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At least 1,333 records · Page 74
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δk eff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Hybrid Computing Technique for Nuclear Facility Design with Best Estimate Source Terms - 12313

When overall source term representing a variety of spent fuels or rad-wastes with different irradiation and decay profile is estimated, there are some restrictions in conventional analysis tools. A hybrid computing program that employs several functional modules was developed to support R and D action plan on an advanced fuel cycle adopting a pyro-process in connection with a sodium-cooled fast reactor in South Korea. Requirements for the advanced program were first defined, followed by development of several functional modules. With the developed modules, it was possible to carry out a source term assessment taking into account respective physical, irradiation, and aging characteristics of each spent fuel. And it was found that typical conservativeness in the design of nuclear facilities can be minimized through the hybrid technique and program proposed in this study. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS

This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.

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

Shell model analysis of the B ( E 2,2 + → 0 + ) values in the A = 70, T = 1 triplet 70 Kr, 70 Br, and 70 Se

The B(E2, 2 + → 0 + ) transition strengths of the T=1 isobaric triplet 70 Kr, 70 Br, 70 Se, recently measured at the RIKEN Radioactive Isotope Beam Factory (RIBF), are discussed in terms of state-of-the-art large scale shell model calculations using the JUN45 and JUN45 + LNPS plus Coulomb interactions. In this Letter we argue that, depending on the effective charges used, the calculations are either in line with the experimental data within statistical uncertainties, or the anomaly happens in 70 Br, rather than 70 Kr. In the latter case, we suggest that it can be due to the presence of a hitherto undetected 1 + T = 0 state below the yrast 2 + T = 1 state. Furthermore, our results do not support a shape change of 70 Kr with respect to the other members of the isobaric multiplet.

59 ≤ A ≤ 89↗
NASA NTRS1998

Re-examination of a Classic Experiment to Measure the Positronium-Helium Cross Section

In 1975, before the advent of positronium beams, a clever experiment was carried out in an attempt to measure low-energy Ps-atom scattering cross-sections, especially that of Ps-He. A series of parallel thin plates was placed in the vessel containing the sample of gas, and positrons were allowed to enter the chamber from a radioactive source. The time spectrum of annihilation radiation was observed in the traditional way, and the pickoff annihilation due to thermalized Ps striking one of the plates was to be the special feature of the experiment. Using a diffusion analysis the authors derived cross- sections for several gases, and for helium the value obtained was sigma = 0.0166 pi(alpha)(sub o, sup 2). Even then this value was thought to be very much too small, while recent measurements and some calculations suggest a more reasonable value would be sigma = 3 to 13 pi(alpha)(sub o,sup 2). It has always been puzzling that an apparently well-designed experiment should give such an unsatisfactory result. We have now re-examined the analysis of the data with some interesting consequences which may explain the discrepancy. Two main observations support our re-analysis. First, we note that the mean free path for Ps-He scattering would be quite long if the cross-section were as small as that quoted above; the diffusion method itself would seem to be questionable. For the larger values, however, there would be no such problem. Second, it was assumed that when the annihilation rate had settled down and was following an exponential decay curve the asymptotic solution of the diffusion equation had been reached. We find, instead, that a superposition of exponentials can accurately represent the decay curve, consistent with the higher cross-section value.

Drachman, Richard J.↗
DOE OSTI2021

Thermal Evaluation of the SAVY-4000 1 Quart Container at High Heat Loads

The SAVY-4000 Safety Analysis Report (SAR) was published in 2013 which established the design life of the container series through physical testing that meets applicable requirements from DOE M 441.1-1, Nuclear Material Packaging Manual. One critical requirement in DOE M 441.1-1 is to set a maximum heat load for the entire series to ensure that throughout the lifetime of the container, each of the components still performs within manufacture specifications. The original maximum decay heat limit has been defined within the SAVY-4000 SAR, which was accomplished by completing a series of tests that heat loaded the SAVY-4000 series and measured the steady state temperature and graded it against manufacture specifications. The container series is comprised of a corrosion resistant 316L stainless steel containment boundary, ceramic composite filter for the prevention of radiological particulate release and a chemically stable O-ring made of Viton ® . Currently, the SAVY 4000 maximum decay heat load is set at 25 Watts and the heat load expected during normal handling and storage is considered sufficient that the SAVY 4000 container requirements will not be the limiting factor for storage of most common material in use at TA-55, e.g. weapons grade oxide, americium, Pu-238 oxide and encapsulated heat sources. The SAVY 4000 lifetime was originally set as 5 years within the SAVY-4000 SAR, but in July 2019 the lifetime was extended to 15 years through a technical basis justification approved by the Department of Energy (DOE) Los Alamos Field Office. The 25 Watt limit applies to all SAVY 4000 container sizes; 1-quart, 2-quart, 3-quart, 5-quart, 8-quart, 12-quart, 5-gallon, and 10-gallon.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Investigations of dual-purpose canister direct disposal feasibility - 15106

Direct disposal of commercial spent nuclear fuel (SNF) in dual-purpose canisters (DPCs) could avoid the cost and complexity of re-packaging spent fuel into smaller purpose-built containers, avoid disposing of the DPC hulls as low-level waste, and potentially decrease worker dose. Technical objectives for disposal are: 1) safety of workers and the public, 2) engineering feasibility; 3) thermal management; and 4) criticality control after permanent closure of a repository. The postclosure safety case for DPC direct disposal would resemble that for any repository-waste isolation would be enhanced by choosing host geology in which radionuclide transport is diffusion dominated, and for which transport properties and pathways are insensitive to the expected temperature history. Containment functions would be assigned to the disposal overpack, and published data show that suitable materials exist for most possible disposal environments. Earlier studies showed that DPC-based waste packages would be only slightly larger and heavier than those proposed for a repository in volcanic tuff. Engineered solutions are available for transporting and emplacing them underground, although some could be the largest of their kind. The cost of DPC direct disposal in various geologic settings has been compared to options involving a change to loading standardized multi-purpose canisters, or continuing to load DPCs and re-packaging all of them for disposal. In general DPC direct disposal would cost tens of billions of dollars less, in packaging and disposal costs. Calculations for emplacement in hard rock and sedimentary rock show that host rock peak temperature targets of 100 to 200 deg.C could be met by managing fuel age and burnup, with repository closure at up to 150 years from reactor discharge. For disposal in salt, DPC-based waste packages could be emplaced at only 50 to 100 years from discharge, depending on burnup. Sedimentary media such as shales pose special challenges especially if they have low thermal conductivity. If backfill is used around DPC-based packages its peak temperature could be 150 to 200 deg.C. Without flooding of waste packages with ground water, criticality can never occur. However, some small number of disposal overpacks could fail within the postclosure performance period due to corrosion or disruptive events. Once flooded, the aluminum-based neutron absorber materials used in most DPCs would readily degrade. Analysis has shown that even without neutron absorbers, virtually all DPCs could be subcritical if flooded with chloride brines such as would occur in a salt repository. The licensing basis for existing DPCs involves criticality analysis that may be conservative compared to fuel that is actually loaded, and the difference provides uncredited reactivity margin. Uncredited margin can offset increases in reactivity associated with flooding and basket degradation. For flooding with fresh water, with loss of neutron absorbers, many DPCs have been shown to be subcritical using uncredited margin. Significantly fewer are subcritical if the fuel basket also degrades. Thus, the materials of DPC basket construction (e.g., aluminum or carbon steel) could be deleterious if they fully corrode, along with the neutron absorbers, during the performance period. Logistical simulations are used to better understand the relationship between the needed DPC decay storage time for disposal, and the timing of future events such as the repository opening date, or a transition to loading multi-purpose (disposable) canisters at nuclear power plants. Results show that the maximum benefit from implementing a new canister design, with respect to shortening the cooling time for all SNF including that in DPCs, is associated with the earliest repository start date. As time passes more of the total SNF inventory will be in DPCs such that the value of a transition to multi-purpose canisters will decline, and the remaining options are DPC direct disposal and re-packaging in purpose-designed disposal canisters. Fuel age at emplacement is of interest to evaluate potential impacts from future changes in the fuel management system that could place fuel age limits on dry storage or transportation. The minimum fuel age at emplacement (best-case) is obtained by re-packaging all DPCs into smaller canisters for disposal, thus drastically decreasing the needed decay storage time for any geologic disposal setting. If the industry transitions to smaller, multi-purpose canisters, and without re-packaging, the fuel age at emplacement would be comparable to the best case if: 1) the emplacement power limit is high enough, or 2) both the transition and the repository start date occur early (e.g., 2036 was analyzed). Technical feasibility evaluation continues to show that direct disposal could be safe and cost effective, for some DPCs and in some geologic settings. Direct disposal of a substantial fraction of existing DPCs, limited mainly by thermal management, uncredited criticality margin, materials of DPC basket construction, and the salinity of ground water in the disposal environment, could be part of the overall waste management system. (authors)

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

Stabilization of Preternatural Barium Oxidation States as an Unexpected Byproduct of β-Decay: Discovery of a New Halide Semiconductor Alloy

137 Cs has a wide range of roles in the nuclear industry. The solid material, safely encapsulated in CsCl as 137 CsCl, is stored as fission product waste from nuclear power production and legacy waste from nuclear weapons production; it has also served as a radiation source in food and sewage irradiators as well as medical devices. However, because of the solubility of the chloride salt and the relatively high specific activity of 137 Cs, damaged or broken capsules can lead to severe radiological accidents. Safe capsule design and material recycling are complicated by the unclear structural evolution during β-decay, which remains ambiguous due to the differing oxidation states of Cs (1+) and Ba (2+). Here, in this study, we use first-principles calculations to investigate the evolving structure–property relationships of Cs 1–x Ba x Cl during β-decay. Despite the well-established 2+ formal oxidation state of alkali-earth metals, we find that Ba 1+ can be stabilized in the form of a mixed-valence alloy at low concentrations. Specifically, we identify three regimes for the β-decay of 137 Cs into CsCl: Ba-doped CsCl (Ba ≤ 14%), wherein Ba has the expected 2+ oxidation state; Cs–Ba–Cl alloys, where Ba has a mix of the usual Ba 2+ and highly unusual Ba 1+ oxidation state in the form of a quasi-disordered mixed-valence alloy (Ba = 25%); and phase separation into a CsCl + BaCl 2 + Ba (m) mechanical mixture, where Ba reverts to its expected 2+ oxidation state (Ba > 25%). Surprisingly, the Cs 0.75 Ba 0.25 Cl mixed-valence alloy is a narrow indirect band gap semiconductor (1.05 eV) despite the insulating nature of both CsCl and BaCl 2 . It also exhibits strongly excitonic polarized optical properties, has glass-like ultralow thermal conductivity (directional average of 0.21 W/mK at 300 K), and shows greater resistance to deformation under both tensile and volumetric strengths compared with the original CsCl structure (e.g., shear and Young’s modulus of 9.04 and 31.62 GPa, respectively). These findings imply that transmutation of 137 Cs leads to highly unusual chemical bonding that stabilizes Ba 1+ in local regions of the quasi-disordered Cs 0.25 Ba 0.75 Cl, resulting in anomalous physical properties. Moreover, this discovery provides valuable insight for safe nuclear waste capsule design, which can aid in preventing environmental or human exposure to radioactive materials.

Fuhr, Addis S. [Oak Ridge National Laboratory (ORN↗
DOE OSTI2019

Characterization and Tritium Aging of LANA.85 for Regenerative Testing

LANA.85 has been suggested for use in both the Four-Inch Short Hydride (FISH) as well as the Forced Atmosphere Cooled Electrically heated (FACE) bed designs that may be implemented in the Tritium Facilities. LANA.85 has a lower plateau pressure than LANA.75 at a given temperature and composition. Though data is limited, previous tritium aging studies initiated on LANA.85 suggest that radiolytic decay of tritium to He-3 within the metal matrix causes performance degradation similar to LANA.75. Tritium aging studies on LANA.85 are needed to characterize behavior on a small scale before implementation occurs on a large scale. Characterization and tritium-aging of LANA.85 for Regeneration Testing is a multi-year project. Objectives included: fabrication of an inherently safe test cell, characterization of the sample via deuterium and tritium isotherms, and initiation of tritium aging of the sample. During FY19, baseline tritium desorption isotherms were collected at 100, 125, and 150 °C on the LANA85 sample. This was the same sample used for deuterium testing in FY18. Following testing, the sample was reloaded with tritium for future aging studies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Computational scheme for burnup credit applied to long term waste disposal - 14448

The application of burnup credit (BUC) to criticality safety evaluations (CSE) of geological repositories for long term disposal of spent nuclear fuel (SNF) is under consideration in Switzerland. Therefore, in collaboration with the national waste management organization, PSI is conducting research towards the development of a BUC methodology tailored for Swiss specific applications. As part of this research, a computational sequence from the in-core burnup phase, through the interim storage, until final disposal is being established. The tools selected for this methodology, referred to as BUCSS-R (Burnup Credit System for Swiss reactors and Repositories), are CASMO for the in-core fuel depletion, the decay module of SERPENT for the decay calculations, and MCNP for the criticality calculations. In connection to BUCSS-R, a computational interface to validated reference core models of the Swiss reactors is being developed. The purpose is to apply reconstructed depletion histories based on real core operation conditions with CASMO assembly calculations, to transfer the pin wise nuclide compositions at discharge to the subsequent SERPENT decay calculations, and afterwards to the final MCNP criticality calculations. An objective of this paper is to present the elements of the BUCSS-R methodology, and to show a first application to so-called bounding analyses of a preliminary Swiss disposal canister design loaded with fuel assembly types operated in a Swiss PWR reactor. The rationale for these bounding analyses is that before starting the integration of methodologies for bias and uncertainty quantifications, the span of the k-eff variation between fresh and burnt fuel loadings should first be quantified in a realistic manner. The obtained results are reported in this paper together with studies performed in order to evaluate the effects from burnup levels and nuclides credit at flooded conditions on the k-eff values. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Exploring Soft Donor Character of the N -2-Pyrazinylmethyl Group by Coordinating Trivalent Actinides and Lanthanides Using Aminopolycarboxylates

The trivalent f-element coordination chemistry of a novel aminopolycarboxylate complexant was investigated. The novel reagent is an octadentate complexant that resembles diethylenetriamine-N,N,N',N",N"-pentaacetic acid (DTPA), but a single N-acetate pendant arm was substituted with a N-2-pyrazinylmethyl functional group. Thermodynamic studies of ligand protonation and trivalent lanthanide, americium and curium, complexation by N-2-pyrazinylmethyldiethylenetriamine-N,N',N",N"-tetraacetic acid (DTTA-PzM) emphasize the strong electron withdrawing influence of the N-2-pyrazinylmethyl group. Particularly, DTTA-PzM is more acidic compared to a N-2-pyridinylmethyl-substituted structural equivalent, DTTA-PyM, with a substantial lowering of pK7, corresponding to the protonation of a second aliphatic amine site. The participation of the pyrizyl nitrogen in the metal ion coordination sphere is observed from the fluorescence lifetime decay measurements of metal hydration and the interpretation of the stability constants for ML– and MHL(aq) complexes. The overall conditional stability constants for the trivalent f-element complexation by DTTA-PzM complexes decrease, relative to DTTA-PyM, as expected based on the lower basicity of pyrazine in water relative to pyridine. Replacement of the N-2-pyridinylmethyl group with N-2-pyrazinylmethyl, while enhancing the total acidity of DTTA-PzM, also reduces its softness, as manifested by a small lowering of β 101 Am/Nd and liquid–liquid separation of trivalent lanthanides from trivalent americium. Despite this, the 4f/5f separation is doubled when DTTA-PzM replaces DTPA as an aqueous complexant in solvent extraction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can be accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized. Results will be presented showing the potential for thorium to reach a high TRU transmutation rate over a wide variety of fuel types (oxide, metal, nitride and carbide) and transmutation schemes (recycle or partition of in-bred U-233). In addition, a sustainable scheme has been devised to burn the TRU accumulated in the core inventory once the legacy TRU supply has been exhausted, thereby achieving long-term virtually TRU-free. A comprehensive 'back-to-front' approach to the fuel cycle has recently been proposed by Westinghouse which emphasizes achieving 'acceptable', low-radiotoxicity, high-level waste, with the intent not only to satisfy all technical constraints but also to improve public acceptance of nuclear energy. Following this approach, the thorium fuel cycle, due to its low radiotoxicity and high potential for TRU transmutation has been selected as a promising solution. Additional studies not shown here have shown significant reduction of decay heat. The TRU burning potential of the Th-based fuel cycle has been illustrated with a variety of fuel types, using the Toshiba ARR to perform the analysis, including scenarios with continued LWR operation of either uranium fueled or thorium fueled LWRs. These scenarios will afford overall reduction in actinide radiotoxicity, however when the TRU supply is exhausted, a continued U- 235 LWR operation must be assumed to provide TRU makeup feed. This scenario will never reach the characteristically low TRU content of a closed thorium fuel cycle with its associated potential benefits on waste radiotoxicity, as exemplified by the transition scenario studied. At present, the cases studied indicate ThC as a potential fuel for maximizing TRU burning, while ThN with nitrogen enriched to 95% N-15 shows the highest breeding potential. As a result, a transition scenario with ThN was developed to show that a sustainable, closed Th-cycle can be achieved starting from burning the legacy TRU stock and completing the transmutation of the residual TRU remaining in the core inventory after the legacy TRU external supply has been exhausted. The radiotoxicity of the actinide waste during the various phases has been characterized, showing the beneficial effect of the decreasing content of TRU in the recycled fuel as the transition to a closed Th-based fuel cycle is undertaken. Due to the back-to-front nature of the proposed methodology, detailed designs are not the first step taken when assessing a fuel cycle scenario potential. As a result, design refinement is still required and should be expected in future studies. Moreover, significant safety assessment, including determination of associated reactivity coefficients, fuel and reprocessing feasibility studies and economic assessments will still be needed for a more comprehensive and meaningful comparison against other potential solutions. With the above considerations in mind, the potential advantages of thorium fuelled reactors on HLW management optimization lead us to believe that thorium fuelled reactor systems can play a significant role in the future and deserve further consideration. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Stopped, Bunched Beams for the TwinSol Facility

Tests of the unitarity of the Cabbibo-Kobayashi-Masakawa (CKM) matrix offer an important avenue for constraining the Standard Model of the electroweak interaction. Several methods are currently used to determine V-ud, the largest element in the top-row normalization test. One such method is through the study of superallowed T =1/2 mixed mirror transitions, which offers a complementary method to the current most-precise value that is determined from superallowed pure Fermi 0(+) 0(+) transitions. The precision currently achievable by this method is currently limited by the very low number of transitions for which the Fermi-to-Gamow-Teller mixing ratio has been measured. St. Benedict, the Superallowed Transition Beta-Neutrino Decay-Ion-Coincidence Trap, is currently under development at the University of Notre Dame's Nuclear Science Laboratory, and intends to determine for a range of new isotopes through measurements of the -neutrino asymmetry parameter a using a linear Paul trap. In order to trap these ions, the fast, continuous secondary beam separated by the TwinSol twin solenoid separator must be thermalized and bunched. The system through which this will be done will feature a large-volume gas cell in which the ions will be thermalized, a double-RF-funnel-based ion guide system for the extraction of the ions, and a radiofrequency quadrupole (RFQ) to provide cooled ion bunches for capture in the Paul trap.

beam cooling↗
DOE OSTI2020

PORFLOW modeling of Vadose zone flow and transport for the E-area intermediate level vault

In support of the E-Area Performance Assessment, a two-dimensional model of water flow and radionuclide transport through the E-Area Intermediate Level Vault (ILV) and local vadose zone has been developed using the PORFLOWTM software. The purpose of the model is to calculate flux to the water table for radionuclides eluted from the ILV during its operational life, the period of institutional control, and times following site closure. Results of model calculations will be used by a three-dimensional PORFLOW model of transport through the aquifer to determine radionuclide concentrations at a hypothetical 100 meter well and at the site boundary where contaminated groundwater could be assessed by members of the public following site closure. While newly developed, the model structure closely follows that used in the 2008 PA while incorporating a refined computational mesh, updated material properties for the vadose zone soil and vault concrete, revised infiltration rates, and new solid-liquid distribution coefficients. The model also addresses degradation of concrete hydraulic properties by blending soil and concrete water retention curves over a 500 year period. That is, the hydraulic properties of ILV concrete start out as fresh concrete at the time of site closure and degrade to soil properties over a 500 year time period. This approach has not been used previously in E-Area PA’s. The model has been used to calculate water flow through the ILV vadose zone and, on a trial basis, transport of a limited number of radionuclides from the waste region inside the ILV to the water table. Figure ES-1 provides a view of the model and an example of water flow after the concrete has substantially degraded. Figure dimensions are in centimeters. The red area in the left hand side figure is the waste disposal region within the ILV. Colored bands in the figures indicate different soil and concrete regions. Figure ES-2 shows a calculation of flux to the water table (mol/yr) for U-234 and its one-year half-life decay chain (U-234 → Th-230 → Ra-226 → Pb-210). Pb-210 flux is too small to show on the same scale used for Th-230 and Ra-226 in Figure ES-2. The model has also been used to make a preliminary evaluation of the release of tritium from TPBAR disposal containers that will be place in the ILV. A separately calculated tritium release rate as a function of time was used as a source in the PORFLOW model. The TPBAR disposal schedule and tritium release rate differ from the values used in the 2008 PA. Results from these initial trial runs are compared to calculations made for the 2008 PA ILV analysis. In brief, a new PORFLOW model of flow and transport through the ILV vadose zone has been developed and tested. The model is ready to be used for PA calculations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Time projection chamber for GADGET II

The established Gaseous Detector with Germanium Tagging (GADGET) detection system is used to measure weak, low-energy 𝛽-delayed proton decays. It consists of the Gaseous Proton Detector equipped with a MICROMEGAS (MM) readout to detect protons and other charged particles calorimetrically, surrounded by the Segmented Germanium Array (SeGA) for high-resolution detection of prompt 𝛾 rays. To upgrade GADGET's Proton Detector to operate as a compact time projection chamber (TPC) for the detection, three-dimensional imaging and identification of low-energy 𝛽-delayed single- and multiparticle emissions mainly of interest to astrophysical studies. A new high granularity MM board with 1024 pads has been designed, fabricated, installed, and tested. A high-density data acquisition system based on generic electronics for TPCs (GET) has been installed and optimized to record and process the gas avalanche signals collected on the readout pads. The TPC's performance has been tested using a 220 Rn 𝛼-particle source and cosmic-ray muons. In addition, decay events in the TPC have been simulated by adapting the attpcroot data analysis framework. Furthermore, a novel application of two-dimensional convolutional neural networks for GADGET II event classification is introduced. The optimization of data throughput is also addressed. The GADGET II TPC is capable of detecting and identifying 𝛼 particles as well as measuring their track direction, range, and energy. The extracted energy resolution of the GADGET II TPC using P10 gas is about 5.4% at 6.288 MeV ( 220 Rn 𝛼 events), computed using charge integration. Based on a systematic simulation study, we estimated the detection efficiency of the GADGET II TPC for protons and 𝛼 particles, respectively. It has also been demonstrated that the GADGET II TPC is capable of tracking minimum-ionizing particles (i.e., cosmic-ray muons). From these measurements, the electron drift velocity was measured under typical operating conditions. In addition to being one of the first generation of micropattern gaseous detectors (MPGDs) to utilize a resistive anode applied to low-energy nuclear physics, the GADGET II TPC will also be the first TPC surrounded by a high-efficiency array of high-purity germanium 𝛾-ray detectors. As a result, the TPC of GADGET II has been designed, fabricated, and tested and is ready for operation at the Facility for Rare Isotope Beams for radioactive-beam-line experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Influence of Nanoconfinement and Elevated Temperatures on Geocolloid-Facilitated Transport of Energy-related Contaminants

The main objective of this project is to obtain a better understanding of the transport behavior and interactions of geocolloids in the presence of energy- related contaminants under bulk and nanoconfined conditions. In that, we first established solid protocols to synthesize and fabricate numerous types of geocolloids with a. various sizes ranging from 40 – 800 nm and b. different hydrophile-lipophile balance (HLB) ratios in the range of 25:75 to 75:25. Direct force measurements with geocolloids having different degrees of surface coverages (i.e. HLB to realistically mimic adsorption of energy related contaminants) were conducted using the Surface Forces Apparatus (SFA) over a distance regime starting from 8μm all the way down to molecular contact. Repulsive forces were observed on approach starting from > 3μm, followed by an exponential increase of which magnitude appears to be larger than a decay length obtained from Derjaguin–Landau–Verwey–Overbeek (DLVO) theory in pure water. When the geocolloids were confined in salted water, the magnitude of onset of repulsion was varied as a function of salinity in solution, which can significantly alter a purely repulsive screened electrostatic (coulombic) interaction arising from, among geocolloids as well as between geocolloids and geosurfaces. The viscosity and flow characteristics of geocolloidal suspensions at different degrees of confinement were also investigated where we identified highly discontinuous rheological behaviors below a critical nanoconfinement level. We anticipate that the knowledge gained through this study will enable the scientists and researchers to better assess transport and fate behaviors of geocolloidal dispersions that can carry energy-related contaminants under realistically emulated geosystem.

03 NATURAL GAS↗
DOE OSTI2012

Disposition of Radioisotope Thermoelectric Generators Currently Located at the Oak Ridge National Laboratory - 12232

Under the American Recovery and Reinvestment Act (ARRA), the U.S. Department of Energy (DOE) awarded SEC Federal Services Corporation (SEC) a 34-building demolition and disposal (D and D) project at the Oak Ridge National Laboratory (ORNL) that included the disposition of six Strontium (Sr-90) powered Radioisotope Thermoelectric Generators (RTGs) stored outside of ORNL Building 3517. Disposition of the RTGs is very complex both in terms of complying with disposal facility waste acceptance criteria (WAC) and U.S. Department of Transportation (DOT) requirements for packaging and transportation in commerce. Two of the RTGs contain elemental mercury which requires them to be Land Disposal Restrictions (LDR) compliant prior to disposal. In addition, all of the RTGs exceed the Class C waste concentration limits under Nuclear Regulatory Commission (NRC) Waste Classification Guidelines. In order to meet the LDR requirements and Nevada National Security Site (NNSS) WAC, a site specific treatability variance for mercury was submitted to the U.S. Environmental Protection Agency (EPA) to allow macro-encapsulation to be an acceptable treatment standard for elemental mercury. By identifying and confirming the design configuration of the mercury containing RTGs, the SEC team proved that the current configuration met the macro-encapsulation standard of 40 Code of Federal Regulations (CFR) 268.45. The SEC Team also worked with NNSS to demonstrate that all radioisotope considerations are compliant with the NNSS low-level waste (LLW) disposal facility performance assessment and WAC. Lastly, the SEC team determined that the GE2000 Type B cask met the necessary size, weight, and thermal loading requirements for five of the six RTGs. The sixth RTG (BUP-500) required a one-time DOT shipment exemption request due to the RTG's large size. The DOT exemption justification for the BUP-500 relies on the inherent robust construction and material make-up of the BUP- 500 RTG. DOE-ORO, SEC, and the entire SEC RTG team are nearing the conclusion of the Sr-90 RTG disposition challenge - a legacy now 50 years in the making. Over 600,000 Ci of Sr-90 waste await disposal and its removal from ORNL will mark an historical moment in the clean-up of the cold-war legacy in the ORNL central industrial area. Elimination (i.e., removal) of the RTGs will reduce security risks at ORNL and disposal will permanently eliminate security risks. The RTGs will eventually decay to benign levels within a reasonable timeframe relative to radiological risks posed by long-lived isotopes. The safety authorization basis at ORNL Building 3517 will be reduced enabling greater operational flexibility in future clean-out and D and D campaigns. Upon disposition the Department of Energy will realize reduced direct and indirect surveillance and maintenance costs that can be reapplied to accelerated and enhanced clean-up of the Oak Ridge Reservation. At present, waste profiles for the RTGs are developed and under review by NNSS RWAP staff and approval authorities. Disposition schedule is driven by the availability of compliant shipping casks necessary to safely transport the RTGs from ORNL to NNSS. The first disposal of the RCA RTG is expected in April 2012 and the remaining RTGs disposed in 2012 and 2013. (authors)

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

Achieving the End State for the Pahute Mesa Corrective Action Units at the Nevada National Security Site - 20355

Underground nuclear testing at the Nevada National Security Site (NNSS) ended in 1992. To address the impact of radionuclide contamination from underground nuclear testing on the groundwater resources of Nevada, the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program created the Underground Test Area (UGTA) activity to characterize the radionuclides in groundwater, understand the nature and extent of radionuclide migration, and forecast the distribution of radionuclides in groundwater for 1,000 years. The UGTA activity is regulated by the Federal Facility Agreement and Consent Order (FFACO), an agreement negotiated between the Nevada Division of Environmental Protection (NDEP), DOE, and the U.S. Department of Defense (DoD). DOE is close to achieving the end state (closure in place with monitoring and institutional control) for three of the five UGTA corrective action units (CAUs) on the NNSS. A number of challenges remain to achieve the end state of the other two CAUs, both located on Pahute Mesa. Pahute Mesa was the location of 82 underground nuclear tests, less than 10% of the total number of tests on the NNSS. Yet, Pahute Mesa contains slightly more than 60% of the total radionuclides (in curies). Based on groundwater sampling in monitoring wells, two radionuclide plumes have migrated several kilometers (km) in groundwater from selected test cavities and have crossed the boundary of the NNSS (yet remain within the boundaries of Federally controlled land). The path to achieve the end state continues to evolve as more data are collected and a better understanding of radionuclide migration in groundwater is developed. DOE has invested in drilling and sampling more than 50 characterization and monitoring groundwater wells over the past 25 years. The data indicate that the primary radionuclide of concern is tritium as it is about 89% of the total radionuclide inventory (in curies). As well, only tritium has been measured in groundwater outside of cavities at concentrations exceeding the Safe Drinking Water Act (SDWA) standard. For one of the tritium plumes that has migrated across the NNSS boundary, the average rate of migration has been measured as about 45 meters (m) per year with the rate of migration at the leading edge of the plume of about 85 meters per year. At that rate of migration, the tritium plume will decay to safe levels and not reach the publicly accessible environment at concentrations above the SDWA standard. Other radionuclides, at concentrations below the SDWA standards, will be monitored to ensure they remain at safe levels. Consequently, the end state path forward for Pahute Mesa has evolved to take full advantage of the data from the monitoring network to constrain uncertainty in model forecasts and to reduce reliance on probabilistic simulations. The focus of the end state approach relies on developing a monitoring well network that is protective of human health by increasing confidence that no radionuclide plume in the groundwater will migrate undetected to the accessible environment, located about 22 km from the nearest up-gradient underground nuclear test. Using the measured data to remove uncertainty, the evaluation of radionuclide migration from Pahute Mesa is directed toward meeting the goals of the end state. (authors)

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