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DOE OSTI2023

Ultra-low radioactivity flexible printed cables

Abstract Flexible printed cables and circuitry based on copper-polyimide materials are widely used in experiments looking for rare events due to their unique electrical and mechanical characteristics. However, past studies have found copper-polyimide flexible cables to contain 400-4700 pg 238 U/g, 16-3700 pg 232 Th/g, and 170-2100 ng nat K/g, which can be a significant source of radioactive background for many current and next-generation ultralow background detectors. This study presents a comprehensive investigation into the fabrication process of copper-polyimide flexible cables and the development of custom low radioactivity cables for use in rare-event physics applications. A methodical step-by-step approach was developed and informed by ultrasensitive assay to determine the radiopurity in the starting materials and identify the contaminating production steps in the cable fabrication process. Radiopure material alternatives were identified, and cleaner production processes and treatments were developed to significantly reduce the imparted contamination. Through the newly developed radiopure fabrication process, fully-functioning cables were produced with radiocontaminant concentrations of 20-31 pg 238 U/g, 12-13 pg 232 Th/g, and 40-550 ng nat K/g, which is significantly cleaner than cables from previous work and sufficiently radiopure for current and next-generation detectors. This approach, employing witness samples to investigate each step of the fabrication process, can hopefully serve as a template for investigating radiocontaminants in other material production processes.

36 MATERIALS SCIENCE↗

DOE OSTI2013

Solar Powered Radioactive Air Monitoring Stations

Environmental monitoring of ambient air for radioactive material is required as stipulated in the PNNL Site radioactive air license. Sampling ambient air at identified preferred locations could not be initially accomplished because utilities were not readily available. Therefore, solar powered environmental monitoring systems were considered as a possible option. PNNL purchased two 24-V DC solar powered environmental monitoring systems which consisted of solar panels, battery banks, and sampling units. During an approximate four month performance evaluation period, the solar stations operated satisfactorily at an on-site test location. They were subsequently relocated to their preferred locations in June 2012 where they continue to function adequately under the conditions found in Richland, Washington.

air sampling↗

DOE OSTI2015

Data Collection Handbook to Support Modeling Impacts of Radioactive Material in Soil and Building Structures

This handbook is an update of the 1993 version of the Data Collection Handbook and the Radionuclide Transfer Factors Report to support modeling the impact of radioactive material in soil. Many new parameters have been added to the RESRAD Family of Codes, and new measurement methodologies are available. A detailed review of available parameter databases was conducted in preparation of this new handbook. This handbook is a companion document to the user manuals when using the RESRAD (onsite) and RESRAD-OFFSITE code. It can also be used for RESRAD-BUILD code because some of the building-related parameters are included in this handbook. The RESRAD (onsite) has been developed for implementing U.S. Department of Energy Residual Radioactive Material Guidelines. Hydrogeological, meteorological, geochemical, geometrical (size, area, depth), crops and livestock, human intake, source characteristic, and building characteristic parameters are used in the RESRAD (onsite) code. The RESRAD-OFFSITE code is an extension of the RESRAD (onsite) code and can also model the transport of radionuclides to locations outside the footprint of the primary contamination. This handbook discusses parameter definitions, typical ranges, variations, and measurement methodologies. It also provides references for sources of additional information. Although this handbook was developed primarily to support the application of RESRAD Family of Codes, the discussions and values are valid for use of other pathway analysis models and codes.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2020

The ENDF/B-VIII.0 Compendium Volume IV: Radioactive Decay Rank-Order Analyses

Version 8 of the Evaluated Nuclear Data File (ENDF/B-VIII.0) contains a wealth of nuclear data representing the most recent collection of evaluated neutron cross section measurements, isotopic radioactive decay, and much more. This Compendium has been prepared as a easily-navigable reference of the contents of the ENDF/B-VIII.0 library, while making no statement regarding the accuracy of those contents. The Compendium is split into four volumes. In Volume IV, radioactive decay properties are presented in tabular form, sorted by isotope, energy, element-then- energy, and probability of emission.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Collective Analysis of Alpha Particle Losses Due to Self-Absorption by Mass Loading on Radioactive Particulate Glass Fiber Filters

In this study, we derived a relationship between filter mass loading and the percent loss during analysis using the mass loading data collected from six previous studies of self-absorption. Components of mass loading include particulate dust, radioactive particulates, and filter material. In a research report published in 1984, Higby calculated a minimum burial depth for an alpha particle to be lost due to absorption (100% loss) of about 3.7 mg/cm 2 based on calculations for the range of 239-Pu alpha particles in glass fiber filters. From there, Higby concluded that a correction factor of 0.85 assumes approximately 15% losses in the count rate of both alpha and beta particles. In 2000, Luetzelschwab et al. recommended assuming a 40% loss at a loading of 3.3 mg/cm 2 and a 28% loss for a loading of 2.3 mg/cm 2 which included the frontal face mass of the filter. More recently, the 100% losses due to absorption were reported to be in the 10 mg/cm 2 range. Presented here is a trinomial relationship method of relating percent loss due to self-absorption to filter mass loading, based on data reported by Higby, Luetzelschwab et al., Huang et al., Barnett et al., Smith et al., and Hogue et al. Under normal operating conditions at the stacks monitored by Effluent Management, the mass loading of sample filters averages 0.09 ± 0.12 (2s) mg/cm 2 (excluding negative values and outliers) and ranges from 0 mg/cm 2 to 0.24 mg/cm 2 . Based on current mass loading results for Effluent Management stack sample filters, the forced-zero trinomial relationship method estimated self-absorption losses of less than 5%. Because American National Standards Institute/Health Physics Society N13.1-2011 guidelines indicate a correction factor should be used when the penetration of radioactive material into the collection media or self-absorption of radiation by the material collected would reduce the count rate by more than 5%, it is possible continued application of a correction factor to the Effluent Management stack samples is no longer necessary. Nevertheless, continuing to assign a correction factor at the 5% threshold (i.e., 0.95) would be a conservative approach.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Meta-Study of Particulate Detection Losses on Radioactive Air Sample Filters

Several mathematical relationships between air sample filter mass loading and the correlated analytical self-absorption factor were developed using data from other published research in this meta-study. Gross-alpha and -beta applications are addressed for this research. As filter media becomes loaded with particulate matter, there is potential for measurement losses due to self-absorption by mass loading. Components contributing to absorption include particulate dust, radioactive particulates, and filter material. Standards indicate a correction factor should be used when the penetration of radioactive material into the collection media or self-absorption of radiation by the material collected would reduce the detection rate by more than 5%. Previously, losses due to self-absorption have been reported up to 100% over a range up to ~10 mg∙cm -2 mass loading. These absorption losses then can be used to determine a correction factor for sample results. For low mass loadings (e.g., ≤0.1 mg∙cm -2 ) corrections factors in the 0.85 - 1 range have been recommended and used, while at higher mass loadings nearer to 10 mg∙cm -2 correction factors closer to 0 (representing near 100% losses) are used. Based on data from published studies, the different methods for relating percent loss due to self-absorption to mass loading include linear, exponential, quadratic, and trinomial derived functions. Where applicable, both forced zero and non-forced zero results were evaluated. From the derived functions evaluated, the trinomial function provided the best fit. Once the sample filter mass loading is known, the trinomial function can be applied to estimate losses and the corresponding self-absorption factor. When applied to routine operating conditions for radiological facility stacks monitored at the Pacific Northwest National Laboratory for an average sample filter mass loading of 0.09 ± 0.12 (2σ) mg∙cm -2 (excluding negative values and outliers) and a range from 0 - 0.24 mg∙cm -2 , the estimated trinomial function nominal self-absorption losses are less than 5% at 0.09 mg∙cm -2 and less than 10% at 0.24 mg∙cm -2 . The trinomial function is one method that may be used to adjust the activity results of an air sample when the sample-specific mass loading is determined. The application of no correction factor when the ANSI/HPS N13.1-2021 guidance of a 5% threshold for loss is not reached with typical stack sample mass loadings may be reasonable in high-efficiency particulate air filtered systems. For simplicity, it would be conservative in assigning the self-absorption correction factor at the 5% threshold (i.e., 0.95) for general uses but in cases of heavy mass loading to calculate the factor.

air sampling↗

NASA NTRS1967

Investigations into attachment properties of radioactive aerosols

Attachment of radioactive small ions and of neutral radioactive substances to aerosols, based on solution of diffusion equation

DIFFUSION↗

NASA NTRS2016

Evaluation of Observed and Modelled Aerosol Lifetimes Using Radioactive Tracers of Opportunity and an Ensemble of 19 Global Models

Aerosols have important impacts on air quality and climate, but the processes affecting their removal from the atmosphere are not fully understood and are poorly constrained by observations. This makes modelled aerosol lifetimes uncertain. In this study, we make use of an observational constraint on aerosol lifetimes provided by radionuclide measurements and investigate the causes of differences within a set of global models. During the Fukushima Dai-Ichi nuclear power plant accident of March 2011, the radioactive isotopes cesium-137 (Cs-137) and xenon-133 (Xe-133) were released in large quantities. Cesium attached to particles in the ambient air, approximately according to their available aerosol surface area. Cs-137 size distribution measurements taken close to the power plant suggested that accumulation mode (AM) sulfate aerosols were the main carriers of cesium. Hence, Cs-137 can be used as a proxy tracer for the AM sulfate aerosol's fate in the atmosphere. In contrast, the noble gas Xe-133 behaves almost like a passive transport tracer. Global surface measurements of the two radioactive isotopes taken over several months after the release allow the derivation of a lifetime of the carrier aerosol. We compare this to the lifetimes simulated by 19 different atmospheric transport models initialized with identical emissions of Cs-137that were assigned to an aerosol tracer with each model's default properties of AM sulfate, and Xe-133 emissions that were assigned to a passive tracer. We investigate to what extent the modelled sulfate tracer can reproduce the measurements, especially with respect to the observed loss of aerosol mass with time. Modelled Cs-137and Xe-133 concentrations sampled at the same location and times as station measurements allow a direct comparison between measured and modelled aerosol lifetime. The e-folding lifetime e, calculated from station measurement data taken between 2 and 9 weeks after the start of the emissions, is 14.3 days (95% confidence interval 13.1-15.7 days). The equivalent modelled e lifetimes have a large spread, varying between 4.8 and 26.7 days with a model median of 9.42.3 days, indicating too fast a removal in most models. Because sufficient measurement data were only available from about 2 weeks after the release, the estimated lifetimes apply to aerosols that have undergone long-range transport, i.e. not for freshly emitted aerosol. However, modelled instantaneous lifetimes show that the initial removal in the first 2 weeks was quicker (lifetimes between 1 and 5 days) due to the emissions occurring at low altitudes and co-location of the fresh plume with strong precipitation. Deviations between measured and modelled aerosol lifetimes are largest for the northernmost stations and at later time periods, suggesting that models do not transport enough of the aerosol towards the Arctic. The models underestimate passive tracer (Xe-133) concentrations in the Arctic as well but to a smaller extent than for the aerosol (Cs-137) tracer. This indicates that in addition to too fast an aerosol removal in the models, errors in simulated atmospheric transport towards the Arctic in most models also contribute to the underestimation of the Arctic aerosol concentrations.

Radioactive isotopes↗

DOE OSTI2018

Radioactivity Calculator for Research and Test Reactor Neutron-Activated Materials

The Idaho National Laboratory Nuclear Science User Facilities program is proposing to launch a new online expert system that will assist users in planning experiments and writing proposals. The Combined Materials Experimentation Toolkit (CoMET) will include two nuclear calculators. The first of these, the Neutron Damage Calculator, will assist users in selecting reactor positions in order to obtain desired material damage in Damage Per Atom (DPA) and provide the required irradiation time to reach the damage levels. The second, the Radioactivity Calculator, will estimate the post-irradiation radioactivity levels of sample materials, which will inform technical staff of examination facility requirements and any potential scheduling delays. Both of these calculators were designed to simplify the information needed from users in order to properly plan and execute awarded proposals. This paper presents the primary purpose and methodologies implemented by these two calculators.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Digital Autoradiography as a novel complementary technique for the investigation of radioactive contamination in nuclear facilities under dismantlement

The detection (location and quantification) of nuclear facilities to be dismantled possible contamination with low-range particles emitters ({sup 3}H, other low-energy β emitters, a emitters) remains a tedious and expensive task. Indeed, usual remote counters show a too low sensitivity to these non-penetrating radiations, while conventional wipe tests are irrelevant for fixed radioactivity evaluation. The only method to accurately measure activity levels consists in sampling and running advanced laboratory analyses (spectroscopy, liquid scintillation counting, pyrolysis...). Such measurements generally induce sample preparation, waste production (destructive analyses, solvents), nuclear material transportation, long durations, and significant labor mobilization. Therefore, the search for the limitation of their number and cost easily conflicts with the necessity to perform a dense screening for sampling (to maximize the representativeness of the samples), in installations of thousands of square meters (floors, wells, ceilings), plus furniture, pipes, and other wastes. To overcome this contradiction, Digital Autoradiography (D. A.) was re-routed from bio molecular research to radiological mapping of nuclear installations under dismantling and to waste and sample analysis. After in-situ exposure to the possibly-contaminated areas to investigate, commercial reusable radiosensitive phosphor screens (of a few 100 cm{sup 2}) were scanned in the proper laboratory device and sharp quantitative images of the radioactivity could be obtained. The implementation of geostatistical tools in the data processing software enabled the exhaustive characterization of concrete floors at a rate of 2 weeks / 100 m{sup 2}, at lowest costs. Various samples such as drilled cores, or tank and wood pieces, were also successfully evaluated with this method, for decisive results. Thanks to the accurate location of potential contamination spots, this approach ensures relevant and representative sampling for further laboratory analyses and should be inserted in the range of common tools used in dismantling. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2018

Radioactivity and neutron damage calculators for research and test reactor neutron-activated materials - 25652

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Determination of Erosion/Corrosion Rates in Hanford Tank Farms Radioactive Waste Transfer System Pipelines

The twenty-eight double-shell underground radioactive waste storage tanks at the U. S. Department of Energy’s Hanford Site near Richland, WA are interconnected by the Waste Transfer System network of buried steel encased pipelines and pipe jumpers in below-grade pits. The pipeline material is stainless steel or carbon steel in 51 mm to 152 mm (2 in. to 6 in.) sizes. The pipelines carry slurries ranging up to 20 volume percent solids and supernatants with less than one volume percent solids at velocities necessary to prevent settling. The pipelines, installed between 1976 and 2011, were originally intended to last until the 2028 completion of the double-shell tank storage mission. The mission has been subsequently extended. In 2010 the Tank Operating Contractor began a systematic evaluation of the Waste Transfer System pipeline conditions applying guidelines from API 579-1/ASME FFS-1 (2007), Fitness-For-Service. Between 2010 and 2014 Fitness-for-Service examinations of the Waste Transfer System pipeline materials, sizes, and components were completed. In parallel, waste throughput histories were prepared allowing side-by-side pipeline wall thinning rate comparisons between carbon and stainless steel, slurries and supernatants and throughput volumes. The work showed that for transfer volumes up to 6.1E+05 m 3 (161 million gallons), the highest throughput of any pipeline segment examined, there has been no detectable wall thinning in either stainless or carbon steel pipeline material regardless of waste fluid characteristics or throughput. The paper describes the field and laboratory evaluation methods used for the Fitness-for-Service examinations, the results of the examinations, and the data reduction methodologies used to support Hanford Waste Transfer System pipeline wall thinning conclusions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Environmental Effects of Cavitation Technology for Radioactive Waste Management

Results from studies of the removal from storage and solidification of low- and intermediate-level liquid radioactive waste in inorganic binders (cementation) are presented. Cavitation technology for recycling radiochemical wastes from nuclear power plant decommissioning increases the extraction throughput and strength of the cement compound.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Effect of supplementary cementitious materials on the degradation of cement-based barriers in radioactive waste repository: A case study in Korea

This study focuses on investigating the chemical degradation characteristics of cementitious barriers used in low- and intermediate-level radioactive waste repository by reactive transport modeling. The impact of the blending with supplementary cementitious materials (SCMs) in the barriers on the chemical degradation was evaluated to find the optimum barrier design. A number of different barrier designs were examined by replacing ordinary Portland cement (OPC) by SCMs (i.e., fly ash, silica fume, and blast-furnace slag). The simulation results showed that silica fume blended barrier has better durability against chemical degradation by rainwater compared to fly ash or blast-furnace slag blended barriers. In addition, the chemical durability of silica fume-based barrier increased with increasing replacement level up to about 20 %. It seems that the amount of formed calcium silicate hydrate (CSH) in the initial cement-based barrier highly affects the overall chemical durability. The newly developed reactive transport model demonstrated its capability for understanding the barrier performance and investigating the optimal design of the barrier system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Development of characterization protocol for mixed liquid radioactive waste classification

Mixed liquid organic waste generated from health-care and research activities containing tritium, carbon-14, and other radionuclides posed specific challenges in its management. Often, these wastes become legacy waste in many nuclear facilities and being considered as ‘problematic’ waste. One of the most important recommendations made by IAEA is to perform multistage processes aiming at declassification of the waste. At this moment, approximately 3000 bottles of mixed liquid waste, with estimated volume of 6000 litres are currently stored at the National Radioactive Waste Management Centre, Malaysia and some have been stored for more than 25 years. The aim of this study is to develop a characterization protocol towards reclassification of these wastes. The characterization protocol entails waste identification, waste screening and segregation, and analytical radionuclides profiling using various analytical procedures including gross alpha/ gross beta, gamma spectrometry, and LSC method. The results obtained from the characterization protocol are used to establish criteria for speedy classification of the waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Vacuum arc with a distributed cathode spot as a plasma source for plasma separation of spent nuclear fuel and radioactive waste

Results from experimental studies of a vacuum arc with a distributed cathode spot on the heated cathode are presented. Such an arc can be used as a plasma source for plasma separation of spent nuclear fuel and radioactive waste. The experiments were performed with a gadolinium cathode, the properties of which are similar to those of an uranium arc cathode. The heat flux from the plasma to the cathode (and its volt equivalent) at discharge voltages of 4-15 V and discharge currents of 44-81 A, the radial distribution of the emission intensity of gadolinium atoms and singly charged ions in the arc channel at a voltage of 4.3 V, and the plasma electron temperature behind the anode were measured. The average charge of plasma ions at arc voltages of 3.5-8 V and a discharge current of 52 A and the average rate of gadolinium evaporation in the discharge were also determined.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Nevada National Security Site 2015 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). These data include direct radiation exposure, as well as radiation from the air, groundwater, meteorology, and vadose zone. This report summarizes the 2015 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 developed by National Security Technologies, LLC. Direct radiation monitoring data indicate exposure levels at the 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 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 were all below established investigation levels. Leachate samples collected from the leachate collection system at the mixed low-level waste cell were below established contaminant regulatory limits. During 2015, precipitation at the Area 3 RWMS was 0.9% above average, and precipitation at the Area 5 RWMS was 25% above average. Water balance measurements indicate that evapotranspiration from the vegetated weighing lysimeter dries the soil and prevents downward percolation of precipitation more effectively than evaporation as measured from the bare-soil weighing lysimeter. The 1.8 inches of precipitation in September reached the lowest sensors at 180 cm on the Cell 5S and 5N covers, however the sensors on the floor do not indicate any signs of the precipitation percolating thru the waste to the floor. Moisture from precipitation did not percolate below 150 centimeters (cm) (5 feet [ft]) in the vegetated final cover on the U-3ax/bl disposal unit at the Area 3 RWMS. A second remedial revegetation trial was done on the southern portion of the North Cover. In March 2015 it was determined that germination occurred on both the mulched and non-mulched treatments that were irrigated, but that no germination occurred on the non-irrigated treatments. By June the seedlings observed in March were gone, possibly grazed by rabbits which were observed inside the rabbit-proof fence. In August the entire area was sprayed with the herbicide glyphosate in an attempt to minimize weed seed production. During 2015, 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. Thirty-one percent of the applied precipitation and irrigation drained from the bare-soil drainage lysimeter, which received 3 times the natural precipitation. All 2015 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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Waste Handling and Emplacement Options for Disposal of Radioactive Waste in Deep Boreholes.

Traditional methods cannot be used to handle and emplace radioactive wastes in boreholes up to 16,400 feet (5 km) deep for disposal. This paper describes three systems that can be used for handling and emplacing waste packages in deep borehole: (1) a 2011 reference design that is based on a previous study by Woodward–Clyde in 1983 in which waste packages are assembled into “strings” and lowered using drill pipe; (2) an updated version of the 2011 reference design; and (3) a new concept in which individual waste packages would be lowered to depth using a wireline. Emplacement on coiled tubing was also considered, but not developed in detail. The systems described here are currently designed for U.S. Department of Energy-owned high-level waste (HLW) including the Cesium- 137/Strontium-90 capsules from the Hanford Facility and bulk granular HLW from fuel processing in Idaho.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗