide concentrations assumed to be released to the accessible environment.

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

Applicability of Public Dose Limits to Individuals on the Nevada National Security Site

Historically all persons working on the Nevada National Security Site (NNSS) were considered subject to worker dose limits when calculations of potential dose from the air pathway were conducted to demonstrate compliance with Title 42 United States Code (USC) Chapter 85, “The Clean Air Act,” and Title 40 Code of Federal Regulations (CFR) Part 61, Subpart H, “National Emission Standards for Hazardous Air Pollutants (NESHAP).” Calculations were also conducted for the all-pathways potential dose to demonstrate compliance with U.S. Department of Energy (DOE) Order DOE O 458.1 Chg. 3, “Radiation Protection of the Public and the Environment.” The need to evaluate environmental air sampling locations and past regulator questions about the existence of persons potentially subject to public dose limits being on the NNSS drove this assessment (letter from EPA [U.S. Environmental Protection Agency] Region 9 to DOE/Nevada Operations Office, dated September 3, 1999). This assessment also supports determining those persons subject to public dose limits from the all pathways perspective. A person that is a member of the public is subject to public dose limits.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Analysis of Air-Purifying Respirator (APR) and Powered Air-Purifying Respirator (PAPR) Cartridge Performance Testing on a Hanford AP Tank Farm Exhauster Slipstream: Volume 2 Raw Analytical Data

As the Tank Operations Contractor for U.S. Department of Energy operations at the Hanford site in Washington State, Washington River Protection Solutions (WRPS) is responsible for managing highly radioactive wastes stored in tanks at Hanford. WRPS tests air-purifying respirator (APR) and powered air-purifying respirator (PAPR) chemical cartridges commonly used at Hanford Tank Farms to determine the period of time that the cartridges would provide adequate performance for APRs and PAPRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from any vapors exiting headspaces in the storage tanks. The Occupational Safety and Health Administration (OSHA) Standard promulgated in Title 29 of the Code of the Federal Regulations (CFR) 1910.134(d)(3)(iii)(b)(2) specifies that for protection against gases and vapors, employers shall implement a schedule for cartridges to ensure that change-outs occur before the end of service life. The change schedule can be based on objective information or data that ensures cartridge change-outs occur before the end of their service life.[2-5] The primary function of the WRPS Cartridge Test Program is to obtain objective data to determine service lives for the APR and PAPR cartridges used at Hanford Tank Farms. WRPS contracted with Pacific Northwest National Laboratory to analyze the test data and offer an independent analysis and any recommendations. Volume 1 of this report summarizes data analysis of APR and PAPR cartridge testing on vapors from the AP tank farm exhauster. Previous testing of APR cartridges was conducted on the AP exhauster in June 2016. However, an AP exhauster upgrade was completed in September 2016. Two different APR cartridges from SCOTT Safety (Monroe, North Carolina) were assessed for the new AP exhauster source, along with two different PAPR cartridges—one from MSA Safety Inc. (Pittsburgh, Pennsylvania) and another from 3M (Maplewood, Minnesota). Volume 2 provides an introduction to the raw data, including analytical laboratory analysis results that supported the analysis and conclusions documented in Volume 1.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Assessment of the 3420 Building Filtered Exhaust Stack Sampling Probe Location: Stack Verification Following Fan and Air Blender Additions

The Pacific Northwest National Laboratory (PNNL) 3420 Building, which is part of the Physical Sciences Facility (PSF), houses radiological capabilities that results in the requirement that emissions monitoring must be conducted for potential radionuclides in the exhaust air discharge of this building. The air monitoring system is required to conform to Title 40 of the Code of Federal Regulations part 61 (40 CFR 61) Subpart H, which in turns requires a sampling probe in the exhaust stream to conform to the criteria of American National Standards Institute / Health Physics Society (ANSI/HPS) N13.1-2011, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities. To support the air emissions permit for the 3420 Building on the PNNL campus, stack testing that used computational fluid dynamics (CFD) modeling as a surrogate stack and verification tests of velocity uniformity and flow angle on the retrofitted facility stack was performed. The ANSI/HPS N13.1-2011 criteria for the air monitoring probe location are that velocity uniformity, gaseous tracer uniformity, and particulate tracer uniformity must be less than or equal to 20%COV. Furthermore, no point in the sampling location may have a gaseous tracer concentration that varies from the mean concentration by more than 30%. Additionally, the flow angle at the sampling location must not be more than 20°. The CFD modeling of the stack, as reported by Recknagle et al. (2018) demonstrated that the stack meets the criteria at the probe location. The velocity uniformity and flow angle results from the 3420 stack verification tests, performed in October 2020, demonstrated that the CFD model results may be used to support the qualification of the stack sampling location. The measured velocity uniformity verification test result was 1.4%COV. This value is well within the uniformity criterion, which is that the velocity uniformity be =20%COV. Additionally, this value is well within the criterion that the actual stack measurement must be within 5% of the surrogate stack result of 2.1%COV when all four fans were operating. Additionally, the measured average flow angle at the 3420 stack monitor location was 15.5 degrees. Although this is higher than expected based on the CFD model, the result is =20 degrees, so the criterion is met. Based on these stack verification test results, the reconfigured 3420 Building filtered exhaust stack meets the qualification criteria given in the ANSI/HPS N13.1-2011 standard. Further changes to the system configuration or operating conditions that are outside the bounds described in this and the CFD report (Recknagle et al., 2018) may require additional tests and additional analysis to determine compliance with the standard.

3420 Building↗
DOE OSTI2020

Test Report for the Qualification Testing of the ORNL-SFC-W-1 Special Form Capsule

Three prototypes the ORNL-SFC-W-1 special form capsules (SFCs) of the same design were evaluated to determine if the requirements of Title 49, Code of Federal Regulations (CFR), Part 173.469, Tests for Special Form Class 7 (Radioactive) Materials, were met. The results of the special form tests are documented in this test report. This report describes the special form testing activities performed on the three ORNL-SFC-W-1 capsules. The test units were designated as TU-1, TU-2 and TU-3. TU-1 was subjected to leak testing, impact testing and percussion testing. TU-1 was leak tested using the leak rate test specified in 49 CFR 173.469(a)(4)(i). The impact test was conducted according to the requirements in 49 CFR 173.469 (b)(1). The percussion test that was carried out on TU-1 was conducted in accordance with 49 CFR 173.469 (b)(2). Both TU-2 and TU-3 were subjected to a leak rate test as specified in 173.469(a)(4)(i) and a heat test as specified in 49 CFR 173.469(b)(4). Each test unit was leak tested before and after these respective tests. The leak rate tests performed were helium back-pressure tests and bubble tests as specified in American National Standards Institute (ANSI) N14.5-2014.The measured leak rates were converted to standard condition leak rates as specified in American Society for Testing and Measurement (ASTM) E 493. The determined standardized leak rates obtained from the test and the calculations for all the test units met the requirements for special form certification. The testing was performed under the direction of the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP).

42 ENGINEERING↗
DOE OSTI2020

Summary Overview of Content Guidance For Early Site Permit Applications

The Early Site Permit (ESP) process offered under Title 10 Code of Federal Regulations (CFR) Part 52, Subpart A, was promulgated by the Nuclear Regulatory Commission (NRC) in 1989 to address industry concerns with the former licensing process under 10 CFR Part 50. Previously, the licensing process required large expenditures of time and money by applicants well before key site specific environmental, safety, and emergency planning issues could be resolved. As envisioned, the ESP process is meant to resolve these issues well in advance of when a decision is made to build a nuclear power facility and before substantial capital is invested in the construction of a new nuclear facility. The ESP application is required to provide design information pertaining to structures, systems, and components along with site-specific parameters (such as meteorology, demographics, and hydrology) and address site safety, environmental protection, and emergency planning. Although an ESP applicant does not need to specify a particular nuclear plant design, as in construction permit applications, it does need to provide sufficient surrogate design information (developed to bound the nuclear plant design(s) that are being considered by the applicant) so that the NRC can make a determination on the acceptability of the site and the environmental impacts, and determine whether designs bounded by the surrogate design information provided by the applicant can be qualified for the proposed site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗