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DOE OSTI2019

Numerical Investigation of the Electrical Resistivity Tomography for Detecting the Waste Zone beneath an Engineered Surface - 19066

Surface barriers are an integral part of waste site remediation that are needed to minimize further contamination spread, allow time for additional radionuclide decay, and decrease risks to workers and the environment. Depending on the type of waste, the design life of a surface barrier varies from decades to millennia. The performance and impacts of surface barriers must be monitored beyond the life of most monitoring instruments; hence, long-term monitoring technologies are needed that can non-intrusively assess barrier impacts both in the near surface and to large depths. Innovative geophysical methods such as the electrical resistivity tomography (ERT) are capable of nonintrusively detecting the distribution of soil wetness and/or contaminant plumes in the soil. However, the application of geophysical methods in monitoring a surface barrier system has been very limited because a surface barrier is often composed of multiple components made of natural (e.g., soil, rocks, asphalt) and manufactured (e.g., plastic, metal) materials. In this study, the E4D numerical simulator, developed at Pacific Northwest National Laboratory for inverting the ERT survey data, was used to evaluate the configuration of electrodes for optimized ERT monitoring of waste zone below a surface barrier. The results indicate that, for high-quality mapping of the conditions below the surface barrier, the ERT method requires that some of the electrodes be installed either beneath the surface barrier or in boreholes around the barrier. In reality, for a surface barrier that is being constructed, some electrodes may be installed beneath the non-conductive layer of the surface barrier before or during barrier construction. For a surface barrier that has already been constructed, some electrodes can be installed in a few vertical boreholes around the barrier boundary. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Demonstration of the On-the-Fly Shielding Analysis Method: Spent Fuel and Waste Disposition

This report documents work performed supporting the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management activities under work breakdown structure element 1.08.02.04.01, “Data and Tools Development, Validation, and Maintenance.” In particular, this report fulfills milestone M3SF-21OR020401016, “Implement on-the-fly dose analysis methodology in UNF-ST&DARDS” within work package SF-21OR02040101, “Commercial SNF Characterization - ORNL.” The Used Nuclear Fuel - Storage, Transportation & Disposal Analysis Resource and Data System (UNFST& DARDS) enables automated dose rate calculations for spent nuclear fuel (SNF) transportation packages and storage casks using a Monte Carlo radiation transport code. The explicit method uses a detailed model of the SNF system and its contents. Therefore, a dose rate calculation is required for each as-loaded transportation package or storage cask because the SNF assemblies within a canister typically have unique irradiation characteristics. An alternate method, referred to as the “on-the-fly” shielding analysis method, has been proposed that requires only a set of Monte Carlo dose rate calculations for each transportation packaging/storage cask design. The results of the Monte Carlo dose rate calculations are independent of the SNF assembly irradiation and decay characteristics. The dose rate values may then be combined with the radiation source strength of the SNF assemblies associated with a particular transportation packaging/storage cask design to determine actual dose rates. This report presents on-the-fly dose rate calculations for a representative SNF storage cask and verification of the on-the-fly dose rate calculation results by comparison with reference dose rate calculations using the explicit Monte Carlo dose rate calculation. The on-the-fly shielding analysis method was implemented in UNF-ST&DARDS. A Python program was developed to process the MAVRIC dose rate results obtained by source particle type, energy group, and fuel geometry region. A Python processor created binary files, which were saved as a special UNF-ST&DARDS library for on-the- fly shielding analyses. UNF-ST&DARDS uses the precalculated on-the-fly binary libraries generated by the Python data processor and directly executes the Python code for on-the-fly dose analysis. This Python code unzips the pre-generated binary files mentioned above, reads the data, and combines them with user-specified sources for dose and uncertainty calculations. The Python programs were verified using Excel calculations and by comparison with the values obtained with the MAVRIC post-processing utilities applied to the 3dmap files. This method can currently be used to determine dose rates for as-loaded HI-STORM FW storage casks. The UNF-ST&DARDS analysis wizard for on-the-fly shielding analysis is described in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Non-Neutron Transmutation of Used Nuclear Fuel (Final Report)

The primary goal of this study is to develop a national facility concept for transmuting long-lived fission products (LLFP) to substantially reduce the disposal impact by minimizing the need for a geologic-timescale repository. As a charter for this study, the national transmutation facility was required to reduce the radiotoxicity and decay heat of LLFP isotopes by at least 90% relative to their values at discharge from a commercial LWR, while consuming less than 10% of the reactor's energy. The identified LLFP isotopes are Se-79, Zr-93, Tc-99, I-129, Sn-126, and Cs-135, whose radiotoxicity is about 99% of the total radiotoxicity of all fission products at 1,000 years. Approximately ~72 kg of LLFPs is discharged every year from a 1,000 MWe commercial or advanced nuclear reactor. First, LLFP transmutation options with non-neutron beams (photons and protons) were explored. The study concluded that LLFP transmutation is feasible with high-energy, high-intensity photons or protons, but impractical on an engineering scale due to low transmutation rates and the high energy requirements to produce the desired photon or proton beams. As alternatives, LLFP transmutation options with neutrons from fission, fusion, and spallation reactions were additionally explored. The transmutation options using advanced critical reactors are attractive only for selective LLFP isotopes because the production rates of several LLFP isotopes (Zr-93, Sn-126, and Cs-135) from fission reactions are larger than the transmutation rates. The transmutation options with only spallation neutrons are favorable to transmute all LLFP isotopes, but as a tradeoff, the net transmutation rates are reduced. The national transmutation facility concept was developed following an exploration of transmutation options using various incident particles. The proposed national LLFP transmutation comprises a dedicated molten-salt reactor (MSR), a proton accelerator, and a spallation neutron-based transmuter. The MSR power was set at 300 MWt and 120 MWe, with the thermal power approximately 10% of that of a commercial 1,000 MWe PWR. The electricity generated by the MSR powers the accelerator and transmuter. The accelerator produces 1 GeV, 30 mA protons, which are introduced into the spallation neutron-based transmuter. The spallation neutron-based transmuter consists of a central spallation target and LLFP target pins merged in a heavy water tank. The six LLFP isotopes are separated into two groups. Tc-99, I-129, and Se-79, having larger neutron cross sections, belong to group A, while Zr-93, Sn126, and Cs-135, having smaller neutron cross sections, belong to group B. Then, for effective transmutation, LLFPs in groups A and B are transmuted in the dedicated MSR and in a spallation neutron-based transmuter, respectively. The estimated capital cost of the national transmutation facility is approximately $\$$3.1B, and its annual O&M cost is expected to be ~$\$$182M. Radiotoxicity and decay heat of LLFPs were calculated and compared with those of the original LLFPs. It was assumed that the targets were made with elementwise LLFP rather than isotopic LLFP, owing to the potentially high cost of isotopic separation from used nuclear fuels. The decay heat of LLFPs can be reduced by more than 90% using a single national transmutation facility. However, radiotoxicity decreases by 79–84%, which does not meet the transmutation performance requirement, primarily because Cs-135 is produced rather than depleted. Thus, to meet the design requirement, Cs-135 should be separated from other Cs isotopes and irradiated in a spallation neutron-based transmuter. Then, radiotoxicity decreases by ~92%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Modeling the Drying of a Mock Used Fuel Assembly with COBRA-SFS Cycle-4 - 18597

The processing and storage of nuclear fuel is an ever increasingly pressing issue. Around 70,000 tons of used fuel are currently in either dry or wet storage on site at nuclear power plants or at specified spent fuel storage facilities. Additionally, about 2,000 tons are added to that sum each year. To promote retrievability, limit degradation in the canister, and reduce potential for hydrogen generation, spent fuel that is initially in wet storage must be dried before moving to dry cask storage. Therefore, this research seeks to use a mock used fuel storage cask to replicate the conditions in the industry during the drying process of fuel being moved from storage pools. To simulate this, a mock storage cask has been constructed to closely resemble those used in industry such that any areas that tend to retain water are replicated effectively. It is able to hold one mock used fuel assembly. The assembly used is an AREVA 10A 10 x 10 assembly with 78 rods filled with depleted uranium, 12 heater rods to simulate decay heat, a 3 x 3 BWR water channel, and one interchangeable rod position for tests such as a simulated failed fuel rod, BWR water rod, and PWR guide tube tests. The industry standard procedure is used in drying the assembly which involves stepping the system pressure down and holding for certain periods until reaching 3 torr and holding for 30 minutes. This is believed to sufficiently dry the assembly. The goal is to record data throughout these tests that can provide the industry with a greater understanding of how well the process works. This data will also be used to develop and validate a computer model of the drying process. Mechanistic models are currently under development. Reduced order models are being investigated for incorporation into current related engineering codes such as COBRA-SFS. COBRA-SFS is a member of the COBRA family of codes which is mainly used in the evaluation of thermal hydraulic systems containing spent nuclear fuel. However, it does not currently have the capability to account for processes such as evaporation encountered during drying operations. Therefore, those physical models are to be added to the source code. At present a model of the experiment has been created using the COBRA-SFS Cycle-4 code. A number of steady state models along with transient models have been developed and presented. Each of these yield valuable results for comparison with the experiments conducted. In addition to comparison with experimental results, the data collected from COBRA-SFS is compared with that which has been generated from other computer models which utilize computational fluid dynamics. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Sorption Behavior of Eu(III) into CSH Gel in Imitated Saline Groundwater - 12145

The sorption behavior of Eu(III) (europium (III)) into CSH (Calcium Silicate Hydrate) gel without dried processes was examined in imitated saline groundwater by using the spectro-fluorometer, Raman spectrophotometer and ICP-AES (Inductively Coupled Plasma- Atomic Emission Spectrometry). Ca/Si ratio was set to 0.4, 0.8, 1.2 and 1.6, and NaCl concentration was also set to 0.6, 0.06 and 0.006. The synthesis of each sample was conducted in a glove box saturated with nitrogen gas. The sealed sample tubes were gently shaken with 120 strokes/min. The time-period to contact Eu(III) with the CSH gel was set to 60 days. The fluorescence emission spectra suggested the incorporation of Eu{sup 3+} into CSH gel in the high Ca/Si ratio samples. On the other hand, from the decay behavior of fluorescence emission spectra, even in the low Ca/Si ratio samples, sorption behavior of Eu{sup 3+} into CSH gel was confirmed. Besides, the Raman spectra showed that the degree of polymerization of Si-O in CSH gel was raised with increasing Na ions concentration. These results suggest that the CSH gel, formed as secondary mineral, would retard the migration of radionuclides even in saline groundwater. Considering the inflow of saline groundwater into repository, this study examined the interaction between CSH gel (without dry processes) and Eu{sup 3+} by using the fluorescence emission spectra, the decay behavior of fluorescence and the Raman spectra. As a result, the fluorescence emission spectra of the sample of more than 0.8 Ca/Si ratio confirmed the intensity split into two peaks around 618 nm (5D0→7F2 transition). Furthermore, even in relatively low Ca/Si ratio samples, the fluorescence lifetimes both of the surface sorption sample and the co-precipitated samples exceeded that of the filtrate sample. These suggested that Eu{sup 3+} is not only hydrolyzed to form Eu(OH){sub 3} colloid, but is also stably incorporated into CSH gel (in Ca/Si ratio>1.2) or is forming complex on the surface of solid phase (in Ca/Si<0.8) in the co-presence of Na ions. On the other hand, the concentrations of Na, Ca and Si after 60 days in the solution filtrated through 0.2 μm membrane filter showed the ion exchange of Ca and Eu apparently in the samples of low Ca/Si ratio. Besides, in the samples with high Ca/Si ratio, Ca and Si concentrations in the solution were raised with increment of initial Na concentration. While Na ions may slightly increase the solubility of CSH gel, the CSH samples were stable, mostly maintaining the initial Ca/Si ratio synthesized. Moreover, the Raman spectra showed that the degree of polymerization of silicate chain in CSH samples increased with increasing Na concentration. Such an immobilized Na into the structure of CSH gel did not obstruct the incorporation of Eu{sup 3+} into CSH gel. The results mentioned above suggested that CSH gel (formed as a secondary mineral around the repository) also can retard the migration of radionuclides even if the repository and its surrounding are saturated by saline groundwater. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

ANSTO Synroc Wasteforms for the Immobilization of a Range of Mo-99 Production Wastes - 19326

Tc-99m, the most commonly used radioisotope in nuclear medicine, is produced from the decay of Mo-99 which is predominantly produced in reactors by irradiation of enriched uranium targets. This results in three waste streams, one of which is a uranium filter cake containing fission products. This filter cake could itself be treated as a waste stream, or processed to retrieve the enriched uranium for further use, thus producing additional waste streams during reprocessing. All wastes are considered intermediate-level wastes and require an appropriate wasteform design for termination. ANSTO is collaborating with NECSA (Nuclear Energy Corporation of South Africa) to develop wasteform and processing solutions for the treatment of intermediate level wastes (ILW) from the production of Mo-99. The work is supported by the U.S. Department of Energy's (DOE) National Nuclear Security Administration (NNSA) with technical support from Argonne National Laboratory (ANL). Here we highlight some of the outcomes of this collaboration which has seen the investigation and performance testing of a range of waste treatment technologies. Among the treatment options for the various waste streams, including the enriched uranium filter cake itself, were glass, glass-ceramic, ceramic and geopolymer wasteforms. All wasteforms were characterized (e.g., by X-ray diffraction (XRD) and scanning electron microscopy (SEM)) and their performance characteristics, for example compressive strength and aqueous durability, determined by standard methods and compared. Consideration has been given to the most suitable, practical and cost effective solutions for the immobilization of these Mo-99 wastes. This paper will discuss the down-selection of wasteforms and process technologies for Mo-99 wastes and compare these with regards to wasteform performance and process technology readiness. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

The Solubility of 242 PuO 2 in the Presence of Aqueous Fe(II): The Impact of Precipitate Preparation

The solubility of different forms of precipitated 242 PuO 2 (am) were examined in solutions containing aqueous Fe(II) over a range of pH values. The first series of 242 PuO 2 (am) suspensions were prepared from a 242 Pu(IV) stock that had been treated with thenoyltrifluoroacetone (TTA) to remove the 241 Am originating from the decay of 241 Pu. These 242 PuO 2 (am) suspensions showed much higher solubilities at the same pH value and Fe(II) concentration than previous studies using 239 PuO 2 (am). X ray absorption fine structure (XAFS) spectroscopy of the precipitates showed a substantially reduced Pu-Pu backscatter over that previously observed in 242 PuO 2 (am) precipitates, indicating that the 242 PuO2(am) precipitates purified using TTA lacked the long range order previously found in 239 PuO 2 (am) precipitates. The Pu(IV) stock solution was subsequently repurified using an ion exchange resin and an additional series of 242 PuO 2 (am) precipitates prepared. These suspensions showed higher redox potentials and total aqueous Pu concentrations than the TTA purified stock solution. The higher redox potential and aqueous Pu concentrations were in general agreement with previous studies on 242 PuO 2 (am) precipitates, presumably due to the removal of possible organic compounds originally present in the TTA purified stock. 242 PuO 2 (am) suspensions prepared with both stock solutions showed almost identical solubilities in Fe(II) containing solutions even though the initial aqueous Pu concentrations before the addition of Fe(II) were orders of magnitude different. By examining the solubility of 242 PuO 2 (am) prepared from both stocks in this way we have essentially approached equilibrium from both the undersaturated and oversaturated conditions. The final aqueous Pu concentrations are predictable using a chemical equilibrium model which includes the formation of a nanometer sized Fe(III) reaction product, identified in the 242 PuO 2 (am) suspension both by use of 57 Fe Mössbauer spectroscopy and transmission electron microscopy (TEM) analysis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1961

A Study of the Free Oscillations of the Earth

Published observations on the toroidal oscillations of the earth are critically reviewed. A supplementary analysis of the record obtained by the Lamont strain seismometer is presented. Eleven toroidal modes are identified, and it is concluded that the periods are known to within 1 per cent. A perturbation scheme involving the ratio of the angular velocity of the earth to the resonant frequency is used in calculating the effects due to the rotation of the earth on the resonant frequency. The free oscillations are viewed as a superposition of traveling waves. In a nonrotating system two traveling waves combine to produce a stationary standing wave. In a rotating system, the rotation distinguishes between waves that travel in the direction of rotation and those that travel in the opposite direction. Rotation removes a degeneracy and results in a splitting of a spectral peak of order l into 2 times l plus 1 peaks. The fractional displacement in frequency for the lowest-order toroidal oscillations is 1/206 and of the same order as the Q of the peak, so that splitting will probably not be observed in the toroidal oscillations. Viewed locally, rotation causes a particle to precess about a direction parallel to the axis of rotation. This precession will cause a variation of amplitude with time if the motion is recorded by an instrument with an anisotropic response function. Care is therefore needed in studying the time decay of a given spectral peak. Rotation also couples the normal coordinates so that a motion that is initially purely horizontal will develop a vertical component. It is expected that vertical seismometers should record particle motion with the toroidal frequencies. The perturbations of the toroidal oscillations due to core-mantle interaction are treated in detail. An exact expression is obtained for the rate of energy dissipated by a finitely conducting plate oscillating across a magnetic field. The energy dissipated at the core-mantle boundary due to viscous and hydromagnetic coupling is shown to be insignificant as compared with the energy dissipated within the mantle. The toroidal magnetic field leaking into the lower mantle combines with the dipole field, resulting in a stress on the mantle, tending to stiffen the lower boundary. The stress is of sufficient magnitude to produce a displacement toward higher frequency in the lower-order toroidal oscillations. Observations on the (sub 0) T (sub 2) oscillations lead to an estimate of the toroidal magnetic field in the lower mantle. A calculation of elastic energy in the low-order oscillations suggests a value of 10 (sup 18) ergs per cycles per hour for the energy density at low frequencies in the Chilean earthquake. Each mode of oscillation has a characteristic radial distribution of elastic energy associated with it. This distribution determines which parts of the earth contribute most heavily in determining a particular resonant frequency. The distribution of energy for the lower 17 modes for a homogeneous and a Gutenberg model earth is calculated. The resonant frequencies for models of the earth based on the Gutenberg and Lehmann distribution of elastic properties are presented. It is shown that the Gutenberg model earth fits the observations more closely than the Lehmann model and that a slight alteration of the Gutenberg model gives a significantly better fit to the observations. The alteration involves a lower shear-wave velocity in the lower mantle while the Gutenberg velocity distribution is maintained in the upper mantle. Various studies of the earth's oscillations coupled with surface-wave investigations substantiate Gutenberg's hypothesis of a layer of low velocity in the upper mantle. The physical conditions required for the formation of a region of low velocity are examined in detail. The results confirm Birch's earlier statement that a temperature gradient in excess of 6 degrees to 7 degrees per kilometer is needed to produce a decrease in velocity. The low-velocity layer does not require that the temperature approach or exceed the melting temperature. If tile upper mantle is homogeneous, the region of lower velocity should commence at the base of the crust and extend to 150 kilometers under the oceans and about 100 kilometers under continental regions. The distribution of thermal conductivity and radioactivity consistent with the low-velocity layer is also considered.

MacDonald, Gordon J. F.↗

DOE OSTI2025

Ultra-sensitive radon assay using an electrostatic chamber in a recirculating system

Rare event searches such as neutrinoless double beta decay and Weakly Interacting Massive Particle detection require ultra-low background detectors. Radon contamination is a significant challenge for these experiments, which employ highly sensitive radon assay techniques to identify and select low-emission materials. This work presents the development of ultra-sensitive electrostatic chamber (ESC) instruments designed to measure radon emanation in a recirculating gas loop, for future lower background experiments. Unlike traditional methods that separate emanation and detection steps, this system allows continuous radon transport and detection. This is made possible with a custom-built recirculation pump. A Python-based analysis framework, PyDAn, was developed to process and fit time-dependent radon decay data. Radon emanation rates are given for various materials measured with this instrument. A radon source of known activity provides an absolute calibration, enabling statistically-limited minimal detectable activities of 20 µBq. These devices are powerful tools for screening materials in the development of low-background particle physics experiments.

47 OTHER INSTRUMENTATION↗

DOE OSTI2012

An Investigation into the Transportation of Irradiated Uranium/Aluminum Targets from a Foreign Nuclear Reactor to the Chalk River Laboratories Site in Ontario, Canada - 12249

This investigation required the selection of a suitable cask and development of a device to hold and transport irradiated targets from a foreign nuclear reactor to the Chalk River Laboratories in Ontario, Canada. The main challenge was to design and validate a target holder to protect the irradiated HEU-Al target pencils during transit. Each of the targets was estimated to have an initial decay heat of 118 W prior to transit. As the targets have little thermal mass the potential for high temperature damage and possibly melting was high. Thus, the primary design objective was to conceive a target holder to dissipate heat from the targets. Other design requirements included securing the targets during transportation and providing a simple means to load and unload the targets while submerged five metres under water. A unique target holder (patent pending) was designed and manufactured together with special purpose experimental apparatus including a representative cask. Aluminum dummy targets were fabricated to accept cartridge heaters, to simulate decay heat. Thermocouples were used to measure the temperature of the test targets and selected areas within the target holder and test cask. After obtaining test results, calculations were performed to compensate for differences between experimental and real life conditions. Taking compensation into consideration the maximum target temperature reached was 231 deg. C which was below the designated maximum of 250 deg. C. The design of the aluminum target holder also allowed generous clearance to insert and unload the targets. This clearance was designed to close up as the target holder is placed into the cavity of the transport cask. Springs served to retain and restrain the targets from movement during transportation as well as to facilitate conductive heat transfer. The target holder met the design requirements and as such provided data supporting the feasibility of transporting targets over a relatively long period of time. A suitable transport cask was selected and a device for housing irradiated targets for loading, unloading and transportation has been designed, built and validated. The device was successful in meeting all design requirements for this feasibility study. Experiments were conducted with a custom test facility to confirm that the design met the maximum temperature requirements during shipping. Results from tests showed that the peak temperature in the apparatus was 300 deg. C. By compensating for experimental considerations, such as reduced thermal conductivity of the test cask versus that of the actual cask the expected maximum target temperature reduces to 231 deg. C. This is below the designated peak value of 250 deg. C. It can therefore be concluded, based on the content of this paper and from a heat-removal standpoint, the feasibility of transporting targets from a foreign nuclear reactor to Canada is possible, although further testing with irradiated targets and a full size cask would be a recommended next step. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Invited Article: Radon and thoron intercomparison experiments for integrated monitors at NIRS, Japan

Inhalation of radon ({sup 222}Rn) and its short-lived decay products and of products of the thoron ({sup 220}Rn) series accounts for more than half of the effective dose from natural radiation sources. At this time, many countries have begun large-scale radon and thoron surveys and many different measurement methods and instruments are used in these studies. Consequently, it is necessary to improve and standardize technical methods of measurements and to verify quality assurance by intercomparisons between laboratories. Four international intercomparisons for passive integrating radon and thoron monitors were conducted at the NIRS (National Institute of Radiological Sciences, Japan). Radon exercises were carried out in the 24.4 m{sup 3} inner volume walk-in radon chamber that has systems to control radon concentration, temperature, and humidity. Moreover, the NIRS thoron chamber with a 150 dm{sup 3} inner volume was utilized to provide three thoron intercomparisons. At present, the NIRS is the only laboratory world-wide that has carried out periodic thoron intercomparison of passive monitors. Fifty laboratories from 26 countries participated in the radon intercomparison, using six types of detectors (charcoal, CR-39, LR 115, polycarbonate film, electret plate, and silicon photodiode). Eighteen laboratories from 12 countries participated in the thoron intercomparisons, using two etch-track types (CR-39 and polycarbonate) detectors. The tests were made under one to three different exposures to radon and thoron. The data presented in this paper indicated that the performance quality of laboratories for radon measurement has been gradually increasing. Results of thoron exercises showed that the quality for thoron measurements still needs further development and additional studies are needed to improve its measuring methods. The present paper provides a summary of all radon and thoron international intercomparisons done at NIRS from 2007 to date and it describes the present status on radon and thoron passive, one-time cycle monitors.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2014

Radionuclide Inventory and Distribution: Fourmile Branch, Pen Branch, and Steel Creek IOUs

As a condition to the Department of Energy (DOE) Low Level Waste Disposal Federal Facility Review Group (LFRG) review team approving the Savannah River Site (SRS) Composite Analysis (CA), SRS agreed to follow up on a secondary issue, which consisted of the consolidation of several observations that the team concluded, when evaluated collectively, could potentially impact the integration of the CA results. This report addresses secondary issue observations 4 and 21, which identify the need to improve the CA sensitivity and uncertainty analysis specifically by improving the CA inventory and the estimate of its uncertainty. The purpose of the work described herein was to be responsive to these secondary issue observations by re-examining the radionuclide inventories of the Integrator Operable Units (IOUs), as documented in ERD 2001 and Hiergesell, et. al. 2008. The LFRG concern has been partially addressed already for the Lower Three Runs (LTR) IOU (Hiergesell and Phifer, 2012). The work described in this investigation is a continuation of the effort to address the LFRG concerns by re-examining the radionuclide inventories associated with Fourmile Branch (FMB) IOU, Pen Branch (PB) IOU and Steel Creek (SC) IOU. The overall approach to computing radionuclide inventories for each of the IOUs involved the following components: Defining contaminated reaches of sediments along the IOU waterways, Identifying separate segments within each IOU waterway to evaluate individually, Computing the volume and mass of contaminated soil associated with each segment, or “compartment”, Obtaining the available and appropriate Sediment and Sediment/Soil analytical results associated with each IOU, Standardizing all radionuclide activity by decay-correcting all sample analytical results from sample date to the current point in time, Computing representative concentrations for all radionuclides associated with each compartment in each of the IOUs, Computing the radionuclide inventory of each DOE-added radionuclide for the compartments of each IOU by applying the representative, central value concentration to the mass of contaminated soil, Totaling the inventory for all compartments associated with each of the IOUs Using this approach the 2013 radionuclide inventories for each sub-compartment associated with each of the three IOUs were computed, by radionuclide. The inventories from all IOU compartments were then rolled-up into a total inventory for each IOU. To put the computed estimate of radionuclide activities within FMB, PB, and SC IOUs into context, attention was drawn to Cs-137, which was the radionuclide with the largest contributor to the calculated dose to a member of the public at the perimeter of SRS within the 2010 SRS CA (SRNL 2010). The total Cs-137 activity in each of the IOUs was calculated to be 9.13, 1.5, and 17.4 Ci for FMB, PB, and SC IOUs, respectively. Another objective of this investigation was to address the degree of uncertainty associated with the estimated residual radionuclide activity that is calculated for the FMB, PB, and SC IOUs. Two primary contributing factors to overall uncertainty of inventory estimates were identified and evaluated. The first related to the computation of the mass of contaminated material in a particular IOU compartment and the second to the uncertainty associated with analytical counting errors. The error ranges for the mass of contaminated material in each IOU compartment were all calculated to be approximately +/- 9.6%, or a nominal +/-10%. This nominal value was added to the uncertainty associated with the analytical counting errors that were associated with each radionuclide, individually. This total uncertainty was then used to calculate a maximum and minimum estimated radionuclide inventories for each IOU.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Applying fast calorimetry analysis on spent fuel calorimeter - 15620

Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Chloride-Induced Stress Corrosion Cracking Characteristics of the Used Nuclear Fuel Canister Material

The used nuclear fuel (UNF) dry storage systems are located in the coastal and lake/river-side regions, the service time of which is 60 years and more. One type of dry storage systems is composite of a thin-walled austenitic stainless steel (SS) welded canister and a concrete overpack. The air flows through the canister surface and brings away the decay heat. Meanwhile, the air brings the salt species into the canister and make them deposit on the surface. The salt deposit will be deliquescent when absorbing the moisture air, being corrosive to the surface. NRC and other programs claimed that chloride induced stress corrosion cracking (CISCC) would lead to the canister failure, being one of the major technical gaps for extended dry storage systems. CISCC is induced by the combined influence of the aggressive environment (chloride ions in our case) and tensile stresses (residual stresses in our case) on the susceptible materials (the austenitic SS in our case). Chrome (Cr), as one of the chemical compositions of SSs, plays an important role in the corrosion resistance. It can form a chrome-rich inner barrier oxide layer to prevent the SS from corrosion. However, chloride ions can easily break down the barrier and accelerate dissolution of the underlying metal, the phenomena of which is pitting corrosion. Pitting is the precursor to CISCC as which provides the combination of a local aggressive solution chemistry and a stress concentrating feature. Accordingly, CISCC invariably originates from the pits, and cracks that have nucleated on pit walls can grow around the pit and coalesce to form a complete through-crack. Significant tensile residual stresses, caused by the welding process, exist throughout the thickness of the canister wall. Enos and Bryan found that large residual stresses were in the full-scale mockup container simulating a canister, which supports the SCC propagation. It is therefore interesting to achieve a good knowledge about the effect of welding on the durability of SSs, especially in the corrosive environment. In order to understand the fundamental of the cracking behaviors, it's vital to elucidate the microstructure and microchemistry of cracking. This study focused on observing the CISCC characteristics of the UNF canister material exposed to the humid environment with salt deposit. Extensive micro-characteristics have been achieved by the techniques include focused ion beam/ scanning electron microscope (FIB/SEM), transmission electron microscopy (TEM), energy dispersive spectroscopy (EDS), and optical profiler. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2026

Persistence of collectivity in the low-lying states of 30,31 Na inside the 𝑁 = 20 island of inversion

Near the 𝛽-stability line, nuclei with magic numbers are expected to exhibit spherical ground states. However, mass measurements of neutron-rich nuclei near the 𝑁 = 20 island of inversion have indicated an excess in the binding energies. These results, together with subsequent spectroscopic studies and theoretical works, point to the structural evolution from normal configurations to intruder-deformed configurations characterized by particle-hole excitations across the shell gap. Despite evidence of intruder-dominant ground states for the 30,31 Na isotopes in the vicinity of 𝑁 = 20, a question remains as to whether the deformation persists into the ground-state bands beyond the first excited states. Here, in this work, we report on a heavy-ion inelastic scattering measurement performed with GRETINA, the TRIPLEX device, and the S800 spectrograph to study the low-lying excitations in neutron-rich 30,31 Na . The observed gamma rays for the (4 + ) → (3 + ), (3 + ) → 2$^+_{g.s.}$, and (4 + ) → 2$^+_{g.s.}$ decays in 30 Na and for the (7/2 + ) → (5/2 + ) and (5/2 + )→3/2$^+_{g.s.}$ decays in 31 Na were used to extract the reduced transition probabilities 𝐵⁡(𝐸⁢2↑). We report the first measurement of the 𝐵⁡(𝐸⁢2↑) between the 3/2$^+_{g.s.}$ and the (7/2 + ) states in 31 Na . The results from this study are compared to shell-model calculations, confirming the persistence of large collectivity in the low-lying excited states consistent with the formation of the well-deformed ground-state bands in 30,31 Na .

collective models↗

DOE OSTI2021

SR19021 Tritium Aging of Regenerated LANA.75 (Final Project Report)

The Savannah River Tritium Enterprise (SRTE) has used the metal hydride LaNi 4.25 Al 0.75 (LANA.75) in the Tritium Facilities for over two decades. LANA.75 beds store significant quantities of tritium but have a limited service life due to the radiolytic decay of tritium to He-3 within the metal matrix. It has been shown that heating tritium-aged LANA.75 under vacuum can reverse tritium aging effects, eliminating the heel of trapped hydrogen, and restoring the reversible capacity. Additional investigation is needed to ensure there are no unexpected changes to the hydride before this restoration technique is employed in full scale beds in the Tritium Facilities. This project was to be comprised of three distinct scopes: obtain tritium aging data on the sample regenerated in 2018, regenerate a tritium-aged sample at 600 °C, and perform thermal stability testing on a non-tritiated sample. Isotherms were collected on the previously regenerated sample after approximately two years of tritium aging. Isotherms were collected at 80, 100, and 120 °C. As expected, there was a decrease in the plateau pressure, an increase in plateau slope, and a portion of the “heel” of tritium trapped in the metal had been reestablished. Unexpectedly, it appeared that the plateau had shortened at the higher tritium to metal ratios as well. This is typically seen in older samples. The second scope, to regenerate a second tritium-aged LANA.75 sample, was not completed. A Task Technical and Quality Assurance Plan was written and approved, a high temperature test cell was fabricated, several pre-job briefs were held, and the hydride sample was passivated with air. Despite these successes, the hydride sample was not recovered from the legacy test cell. The third scope was to perform thermal stability testing on a “cold” LANA.75 sample. A non-tritiated sample of LANA.75 was held at 750 °C under vacuum for 200 hours to simulate exposure to multiple regeneration evolutions. Hydride isotherm performance, chemical composition, crystallinity, particle size, and morphology are compared between the pre- and post-regeneration samples. No significant changes were observed in composition, crystallinity, or particle size. Comparison of before and after isotherms showed that performance improved rather than deteriorated during the evolution. Scanning Electron Microscopy (SEM) analysis showed small growths on the particle surface after exposure to regeneration conditions. Additional testing will be required to determine the cause of these growths.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2024

Data-driven background model for the CUORE experiment

Here, we present the model we developed to reconstruct the CUORE radioactive background based on the analysis of an experimental exposure of 1038.4 kg yr. The data reconstruction relies on a simultaneous Bayesian fit applied to energy spectra over a broad energy range. The high granularity of the CUORE detector, together with the large exposure and extended stable operations, allow for an in-depth exploration of both spatial and time dependence of backgrounds. We achieve high sensitivity to both bulk and surface activities of the materials of the setup, detecting levels as low as 10 nBq kg −1 and 0.1 nBq cm −2 , respectively. We compare the contamination levels we extract from the background model with prior radio-assay data, which informs future background risk mitigation strategies. The results of this background model play a crucial role in constructing the background budget for the CUPID experiment as it will exploit the same CUORE infrastructure.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗