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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Nuclear Structure and Decay Data for A=71 Isobars

Experimental nuclear spectroscopic data are evaluated for 12 known nuclides of mass number=71 (Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se, Br, Kr). Detailed compiled and evaluated information is presented for each reaction and decay experiment. The β – n decay of 72 Co to 71 Ni is included in this work, while for β – n decay of 71 Co to 70 Ni, consult Nuclear Data Sheets for A=70 (2016Gu11) or the ENSDF database for 70 Ni. Combining all the available data, recommended values are provided for energies, spins and parities, and half-lives of levels, with energies, branching ratios and multipolarities of γ radiations, and characteristics of β and α radiations in radioactive decays. Excited-states have not yet been identified in 71 Mn, 71 Fe, and 71 Kr, with the ground-state half-life remaining unknown only for 71 Mn; data for excited states in 71 Co and 71 Ni are very limited; 71 Ga, 71 Ge and 71 As are the most extensively studied nuclides via various reactions and decays, followed by 71 Cu, 71 Zn, 71 Se, and 71 Br, however, except for 71 Ge, the decay schemes of all other nuclides are considered as incomplete due to a large gap between the decay Q-value and the highest observed level. Lastly, this work supersedes earlier evaluations of A=71 by 2011Ab01.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1988

SN 1987A - Out on the tail

The expected light curve of Supernova 1987A is calculated for the next 10 yr based solely upon energy input from radioactive decay of Co-56 and Co-57. During the third year, energy input from Co-57 begins to dominate. How the bolometric (ultraviolet-optical-infrared) light curve deviates from that given by 100 percent deposition of the gamma-rays is a measure of the mixing that has gone on in the supernova. Current measurements favor a mixed model. The radioactive light curve decreases beneath 10 to the 39th ergs/s at an age of about 900 days, by which time the neutron star remnant should have made its presence known.

Pinto, Philip A.↗
DOE OSTI2020

Issue Resolution During the Development of the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20125

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high-activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance (i.e., 100 m from the SDUs). In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. A number of issues were overcome during the development of these simulations. These issues were identified as part of internal technical reviews. Simulations are developed by people and people make mistakes, so the internal technical review process is a vital step in PA development. Specific examples of resolved issues include a unit-conversion error, an inappropriate definition for a model boundary condition, a model time-stepping issue, and an error in the calculation for the buildup of contaminants in soil. Actions taken to address these issues resulted in an improved product with a better supported technical basis and more defensible results. The identification and correction of these issues are discussed. By understanding these issues, model developers and technical reviewers working on PAs in the future may avoid repeating these types of mistakes. Transparency with respect to these mistakes builds trust between waste management sites, regulators, and stakeholders. (authors)

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

Decay Curve Correction Analysis Report

The decay curve analysis that is done on the short-lived radionuclide gas samples is used to differentiate between gaseous radionuclides that have the same characteristic gamma decay energy, 511 kiloelectron-volts (keV). A sample of stack gas is isolated and the total counts in the 511 keV peak are counted repeatedly in 10-second intervals to evaluate the decay rate of the sample over time. Analysis of this decay data required a series of steps. First, a raw data report is generated by the gamma acquisition system, based on an analysis template within the acquisition software. The data report file was then loaded into Microsoft Word, and a macro was used to perform minor formatting (remove colons and insert tabs between data columns) to allow analysis within Excel. The file is then saved as a text file at this point. The text file is then uploaded into Excel and a series of macros are used to add labels, calculate radioactive decay constants, and analyze the gamma decay data using linear regression techniques. The analysis template has been used since 1998 for stack 53000303 (TA-53, building 0003, exhaust stack 03) and 2000 for stack 53000702. The overall process, including the gamma report format and the macros used in Word and Excel for processing the report, had remained unchanged until 2015. In October of 2015, staff made a change to the report template in the gamma acquisition software which resulted in an error in the calculations later performed by the Excel macro. This error was not caught until a more in-depth review of the analysis took place regarding 2020 data. This report covers a much more complete review of the issue that occurred regarding the decay curve analysis, a review of the calculations completed to correct the issue, a review of the updated decay curve analysis process, and recommendations for moving forward.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS2017

Cassini Power Subsystem

The electrical output of Cassini’s power system has been decaying consistently as predicted during its 20-year mission between October 1997 and September 2017. The power telemetry data is presented for the entire Cassini mission, including launch, cruise, and Saturn tour, up to the most recent available data. The spacecraft has been powered by three independent Radioisotope Thermoelectric Generators (RTGs) connected in parallel, which were able to generate 882 W at the beginning of the mission shortly after launch. The decrease in power energy output has mainly been driven by the heat reduction of the hot side of the RTGs due to the natural radioactive decay of its heat source plutonium (mostly plutonium-238 [238Pu]), degradation of thermoelectric material performance, and interface degradation.

Carr, Gregory A.↗
DOE OSTI2022

Building surrogate models of nuclear density functional theory with Gaussian processes and autoencoders

From the lightest Hydrogen isotopes up to the recently synthesized Oganesson (Z = 118), it is estimated that as many as about 8,000 atomic nuclei could exist in nature. Most of these nuclei are too short-lived to be occurring on Earth, but they play an essential role in astrophysical events such as supernova explosions or neutron star mergers that are presumed to be at the origin of most heavy elements in the Universe. Understanding the structure, reactions, and decays of nuclei across the entire chart of nuclides is an enormous challenge because of the experimental difficulties in measuring properties of interest in such fleeting objects and the theoretical and computational issues of simulating strongly-interacting quantum many-body systems. Nuclear density functional theory (DFT) is a fully microscopic theoretical framework which has the potential of providing such a quantitatively accurate description of nuclear properties for every nucleus in the chart of nuclides. Thanks to high-performance computing facilities, it has already been successfully applied to predict nuclear masses, global patterns of radioactive decay like β or γ decay, and several aspects of the nuclear fission process such as, e.g., spontaneous fission half-lives. Yet, predictive simulations of nuclear spectroscopy—the low-lying excited states and transitions between them—or of nuclear fission, or the quantification of theoretical uncertainties and their propagation to basic or applied nuclear science applications, would require several orders of magnitude more calculations than currently possible. However, most of this computational effort would be spent into generating a suitable basis of DFT wavefunctions. Such a task could potentially be considerably accelerated by borrowing tools from the field of machine learning and artificial intelligence. In this paper, we review different approaches to applying supervised and unsupervised learning techniques to nuclear DFT.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Long Term Assessment of Radiocesium in Deer, Fish, Soil, and Vegetation at the Savannah River Site

Throughout the Savannah River Site's (SRS) operational history, many radionuclides have been released into the environment. However, due to fallout from weapons testing, the amount released from SRS, its persistence in the environment, and the seclusion of the Site, radiocesium (Cs-137) is one of the most critical radionuclides in the SRS environment. The Effective half-life (Te) is the time it takes for a radionuclide in the environment to decrease by 50% of its initial amount. There are three means of radionuclide removal: Physical: Radioactive decay, sedimentation, and washout; Chemical: Changes in pH, oxidation state or adsorption; Biological: Changes in the food web or translocation. Effective half-life (T{sub e}) = log{sub e}2/λ{sub e} Where λ{sub e} is estimated from the slopes of loge-transformed Cs-137 activity concentrations regressed on year. The objective of this work was to calculate the effective half-life of Cs-137 in various biota at SRS and to perform an assessment of the changes in the effective half life over time. Compared to 2013 data, there is a slight increase in the effective half-lives for fish at most sampling locations. This effective half-life increase is an indication that the rate of Cs-137 loss is starting to plateau. The effective half-life of Cs-137 in SRS fish ranges from 4-11 years. The effective half-life of Cs-137 in soil at the Creek plantation is about 15 years. The effective half-life of Cs-137 in grassy vegetation at the Creek Plantation is about 14 years. The effective half-life of deer hunted on the SRS is about 15.5 years. After termination of all reactor activity by 1980, the Cs-137 activity began to drop drastically.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

First experimental determination of the ⁴⁰Ar(𝑛, 2𝑛)³⁹Ar reaction cross section and ³⁹Ar production in Earth’s atmosphere

The cosmogenic ³⁹Ar(t1∕2 = 268 years) isotope of argon is used for geophysical dating and tracing of underground and ocean water, as well as ice owing to its appropriate half-life and chemical inertness as a noble gas; ³⁹Ar serves also in nuclear weapon test monitoring. We measured for the first time the total cross section of the main ³⁹Ar cosmogenic production reaction in the atmosphere, namely ⁴⁰Ar(𝑛, 2𝑛)39 Ar, using 14.8 ± 0.3 MeV neutrons. The neutrons, produced by a deuterium-tritium generator, impinged on a stainless steel sphere filled with Ar gas highly enriched in the ⁴⁰Ar isotope and were monitored by a stack of fast-neutron activation foils. The reaction yield was measured by atom counting of long-lived ³⁹Ar with noble gas accelerator mass spectrometry and, independently, by decay counting relative to atmospheric argon (³⁹Ar/Ar= 8.12 × 10−16 ). A total ⁴⁰Ar(𝑛, 2𝑛)39 Ar cross section of 610 ± 100 mb was determined at 14.8 ± 0.3 MeV incident neutron energy. This result serves as a benchmark for recent theoretical calculations and evaluations, found to reproduce well the experimental total cross section. We use these energy-dependent theoretical cross sections together with experimental spectra of cosmogenic neutrons at different altitudes to calculate the global average rate of neutron-induced ³⁹Ar atmospheric production, resulting in 770 ± 240 39 Ar atoms/cm²/day. The secular equilibrium between the ³⁹Ar calculated production rate and radioactive decay rate leads to a partial isotopic abundance ³⁹Ar/Ar = (5.9 ± 1.8) × 10−16 , showing that ≈73% of atmospheric ³⁹Ar is produced by cosmogenic neutrons, the remaining part believed to be induced by muons and high-energy 𝛾 rays. The ⁴⁰Ar(𝑛, 2𝑛)³⁹Ar cross section at 14 MeV is also a key parameter for quantifying the anthropogenic contribution to atmospheric 39 Ar produced during the thermonuclear tests of the 1960s. We estimate that anthropogenic ³⁹Ar accounts for roughly 20% of the present atmospheric inventory.

39 Ar atmospheric production↗
DOE OSTI2024

DECOVALEX-2023: Task B Final Report

In all repository concepts for the geological disposal of radioactive waste, an engineered barrier system (EBS) is used to encapsulate the waste canister, or, to act as borehole or gallery seals. These systems are often based on bentonite clays due to their low permeability and high swelling capacity enabling the closure of engineering voids. However, in all repository concepts gases will be generated through the corrosion of metallic materials (under anoxic conditions), the radioactive decay of waste and the radiolysis of water. Thus, understanding the processes and mechanisms controlling the advective movement of gas (as a discrete phase) in clay-based materials is a key aspect when assessing the impact of gas flow in a repository safety case.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2018

Low Power Radioisotope Conversion Technology and Performance Summary

This work presents a summary of low-power radioisotope-based power sources. This includes systems that use direct energy conversion from radioactive decay such as betavoltaics and systems that convert heat generated by the isotope decay. These types of power systems have terrestrial and space-based applications. Their main application is to provide low levels of power in remote environments where long life and reliability are the main objectives. This work looked at the potential performance of these types of systems and the range of output power they can achieve. Also considered was the amount of power that could be achieved from these different conversion technologies, while remaining under the critical limit acceptable for space flight without the need for special handling and controls. This is of interest for space missions where low power levels can be used for operating sensors or other devices, but the expense associated with the incorporation of radioactive material above this critical level would be cost prohibitive.

Colozza, Anthony J.↗
DOE OSTI2020

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

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

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

Direct Mass Measurements to Inform the Behavior of 128m $\mathrm{Sb}$ in Nucleosynthetic Environments

Nuclear isomer effects are pivotal in understanding nuclear astrophysics, particularly in the rapid neutron-capture process where the population of metastable isomers can alter the radioactive decay paths of nuclei produced during astrophysical events. The β-decaying isomer 128m Sb was identified as potentially impactful since the β-decay pathway along the A = 128 isobar funnels into this state bypassing the ground state. Here we report the first direct mass measurements of the 128 Sb isomer and ground state using the Canadian Penning Trap mass spectrometer at Argonne National Laboratory. We find mass excesses of -84564.8(25) keV and -84608.8(21) keV, respectively, resulting in an excitation energy for the isomer of 43.9(33) keV. These results provide the first key nuclear data input for understanding the role of 128m Sb in nucleosynthesis, and we show that it will influence the flow of the rapid neutron-capture process.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS2006

Alpha-Voltaic Sources Using Liquid Ga as Conversion Medium

A family of proposed miniature sources of power would exploit the direct conversion of the kinetic energy of alpha particles into electricity. In addition to having long operational lives, these sources are expected to operate with energy-conversion efficiencies from 70 to 90 percent. A power source as proposed (see figure) would be an electrolytic cell in which liquid gallium would serve as both an electrolyte and an energy-conversion medium. The cell would contain an iridium cathode and a zirconium anode. The alpha particles, each with a kinetic energy approx.5.8 MeV, would be emitted by radioactive decay of Cm-244, which has a half-life of 18 years. The Cm-244 source would be positioned so that the a particles would enter the liquid gallium, where their kinetic energy would be dissipated mostly through ionization of Ga atoms, creating Ga(+) ions and free electrons. The electrons would be collected by iridium cathode, and the Ga(+) ions would be neutralized at the zirconium cathode by electrons returning after flowing through an external circuit. Gallium is a candidate for use as the electrolyte and the energy-conversion medium because in the liquid state it is a semimetal: its electrical conductivity is greater than that of a typical semiconductor but small in comparison with the conductivities of metals. Consequently, in liquid gallium, electrons and Ga(+) can exist without immediate recombination and can be moved by electric fields. It is expected that electric fields, resulting at least partly from the difference between the work functions of the electrode metals, would move the electrons and ions to their respective electrodes. The open-circuit potential of the cell is expected to be 1.62 V - equal to the difference between the work functions of iridium and zirconium. Unlike in a solid-state energy conversion medium, the impingement of energetic a particles would not give rise to displacement damage in the liquid gallium. Hence, the cell should have a long life, limited only by the half-life of Cm-244. A cell having a volume less than 25 cu mm, containing 1 curie of Cm-244 (the curie is a unit of radioactivity equal to 3.7 10(exp 10) disintegrations per second) is expected to deliver a current between 7 and 12 mA, which, at the expected open-circuit potential, would provide a power in the approximate range of 11 to 20 mW.

Patel, Jagdish U.↗
DOE OSTI2020

The TRANSCEND University Consortium: Theme 4: Nuclear Materials - 20431

The safe and secure management of Pu is a matter of international concern, with ∼250 t of separated Pu currently stockpiled worldwide. The UK's civil inventory of nuclear materials contains significant stocks of separated Pu from the reprocessing of Magnox and AGR spent fuels. The preferred option for the 138.5 tonnes of Pu is re-use as mixed oxide (MOx) fuel, although 5% is not suitable for re-use and is recommended for direct disposal. However, it will take more than 15 years to implement re-use, requiring that the Pu be kept in interim storage in its current state for that period, i.e. as PuO{sub 2} powder within inert steel storage cans at Sellafield. The focus of the work presented here is thus plutonium storage and the direct disposal of plutonium. The Research and Development needs of both are now pressing: in the case of storage due to it being the current default; in the case of immobilization and disposal because of a comparative lack of Research and Development on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, criticality, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on Pu bearing materials in these two contexts. In response to these needs, the TRANSCEND Consortium (Transformative Science and Engineering for Nuclear Decommissioning, a multi-disciplinary collaboration of 11 universities and 8 key industry partners from across the UK's civil nuclear sector) is seeking to provide technical underpinning to ongoing option development for the UK's civil Pu stockpile. Whilst understanding the behaviour of plutonium during its re-use as MOx is beyond the scope of the TRANSCEND Consortium work plan, the main objectives of the work are: (1) For interim storage: to understand how the surface structure and properties of pristine and radiation damaged PuO{sub 2} change with time in the absence and presence of water; and (2) For immobilization and disposal: to understand the mechanisms of incorporation of Pu into ceramic and glass-ceramic waste-forms, as well as the effect on these of self-induced radiation damage. Each objective is being addressed through separate work packages, the details of which are discussed in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Variation of lead isotopic composition and atomic weight in terrestrial materials (IUPAC Technical Report)

The isotopic composition and atomic weight of lead are variable in terrestrial materials because its three heaviest stable isotopes are stable end-products of the radioactive decay of uranium ( 238 U to 206 Pb; 235 U to 207 Pb) and thorium ( 232 Th to 208 Pb). The lightest stable isotope, 204 Pb, is primordial. These variations in isotope ratios and atomic weights provide useful information in many areas of science, including geochronology, archaeology, environmental studies, and forensic science. While elemental lead can serve as an abundant and homogeneous isotopic reference, deviations from the isotope ratios in other lead occurrences limit the accuracy with which a standard atomic weight can be given for lead. In a comprehensive review of several hundred publications and analyses of more than 8000 samples, published isotope data indicate that the lowest reported lead atomic weight of a normal terrestrial materials is 206.1462 ± 0.0028 (k = 2), determined for a growth of the phosphate mineral monazite around a garnet relic from an Archean high-grade metamorphic terrain in north-western Scotland, which contains mostly 206 Pb and almost no 204 Pb. The highest published lead atomic weight is 207.9351 ± 0.0005 (k = 2) for monazite from a micro-inclusion in a garnet relic, also from a high-grade metamorphic terrain in north-western Scotland, which contains almost pure radiogenic 208 Pb. When expressed as an interval, the lead atomic weight is [206.14, 207.94]. It is proposed that a value of 207.2 be adopted for the single lead atomic-weight value for education, commerce, and industry, corresponding to previously published conventional atomic-weight values.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

The Nature and State of Groundwater Contamination at the Nevada National Security Site: What Have We Learned from Decades of Groundwater Analysis? - 20337

The regulatory framework for remediating radionuclide contamination from underground nuclear testing at the Nevada National Security Site (NNSS) is based on a combination of characterization and modeling studies, monitoring, and institutional controls [1]. Currently, tritium is the largest contributor (∼90%) to the estimated 44.6 million-curie radionuclide inventory resulting from underground testing [2]. Because of its short half-life (12.32 years), its relative contribution reduces below 10% of the total radiologic inventory over the next 120 years as a result of radioactive decay. Although tritium levels are observed well above the Safe Drinking Water Act (SDWA) maximum contaminant levels (MCLs) in groundwater, other radionuclides are well below their MCLs except within the nuclear test near-field (nuclear test cavity and chimney) environment. In fact, most device-derived radionuclides are below their MCL in groundwater even in samples collected from this near-field environment. The distribution of radionuclides following the nuclear detonation greatly influences the availability of potential contaminants for groundwater transport. Tritium is initially distributed in the gas phase, later as tritiated water in steam, and finally as liquid water, and is available to groundwater transport away from the near-field environment. Other radionuclides that are mobile in groundwater are {sup 14}C, {sup 36}Cl, {sup 99}Tc, and {sup 129}I though their radiologic inventory is small when compared to tritium. Many radionuclides (e.g. U, Pu, Am) are incorporated to a significant extent into the melt glass at the bottom of the cavity and are accessible to groundwater primarily through the slow process of glass dissolution. These radionuclides are also adsorbed to the surfaces of the crushed rock within the cavity and chimney which limits their migration in groundwater. Although colloid facilitated transport of radionuclides at the NNSS has been observed [3], radionuclide concentrations decrease with time and migration distances due to desorption and colloid filtration processes. Current studies indicate that radionuclides associated with colloids are unlikely to migrate downgradient from NNSS underground nuclear tests at concentrations above the SDWA MCL [4][5]. The results of over 50 years of sampling, along with an understanding of these post detonation processes, indicate that tritium is the only contaminant of concern downgradient of testing and that even tritium will not exceed its MCL in groundwater after ∼120 years. Although other longer-lived radionuclides may continue to be released slowly from the near-field environment they will likely never reach levels exceeding their MCLs in groundwater downgradient of the NNSS. Groundwater monitoring will continue to verify these observations. (authors)

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

Radiochemistry and nuclear chemistry workforce in the United States

The disciplines of radiochemistry and nuclear chemistry have direct applications in the fields of national security, nuclear medicine, nuclear power production, and environmental management. Although, often, nuclear and radiochemistry are grouped together and many experts work in both areas, the definition for each field is slightly different. For example, radiochemistry may be defined as the application of the phenomena of radioactive decay and techniques common to nuclear physics so as to solve problems in the field of chemistry. In contrast, nuclear chemistry may be defined as the application of procedures and techniques common to chemistry to study the structure of the atomic nucleus. This chapter provides a brief update of the current state of, and critical U.S. needs for, nuclear chemistry and radiochemistry expertise as the Assuring a Future U.S.-Based Nuclear and Radiochemistry Expertise report was published by National Academy of Sciences (NAS) in 2012.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Steady-state radiochemical transport model of the molten salt reactor experiment

A radiochemical transport model of the Molten Salt Reactor Experiment (MSRE) was formulated based on the material derivative featuring system level 1-D flow providing spatial dependence accounting for multiple phenomena including radioactive decay, chemical removal, and fuel depletion across custom volume regions. Further, the steady-state version, which is based on a system of ODE's, can be solved efficiently and is useful for sensitivity studies utilizing Monte Carlo sampling. Accounting for multiple phenomena in a fully coupled model is necessary to understand the complexities of transport and distribution of radionuclides throughout the primary loop of a flowing fuel reactor.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗