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At least 865 records · Page 48

2020 NEPA Polychlorinated Biphenyl (PCB) Risk Assessment

The Waste Isolation Pilot Plant (WIPP) facility is a U.S. Department of Energy (DOE) operating repository 654 m below the surface in a thick salt formation in southeastern New Mexico. The DOE disposes transuranic (TRU) waste produced from atomic energy defense activities at the WIPP facility. A portion of the waste shipped to the WIPP facility contains TRU radionuclides co-mingled with polychlorinated biphenyls (PCBs), which fall under U.S. Environmental Protection Agency (EPA) regulations implementing the Toxic Substances Control Act (TSCA). This report documents the risks of PCBs co-mingled with TRU waste (hereafter designated as PCB/TRU waste) designated for disposal at the WIPP facility. This analysis is input to the National Environmental Policy Act (NEPA) assessment by the DOE Carlsbad Field Office (CBFO) for the proposed increase of the WIPP facility disposal area to include additional waste panels (but not to increase the legislated WIPP volume). This analysis is not a compliance calculation to support a certification renewal nor does it support a planned change request (PCR) or planned change notice (PCN) to be submitted to the EPA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Turbo FRMAC Implemetation of IAEA Radiological Assessment Methodologies for Nuclear and Radiological Emergencies.

This report documents the findings of an assessment of the Turbo FRMAC software's ability to implement International Atomic Energy Agency (IAEA) guidance for calculating Operational Intervention Levels (OIL) 1 & 2 for nuclear and radiological emergencies. The IAEA OIL and U.S. Federal Radiological Monitoring and Assessment Center (FRMAC) Derived Response Level methodology and implementation in respective tools were compared, as demonstrated through benchmarking activities for a nuclear power plant source term and potential radionuclides of concern for radiological dispersal devices. This comparison revealed some shortcomings in Turbo FRMACs ability to perform IAEA OIL calculations and resulted in recommended software modifications to be considered for future development.

61 RADIATION PROTECTION AND DOSIMETRY↗

Characterization of Tank 11H samples from tank closure cesium removal (TCCR) Batch 3 - intermediate and final samples

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. The TCCR unit processes dissolved salt from Tank 10H through a series of ion exchange columns containing crystalline silicotitanate (CST) and the effluent is then discharged to Tank 11H. Four interim samples pulled from Tank 11H during and just after the completion of processing of Batch 3 through the TCCR process have been analyzed for 137 Cs activity and density. The 137 Cs activity was found to decrease with each subsequent sample, which is consistent with the addition of decontaminated solution to Tank 11H. When compared to the expected composition from mixing the Tank 10H Batch 3 feed with the material already present in Tank 11H, the bulk chemical composition was as expected. A corrosion control sample collected from Tank 11H in June 2020 showed changes in the chemical composition and 137 Cs activity when compared to the composition measured at the end of Batch 2 processing. As there were no additions made to the tank during this period, these changes were attributed to leaching of the solids present in Tank 11H. Revision 1 of this report contains additional analyses of the fourth interim sample including inductively coupled plasma – mass spectrometry (ICP-MS) results and activities of other radionuclides. These results are consistent with the previously reported results, showing leaching from the Tank 11H heel. Notably, the 90 Sr activity was about 2.7x higher than calculated from a mixture of the original Tank 11H supernate and the Tank 10H supernate as treated by TCCR, suggesting additional leaching of Sr from the Tank 11H heel solids. In addition, Revision 1 includes analysis results for the Batch 3 Post-Production surface and variable depth samples collected in October of 2020. The Post-Production surface sample was similar in composition to the fourth interim surface sample, except for the increased Cs-137 and oxalate concentrations that were presumably associated with leaching of the Tank 11H heel solids.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nevada National Security Site 2019 Waste Management Monitoring Report Area 3 and Area 5 Radioactive Waste Management Sites

Environmental monitoring data are collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) within the Nevada National Security Site (NNSS). This report summarizes the 2019 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities. Some of these data (e.g., radiation exposure, air, and groundwater) are presented in other reports (Mission Support and Test Services, LLC 2019, 2020a, 2020b). Direct radiation monitoring data indicate exposure levels at the Area 3 and Area 5 RWMSs are within the range of background levels measured at the NNSS. Slightly elevated exposure levels outside the Area 3 RWMS are attributed to nearby historical aboveground nuclear weapons tests. Air monitoring data at the Area 3 and Area 5 RWMSs show that tritium concentrations in water vapor and americium and plutonium concentrations in air particles are below Derived Concentration Standards for these radionuclides. Groundwater monitoring data indicate the groundwater in the uppermost aquifer beneath the Area 5 RWMS is not impacted by RWMS operations. Results of groundwater analysis from wells around the Area 5 RWMS are below established investigation levels. Leachate samples collected from the leachate collection systems at the Area 5 mixed low-level waste disposal unit are below established contaminant regulatory limits. During 2019, precipitation at the Area 3 RWMS was 77 percent above average, and precipitation at the Area 5 RWMS was 69 percent above average. Water balance measurements indicate that evapotranspiration from the vegetated weighing lysimeter at the Area 5 RWMS dries the soil and prevents downward percolation of precipitation more effectively than evaporation as measured from the bare-soil weighing lysimeter. Vadose zone monitoring in the Area 3 and Area 5 RWMS soil covers shows no evidence of precipitation percolating through the covers to the waste. Moisture from precipitation did not percolate below 120 centimeters (3.9 feet [ft]) in the vegetated final cover on the U-3ax/bl disposal unit at the Area 3 RWMS during 2019. There was no drainage through 2.4 meters (8 ft) of soil as indicated from the Area 3 drainage lysimeters that received only natural precipitation. At the Area 3 RWMS, which received three times the natural precipitation, 57 percent of the applied precipitation and irrigation drained from the bare-soil drainage lysimeter. All 2019 monitoring data indicate that the Area 3 and Area 5 RWMSs are performing within expectations of the model and parameter assumptions for the facilities’ PAs.

2019↗

Operational and Mission Highlights A Monthly Summary of Top Achievements

In September 2020, the Laboratory’s Chemical Microscopy Facility tripled in size, expanding to 1,000 square feet. Located at TA-46, Building158, this facility houses a unique Laboratory resource. Operated by the Physical Chemistry and Applied Spectroscopy (C-PCS) group, this facility was established to renew the Laboratory’s rich history in radiobiology. The facility combines biological research capabilities with isotope production, as well as radioanalytical and chemical infrastructure. This facility provides space for Biosafety Level 2 (BSL-2) research, with alpha-emitting and other radionuclides of interest associated with biological systems. Equipped with a Class ll/A2 biological safety cabinet, a radiological fume hood and a host of other analytical equipment, this facility serves an effective resource to evaluate the efficacy of targeted alpha therapeutics, as well as the deleterious effects of radiological materials on living organisms. The facility receives funding through LDRD- 20180005DR, “Establishing a Radiotherapeutic Capability to Counter Biothreats,” and the BES Heavy Element Chemistry Program at the Laboratory.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

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↗

Decovalex-2019 (Executive Summary)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. The overall aim of DECOVALEX-2019 was to increase the understanding of various thermo-hydro-mechanical-chemical processes of importance for radionuclide release and transport from a repository to the biosphere and how they can be described and modelled using mathematical models. The scientific and technical objectives are: to increase the basic understanding of T-H-M-C coupled processes in fractured rocks (crystalline, sedimentary, argillaceous) and buffer materials; to investigate the predictive capabilities of different codes to field experiments and to perform verification of codes; to exchange experimental data, and improve the understanding of the constitutive behavior of crystalline and argillaceous rock masses and buffer materials; and to perform THMC calculations in a performance/safety assessment context.

58 GEOSCIENCES↗

Economic Model for Estimation of GDP Losses in the MACCS Offsite Consequence Analysis Code

The MACCS (MELCOR Accident Consequence Code System) code is the U.S. Nuclear Regulatory Commission (NRC) tool used to perform probabilistic health and economic consequence assessments for atmospheric releases of radionuclides. It is also used by international organizations, both reactor owners and regulators. It is intended and most commonly used for hypothetical accidents that could potentially occur in the future rather than to evaluate past accidents or to provide emergency response during an ongoing accident. It is designed to support probabilistic risk and consequence analyses and is used by the NRC, U.S. nuclear licensees, the Department of Energy, and international vendors, licensees, and regulators. This report describes the modeling framework, implementation, verification, and benchmarking of a GDP-based model for economic losses that has recently been developed as an alternative to the original cost-based economic loss model in MACCS. The GDP-based model has its roots in a code developed by Sandia National Laboratories for the Department of Homeland Security to estimate short-term losses from natural and manmade accidents, called the Regional Economic Accounting analysis tool (REAcct). This model was adapted and modified for MACCS and is now called the Regional Disruption Economic Impact Model (RDEIM). It is based on input-output theory, which is widely used in economic modeling. It accounts for direct losses to a disrupted region affected by an accident, indirect losses to the national economy due to disruption of the supply chain, and induced losses from reduced spending by displaced workers. RDEIM differs from REAcct in its treatment and estimation of indirect loss multipliers, elimination of double counting associated with inter-industry trade in the affected area, and that it is designed to be used to estimate impacts for extended periods that can occur from a major nuclear reactor accident, such as the one that occurred at the Fukushima Daiichi site in Japan. Most input-output models do not account for economic adaptation and recovery, and in this regard RDEIM differs from its parent, REAcct, because it allows for a user-definable national recovery period. Implementation of a recovery period was one of several recommendations made by an independent peer review panel to ensure that RDEIM is state-of-practice. For this and several other reasons, RDEIM differs from REAcct. Both the original and the RDEIM economic loss models account for costs from evacuation and relocation, decontamination, depreciation, and condemnation. Where the original model accounts for an expected rate of return, based on the value of property, that is lost during interdiction, the RDEIM model instead accounts for losses of GDP based on the industrial sectors located within a county. The original model includes costs for disposal of crops and milk that the RDEIM model currently does not, but these costs tend to contribute insignificantly to the overall losses. This document discusses three verification exercises to demonstrate that the RDEIM model is implemented correctly in MACCS. It also describes a benchmark study at five nuclear power plants chosen to represent the spectrum of U.S. commercial sites. The benchmarks provide perspective on the expected differences between the RDEIM and the original cost-based economic loss models. The RDEIM model is shown to consistently predict larger losses than the original model, probably in part because it accounts for national losses by including indirect and induced losses; whereas, the original model only accounts for regional losses. Nonetheless, the RDEIM model predicts losses that are remarkably consistent with the original cost-based model, differing by 16% at most for the five sites combined with three source terms considered in this benchmark.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Summary Report for the Analysis of the Sludge Batch 7b (Macrobatch 9) DWPF Pour Stream Glass Sample for Canister S04023

In order to comply with the Defense Waste Processing Facility (DWPF) Waste Form Compliance Plan for Sludge Batch 7b, Savannah River National Laboratory (SRNL) personnel characterized the Defense Waste Processing Facility (DWPF) pour stream (PS) glass sample collected while filling canister S04023. This report summarizes the results of the compositional analysis for reportable oxides and radionuclides and the normalized Product Consistency Test (PCT) results. The PCT responses indicate that the DWPF produced glass that is significantly more durable than the Environmental Assessment glass. Results and further details are documented in SRNL-STI-2013-00462, Revision 1 “Analysis of the DWPF Sludge Batch 7b (Macrobatch 9) Pour Stream Glass Sample.”

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Acceptable Knowledge Technologist Procedure

Process acceptable knowledge (AK) is a waste stream characterization method that can be used to meet all or part of the waste analysis requirements appropriate for the waste. The method may include documented knowledge of process (KOP), additional characterization data, and/or facility records of analysis. LANL will use process AK to ensure chemical compatibility; assign matrix parameter categories and Environmental Protection Agency (EPA) hazardous waste numbers (HWNs) commonly referred to as EPA codes to waste streams; and determine the waste material parameters and radionuclides present in waste streams.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Vitrification of Hanford Tank 241-AP-107 with Recycled Condensate

During the vitrification of nuclear waste at the Hanford Waste Treatment and Immobilization Plant (WTP) – the primary mission of the U.S. Department of Energy Office of River Protection – the offgas condensate generated from the waste-to-glass conversion is currently planned to be concentrated by evaporation in the Effluent Management Facility (EMF). This concentrated condensate can then be recycled back to the incoming waste and vitrified. To test the recycle process, a test apparatus was designed to mimic the EMF evaporator and used to concentrate a volume of condensate that had been previously produced during the vitrification of Hanford tank 241-AP-107 (referred to herein as AP-107) waste in a continuous laboratory-scale melter (CLSM). The test apparatus successfully concentrated the AP-107 condensate by a factor of 10 while retaining over 90 % of the technetium-99 ( 99 Tc), Cs, and I inventory. A second portion of AP-107 waste was retrieved by Washington River Protection Solutions, LLC, given to Pacific Northwest National Laboratory, and combined with the AP-107 condensate concentrate after undergoing solids filtration and cesium removal by ion exchange. This combination served to approximate the recycling action to be performed at the WTP. After the addition of glass-forming chemicals (GFCs), the combined AP-107 waste and AP-107 condensate concentrate were processed in the CLSM to produce a glass, called AP-107-1R, that was designed to satisfy the WTP baseline requirements (Kim et al. 2012). During the 8.87 hours of processing, 7.27 kg of AP-107-1R glass were produced for an average glass production rate of 1739 kg m 2 d -1 . Compared to the previous run in the CLSM without recycled condensate, the run with the recycle had a greater average glass production rate, but the rate was within the potential range of variability when processing melter feeds with similar composition in the CLSM. The glass produced from the AP-107 recycle run in the CLSM was within 10 % of the target AP-107-1R glass composition with respect to the primary glass components. Analysis of the minor component impurities revealed that their content in the glass product had approached their nominal target after 2 turnovers of the glass inventory in the CLSM while the activity of the minor radionuclides was retained in the glass product. The 99 Tc and total cesium content in the combined AP-107 waste and recycled condensate were maintained at concentrations expected to be experienced at the WTP. During processing in the CLSM, at discrete sampling time periods, the target 99 Tc/Cs mass ratio in the glass formulation varied from 0.9 to 62.9. Across this range, the Cs retention in the glass ranged from 53 to 60 %, while the retention from the entire runtime totaled 68 %, values which align with Cs retention in other scaled melter systems while processing LAW melter feeds at 99 Tc/Cs mass ratios ranging from 1 to 100. The 99 Tc retention in the glass ranged from 22 to 32 %, primarily due to the cold-cap coverage on the glass melt surface, the area covered by reacting melter feed, varying from ~80 % to ~95 % during processing, demonstrating greater volatility of 99 Tc from the glass while more surface was exposed, as expected based on previous 99 Tc volatility studies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Physics Flash Winter 2021 [Newsletter]

The Winter 2021 quarterly newsletter highlights articles titled "Putting down new markers for Los Alamos neutrino research", Direct measurements of nuclear reactions on radionuclides", Novel thermal neutron detector safeguards aircraft, HPC semiconductor electronics", along with the usual standard sections.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

C-NR Countroom Overview [Slides]

Perform qualitative and quantitative assay of gamma, beta, and alpha-emitting radionuclides in a variety of matrices and over a wide range of activity levels.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Assessing subtle variations in the actinyl oxo reactivity through characterization of neptunyl complexes. Final Report

Formation of the actinyl cation, [An(V,VI)O 2 ] n+ (n = 1,2), imparts unique chemical properties to the lighter actinide elements (U, Np, and Pu) that has yet to be fully understood. The axial O atoms (oxo) of the [An(VI)O 2 ] 2+ are considered weak Lewis bases that exhibit only meager interact with other metal cations and regarded as poor hydrogen acceptors for nearby water molecules. Oxo atoms of [An(V)O 2 ] + should engage in stronger bonding interactions due to a higher degree of Lewis basicity, but the actual interactions have yet to be fully explored. Intermolecular interactions can impact radionuclide separations, corrosion of spent fuel, and mobility of actinides in environmental systems; there is, therefore, a critical need to understand how the subtle difference in electronic properties of the An(V) and An(VI) metal center influence actinyl chemistry. The objective of this proposal was to determine the chemical components that influence the intermolecular interactions occurring between neptunyl ([Np(V)O 2 ] + and [Np(VI)O 2 ] 2+ ) cations and neighboring species. My overall hypothesis was that the intermolecular attraction that occurs between the neptunyl oxo atoms and neighboring species (H atoms, low-valent cations, actinyl cations) is primarily controlled by the electronic properties of the actinyl cation, but can be further influenced by the electron donating properties of the equatorial ligands. The hypothesis was tested through the following research objectives: (1) Determine the fundamental differences in H-bonding between Np(V)O 2+ and Np(VI)O 2 2+ and neighboring donor molecules; (2) Identify the major differences in bonding between low-valence metal cations and neptunyl oxo groups, and (3) Assess the impact of the equatorial ligand on the formation of cation-cation interactions within molecular species. Major outcomes of the research include: (1) More precisely identifying oxo interactions (actinyl-cation; actinyl-actinyl, actinyl-hydrogen) and the factors that impact the vibrational modes for actinyl cation; (2) Identification of H-bonding modes for the [An(V,VI)O 2 ] n+ (n = 1,2), with direct H-bonding interactions to the oxo group of the Np(V)O 2+ cation more prevalent than in hexavalent species and recognition of the importance of H-bonding networks, that can impart bond asymmetry and activation of vibrational modes with the actinyl moiety; (3) Delineation of the importance of charge density for engaging the cation-oxo interaction for both pentavalent and hexavalent species; (4) Determining the importance of sterics and ligand binding in formation of specific actinyl-actinyl interactions and the resulting vibrational modes associated with this bonding motify and (5) Understanding how the functionalization of the ligand can impact redox stability of neptunyl cations. Deliverables from this work include 11 peer-reviewed manuscripts and 13 (including six by graduate students) oral/poster presentations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Characterization results for the 2020 tank farm 2H and 3H evaporator overhead samples

The 2020 annual 2H and 3H Evaporator overhead samples were received at Savannah River National Laboratory (SRNL) on December 22, 2020. The delivered samples were analyzed for concentration of cesium-137 ( 137 Cs), strontium-90 (90 Sr) and iodine-129 ( 129 I) to validate compliance with the Effluent Treatment Project Waste Acceptance Criteria (ETP WAC, Rev. 7). The concentration of all three radionuclides, 137 Cs, 90 Sr, and 129 I, in the 2H Evaporator overhead sample were found to be less than the corresponding ETP WAC limits. Radiochemical analysis of the 3H Evaporator overhead sample indicated a 137 Cs content about an order of magnitude higher than the acceptable threshold value, and 90 Sr and 129 I concentrations within the acceptance limits.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

U.S. Efforts in Support of Examinations at Fukushima Daiichi- November 2020 Meeting Notes with Updated Information Requests

Much is still not known about the end-state of core materials in each unit that was operating on March 11, 2011 at the Fukushima Daiichi Nuclear Power Station (Daiichi). Information obtained from Daiichi is required to inform Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. commercial nuclear power plants. This document summarizes results from the Fiscal Year 2021 (FY2021) U.S. effort to review Daiichi information and extract insights to enhance the safety of existing and future nuclear power plant designs. This U.S. effort, which was initiated in 2014 by the Department of Energy Office of Nuclear Energy (DOE-NE), is completed by a group of experts in reactor safety and plant operations that identify examination needs and evaluate recent Daiichi examination data to address these needs. Since its inception, annual reports were issued that document significant safety insights being obtained in areas of special emphasis: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties related to severe accident modeling progression, these insights have and continue to be used to update guidance for severe accident prevention, mitigation, and emergency planning. Reduced uncertainties in modeling the events at Daiichi improve the realism of reactor safety evaluations that inform future D&D activities. A key aspect of prior U.S. efforts, the updated list of information requests, is included in this FY2021 report to ensure that they are transmitted to organizations within Japan. This report also continues to emphasize how information obtained from the affected reactors at Daiichi has been and will continue to be used to update severe accident management strategies and reduce uncertainties in systems analysis code models. In addition, recommendations are included that would expand the use of this information to provide insights regarding maintenance, radiation protection, design, and siting activities for existing and new reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Guidelines for Broadband Seismometer Testing

The Comprehensive Nuclear-Test-Ban Treaty mandates the four verification technologies to be used by the International Monitoring System (IMS) to monitor compliance to the Treaty. These technologies are seismic, hydroacoustic, infrasound, and radionuclide. Operational manuals for each of these technologies specify the requirements that equipment installed at IMS stations must meet for data from said stations to be accepted by the International Data Center. Following a model in which a revised set of infrasound sensor specifications and accompanying test procedures were developed in an international collaboration overseen by the Comprehensive Nuclear-Test-Ban Treaty Organization, the same process was recently followed for seismometers potentially of use by the IMS. This document, a product of that collaboration, defines key concepts and recommends test procedures to be followed to characterize broadband seismometers in a consistent and standardized way. Results from these test procedures can be used to verify that a seismometer meets the manufacturers stated specifications and IMS requirements. ACKNOWLEDGEMENTS The writers of this document would like to acknowledge Charles R. Hutt, John R. Evans, Fred Followill, Robert L. Nigbor, and Erhard Wielandt, the authors of the 2009 Guidelines for Standardized Testing of Broadband Seismometers and Accelerometers funded by the U.S. Department of Interior, on which this document is based. Although this document is limited in scope to seismometers and much of the document has been updated, the structure remains essentially the same.

58 GEOSCIENCES↗

CY2019 Annual Closure Monitoring Report for Corrective Action Unit 98, Frenchman Flat, Underground Test Area, Nevada National Security Site, Nevada (January 2019–December 2019), Revision 1

Three types of monitoring are performed for CAU 98: water quality, water level, and institutional control. These are monitored to determine whether the URs remain protective of human health and the environment, and to ensure that the regulatory boundary objectives are being met. Monitoring data will be used in the future, once multiple years of data are available, to evaluate consistency with the groundwater flow and contaminant transport models because the contaminant boundaries calculated with the models are the primary basis of the UR boundaries. Six wells were sampled for water-quality monitoring in 2019. Contaminants of concern were detected only in the two source/plume wells already known to contain contamination as a result of a radionuclide migration experiment. Tritium concentrations in both of these wells, RNM-2S and UE-5n, remain above the Safe Drinking Water Act maximum contaminant level of 20,000 picocuries per liter but declined in 2019 as compared to measurements in 2018. All other contaminants of concern are below the minimum detection level plus analytical error. The water-level monitoring network includes 16 wells. Depth to water measured in 2019 is generally consistent with recent measurements for all wells. Many wells continue to exhibit a long-term downward trend in water level, though changes from 2018 to 2019 are minimal. The sharp 2016 decline in water level in Well ER-5-3-2 remains unexplained, with the lower level persisting through 2019. Rising water-level trends continue to be observed in Well ER-5-3 deep piezometer and former water supply Well WW-5A. Water supply Well WW-5B experienced a rise in water level as a result of an absence of pumping in the first part of the year (due to a mechanical problem), whereas water levels declined in WW-4 and WW-4A in response to greater pumping in 2019. Institutional control monitoring confirmed the URs are recorded in U.S. Department of Energy and U.S. Air Force land management systems, and that no activities within Frenchman Flat basin are occurring that could potentially affect the contaminant boundaries. Survey of groundwater resources in basins surrounding Frenchman Flat similarly identify no current or pending development that would indicate the need to increase monitoring activities or would otherwise cause concern for the closure decision. The URs continue to prevent exposure of the public, workers, and the environment to contaminants of concern by preventing use of potentially contaminated groundwater.

54 ENVIRONMENTAL SCIENCES↗