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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Fracture Toughness Evaluation for Spent Nuclear Fuel Clad Systems Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials. These materials are often held in long-term storage as an interim stage within their lifecycle, including reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations and follow-on SNF transportations to its final repository sites are very important, due to a great concentration of fission products, actinides and activation products. Fracture mechanics approach in applying to SNF system reliability investigation, especially for the high burn up (HBU) SNF, during SNF long term dry storage or SNF transportation is necessary due to the inherited flaws and inhomogeneity structures existed in a SNF system from the reactor operation. For instance, such as hydride and oxide formation, surface flaw induced by SNF assembly contact interactions, or internal flaws induced by pellet-clad mechanical interaction (PCMI). However, none of the existing fracture toughness data deal with fuel cladding specific geometry or SNF material conditions, such as cladding structure with the segment fuel pellets configuration and the PCMI mechanism. Consequently, the application of existing fracture toughness data to SNF system reliability investigation can be error prone; thus, the development of an industry-wide consensus fracture mechanics approach is needed. The objective of this research is to develop an in-situ fracture mechanics testing protocol, including the associated analytical procedure, that is suitable for evaluating the SNF fracture toughness. Fracture toughness data was obtained under quasi-static fracture loading using Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support SNF reliability investigation during long-tern SNF dry storage stage or in the follow-on SNF system transport to the final repository sites. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre cracks are needed for brittle materials specimens. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient pre-crack length to increase the sample localized constraint at crack front. Fatigue pre-crack growth techniques with the associated compliance function estimated was developed for estimating the crack penetration depth to monitor the fatigue crack growth evolution. In order to extend SNTT approach to a thin shell cladding structure material, a new testing protocol and the associated analytical procedure for estimating the SNF fracture toughness was developed. The detailed SNTT approach and its estimated fracture toughness for the Zr-4 cladding structure with segment alumina inserts are presented in this report. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to Zr-4 cladding material. For the medium and the short crack length, the estimate J Q upon fracture for the baseline Zr-4 cladding is at 285.7 lb/in with 2-sigma uncertainty of 18.59 lb/in., and the associated K Q is at 61.4 Ksi√in. For a long crack length, the crack initialization orientation is apparently deviated from the principle tensile stress contour – 45° spiral crack front (i.e., Mode I + Mode III failure mechanism); and the estimated J Q is at 108 lb/in., the associated K Q is at 37.7 Ksi√in. The estimate Mode I J Q (along the 45° pitch principle tensile stress contour) is at 200 lb/in.; which indicate a significant reduction in the fracture toughness in a tubing structure under mixed mode loading condition, Mode I + Mode III, compared to that of Mode I (tensile stress dominated) alone.

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

H-Canyon Jumper Gasket Sealing Evaluation

H-Canyon Documented Safety Analysis approved December 2019 credits the leak integrity of H-Canyon (HCA) process tanks, piping, and jumpers in preventing the release of radioactive material in a post seismic environment. Demonstration of leak tightness capability for this equipment was achieved through substantial analytical effort. However, one potential weak point that affects leak integrity exists at the gasketed joint between Hanford Connectors (HC) and nozzles. Review of the HCA Crane Log over a three-year period indicates the HCs periodically leak. Savannah River National Laboratory (SRNL) was requested to perform gasket compression tests that evaluate gasket relaxation characteristics, i.e., creep, and provide improvements to the HC installation process that can reduce or eliminate the frequency of leaks at HCs. A HC torque value of 312 ft-lbf is the target value used because it is the minimum torque value, with 95% confidence, when an unlubricated HC is impacted for standard sealing period of 9 seconds. In the HCA pipe-jumper connections are installed remotely by using HCs. HCs inherently add to the uncertainty of a gasketed joint because a calibrated torque value cannot be applied. The gasket material is a special material made of asbestos-fiber impregnated with Teflon. This material has been used for decades and generally performs well but does leak intermittently as gaskets creep and preloaded torque is lost. Gasket types are either snap-ring (SR) or lampshade (LS). Both gasket types cover the same seating area. The SR gasket is designed to remain with the jumper when the jumper is removed leaving the nozzle faces clean of any debris. SR gaskets are initially installed when the jumper is fabricated. Changing gaskets is time consuming and the radioactive environment presents risks to personnel. When jumpers are removed LS gaskets could be used to better seal the connection to reduce down time and ALARA concerns. That is, an LS gasket would be remotely positioned over the nozzle and the HC reinstalled. Testing also included two situations not typical of plant operation. One situation was the use of the non-standard double-layered gaskets (DG), and the other situation was the use of a LS without removal of the installed SR gasket. The DG was used when the supply of LS gaskets ended, so that testing could continue. It was of interest to demonstrate this gasket’s creep response because the DG, which was two gaskets glued together, was thicker than the LS gasket and the thicker material was expected to elicit a greater creep characteristics. The second situation of using two gaskets together by installing an LS gasket without removing the installed gasket is not allowed by the ASME Codes. However, by demonstrating the creep history of such a gasket arrangement it could be considered as an option that could possibly seal a connection while significantly reducing operational down time and personnel risk. Forty tests were conducted, and the overall results demonstrate the initial applied force and how the gasket changes with time. The change in sealing force, in terms of torque, was then measured to obtain a force history as the gaskets creep. This report documents the results of those tests as well as the method used to perform the tests.

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

Update of Receptor Locations for INL National Emissions Standard for Hazardous Air Pollutants (NESHAP) Assessments

This report documents the selection of updated public receptor locations for use in conducting regulatory assessments of radioactive air emissions for Idaho National Laboratory (INL) facilities. The basis for this receptor location analysis is found in 40 CFR 61 Subpart H, “National Emissions Standards for Emissions of Radionuclide Other than Radon from Department of Energy Facilities,” where the requirements for identifying National Emission Standards for Hazardous Air Pollutants (NESHAP) receptor locations are outlined. Receptor locations (residences, schools, businesses, or farm operations) were identified using high-resolution satellite imagery. This assessment identified 31 unique receptor locations for all INL Site facilities, and 30 receptor locations for INL in-town facilities that have the potential to emit radioactive materials. Identification of receptors for each in-town facility is new to this report.

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

Chemistry Envelope for Pitting and Stress Corrosion Cracking Mitigation

The present waste chemistry controls for the Hanford double-shell tanks (DSTs) were established in the 1980s in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits were coupled with corrosion testing results from Pacific Northwest Laboratory (PNL) and adopted for corrosion control of the DST waste at the Hanford Site. Since that time, the temperatures of many DST wastes have decreased to less than 50 °C as the radioactive material has decayed, the testing program in Argentina demonstrated that chloride ion in Hanford Site waste could cause pitting, and other work showed that dilute solutions could cause pitting at the liquid air interface (LAI). Moreover, future transfers from the single shell tanks and the waste treatment plant and process changes may shift the DST waste chemistries to a broader range of compositions with higher aggressive anion concentrations than the wastes that are presently stored. The decreasing temperatures led to the formulation of new chemistry controls for the minimization of SCC risks for operations below 50 °C in 2010, and these controls were partially adopted. The Tank Integrity Expert Panel (TIEP) Corrosion Subgroup (CSG) recommended that additional testing be performed to assess the potential for pitting and LAI corrosion. Since 2015, DNV-GL, SRNL, and the Hanford Site laboratory have investigated pitting and LAI corrosion at current and anticipated DST chemistries and Savannah River National Laboratory (SRNL) undertook a statistically based investigation of the role of nitrate and halide ion induced pitting corrosion. The objective was to develop a comprehensive waste chemistry envelope for the simultaneous minimization of the pitting and SCC risks caused by halide and nitrate ions. This report summarizes the technical basis for the new waste chemistry envelope to minimize the potential for both pitting and cracking corrosion induced by nitrate and halide ions, and the strategies used for the validation of the new controls.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Data Analysis and Detector Troubleshooting for the Silicon Muon Scanner

Muon tomography is the calculation of the density distribution of a material from the probable angles of muon Coulomb scattering in a sampled object. This allows the probing of internal object structures that cannot otherwise be observed. It has many important applications for both national security and other fields, such as archaeology and geology. Its uses range from detecting radioactive material intended for nuclear weaponry to searching for hidden chambers in the Egyptian pyramids. In this project we intend to demonstrate the capability of a tomographic system which uses cosmic rays as the muon source and tracks each muon's initial and final trajectories to ascertain information about their scattering pattern. This method can be used to determine the structure or composition of an object, such as testing to see if a nuclear fuel cask is empty or full. We intend this as an early R&D exploration into the practicality of implementing silicon microstrip sensors in border scans for radioactive contraband, as an alternative to existing drift tube technology.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Developments that are making the Versa-Pac Indeed More Versatile (and More Useful to Industry) - 16267

Daher-TLI acquired the assets and employees of Century Industries, a radioactive material packaging manufacturing facility, in 2013. Included in that acquisition was the Versa-Pac Shipping Package. The Versa-Pac is an NRC licensed Type AF packaging (Certificate of Compliance no. USA/9342/AF-96) and was originally designed as a replacement for standard 7A drum-type transportation packaging. It can be used directly or in conjunction with pails, drums or inserts that can be manufactured to accommodate a variety of smaller containers and vessels. The Versa-Pac comes in two sizes, the VP-55 (i.e., 55-gallon drum version) and VP-110 (i.e., 110-gallon drum version). It features a patented design concept in combination with the drum exterior to provide enhanced structural protection to payloads under Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC), per 10CFR71.71 and 10CFR71.73 respectively. Since its introduction in 2010, the Versa-Pac has been utilized mainly as a one-time use package for disposing of radioactive waste. The package is also designed as a reusable package for transporting front-end materials. Discussions with current and prospective clients have resulted in numerous improvements, accomplished and planned, to the package's payload capability through additional engineering analyses and actual design modifications. Revision 10 to the NRC Certificate of Compliance, dated August 6, 2015, introduces an enrichment-loading table that allows shipping larger quantities of U-235 for enrichments less than 100 wt.%. It also authorizes shipment of natural thorium. Other planned and proposed improvements include provisions to transport UF{sub 6} in 1S, 2S, and 5B cylinders, design changes that further increases the amount of U-235 at enrichments less than 100 wt.%, increasing the capacity of the VP-110, developing a lighter 30-gallon version, the VP-30, and introducing content limits specific to uranium-oxides in multiple forms. This paper traces the engineering and design efforts involved in expanding the Versa-Pac's versatility. (authors)

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

Neptunium transition probabilities estimated through laser induced breakdown spectroscopy (LIBS) measurements

in tis report calibration free–laser induced breakdown spectroscopy (CF-LIBS) approaches are promising for the analysis of radioactive materials; however, a lack of reported fundamental properties for these materials is an obstacle to implementation. Spectra of laser induced plasmas containing Np and Sr, along with Saha–Boltzmann methods, were used to estimate the first reported transition probabilities (TPs) of Np. These TPs enabled the first attempt at CF-LIBS analysis of radioactive samples to predict Np/Sr ratios with a mean error of 3.86%. The presented methodology of TP estimation will be applied to the study of other actinides and allow broader applications of CF-15 LIBS in the nuclear field.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2019

ENHANCEMENTS IN THE TECHA RIVER DOSIMETRY SYSTEM: TRDS-2016D CODE FOR RECONSTRUCTION OF DETERMINISTIC ESTIMATES OF DOSE FROM ENVIRONMENTAL EXPOSURES

Waterborne releases to the Techa River from the Mayak plutonium facility in Russia during 1949–1956 resulted in significant doses to persons living downstream. The dosimetry system TRDS-2016D has been developed which provides individual doses of external and internal exposure for the members of the Techa River Cohort (TRC) and other persons, who were exposed to releases of radioactive material to the Southern Urals. The results of computation of individual doses absorbed in red bone marrow and extra-skeletal tissues for the TRC members (29,647 persons) are presented, which are based on residence histories on the contaminated Techa River and the East Urals Radioactive Trace (EURT) which was formed in 1957 as a result of the Kyshtym Accident. Available 90Sr-body-burden measurements and available information on individual household locations have been used for refinement of individual dose estimates. TRDS-2016D-based dose estimates will be used for verification of risk of low-dose-rate effects of ionizing radiation in the TRC.

dose reconstruction, environmental contamination, ↗
DOE OSTI2014

Pill to Treat Radiation Exposure

Rebecca Abergel explains how one pill could be used to treat people exposed to radioactive materials at our '8 Big Ideas' Science at the Theater event on October 8th, 2014, in Oakland, California.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Fiberboard Properties and Degradation Rates for Storage of 9975 Shipping Package in KAC (2020 Status Report)

Savannah River Site (SRS) stores plutonium materials within model 9975 shipping packages in the K-Area Complex (KAC). The 9975 shipping package consists of a 35 gallon stainless steel drum, Celotex fiberboard insulation, lead shield, and primary and secondary containment vessels. The 9975 shipping package design, performance, and analysis for safe transport of radioactive material are described in the Safety Analysis Report for Packaging (SARP). Celotex fiberboard provides three safety functions: thermal insulation to limit internal temperature during a fire, criticality control, and resistance to package crushing. The fiberboard material must retain its dimensions and density within certain ranges to provide the required impact resistance, criticality control, and fire resistance. The SRS Surveillance Program monitors material performance to establish a basis for service life and ensures the continued integrity of 9975 packages. Fiberboard samples, taken from multiple fiberboard assemblies fabricated from cane and softwood fiberboard, have been conditioning in elevated temperature environments since 2005. The samples are periodically examined to monitor thermal, mechanical, and physical properties, and assess degradation trends. Fiberboard properties of interest that are evaluated to demonstrate acceptable long-term performance include dimensional stability, density, compressive strength, thermal conductivity, and specific heat capacity. Duplicate samples from multiple package sources have been tested to identify the range of variability in fiberboard properties and degradation rates. Baseline and long-term testing of fiberboard material properties have been reported previously; reference 6 summarized experimental results of cane and softwood fiberboard through May 2019 and presented degradation models for the measured properties. This report presents the cumulative data collected through June 2020 and the corresponding updated aging models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A Review of Nuclear Forensics: 2016-2020

The study of nuclear forensics harkens back to the Manhattan Project-era, when scientists first started to analyze the debris from the 1945 Trinity test. Political turmoil stemming from the Cold War and the rehabilitation of Germany following WWII has led to new challenges in international security involving nuclear proliferation. Nuclear materials have, on occasion, been lost, misplaced, or stolen from former Soviet countries, and illicit materials have been interdicted all over the world. The National Technical Nuclear Forensics Center (NTNFC) was established in 2006, and has been at the forefront of drive to advance nuclear forensic capabilities in the United States. The ultimate goal of nuclear forensics is to examine nuclear and other radioactive materials using analytical techniques to determine origin and history of the material, particularly in the context of law enforcement investigations. Nuclear forensics can be divided into two parts: predetonation and post-detonation. Pre-detonation forensics, as the name implies, is the investigation of a nuclear material or weapon that has not been detonated or involved in an explosion, whereas postdetonation forensics is the study of activation or fission products in debris or the environment following the use of a nuclear or radiological dispersal device (RDD). Both parts require a number of analytical chemical and radiochemical techniques to determine identification of the material. Many advancements in analytical techniques, including rapidity, sample size, and forensic signatures have been made in recent years. The analytical methods that can be used in a nuclear forensic investigation, such as mass spectrometry and gamma spectroscopy, have been described in detail in previous reviews, including Straub et.al, and will not be explained here. This review will discuss recent publications (from 2016 to present) describing advancements of techniques such as radiochronometry, morphology, development of novel reference materials, and inter-laboratory collaborations for both pre- and post-detonation nuclear forensics.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

3rd Quarter Transportation Report FY 2014: Radioactive Waste Shipments to and from the Nevada National Security Site (NNSS)

This report satisfies the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) commitment to prepare a quarterly summary report of radioactive waste shipments to the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) at Area 5. There were no shipments sent for offsite treatment and returned to the NNSS this quarter. This report summarizes the 3rd quarter of Fiscal Year (FY) 2014 low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW) shipments. This report also includes annual summaries for FY 2014 in Tables 4 and 5. Tabular summaries are provided which include the following: Sources of and carriers for LLW and MLLW shipments to and from the NNSS; Number and external volume of LLW and MLLW shipments; Highway routes used by carriers; and Incident/accident data applicable to LLW and MLLW shipments. In this report shipments are accounted for upon arrival at the NNSS, while disposal volumes are accounted for upon waste burial. The disposal volumes presented in this report do not include minor volumes of non-radioactive materials that were approved for disposal. Volume reports showing cubic feet generated using the Low-Level Waste Information System may vary slightly due to differing rounding conventions.

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

Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium and Uranium Radioactivity in Fractions Separated from Placer Sands in Southeast Bangladesh

The present study reports the likely first attempt of separating radioactive minerals for estimation of activity concentration in the beach placer sands of Bangladesh. Several sand samples from heavy mineral deposits located at the south-eastern coastal belt of Bangladesh were processed to physically upgrade their radioactivity concentrations using plant and laboratory equipment. Following some modified flow procedure, individual fractions were separated and investigated using gamma-ray spectrometry and powder-XRD analysis. The radioactivity measurements indicated contributions of the thorium and uranium radioactive series and of {sup 40}K. The maximum values of {sup 232}Th and {sup 238}U, estimated from the radioactivity of {sup 208}Tl and {sup 234}Th in secular equilibrium, were found to be 152,000 and 63,300 Bq/kg, respectively. The fraction of the moderately conductive part in electric separation contained thorium predominantly, while that of the non-conductive part was found to be uranium rich. The present arrangement of the pilot plant cascade and the fine tuning of setting parameters were found to be effective and economic separation process of the radioactive minerals from placer sands in Bangladesh. Probable radiological impacts and extraction potentiality of such radioactive materials are also discussed.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

2nd Quarter Transportation Report FY 2014

This report satisfies the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) commitment to prepare a quarterly summary report of radioactive waste shipments to the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) at Area 5. There were no shipments sent for offsite treatment and returned to the NNSS this quarter. This report summarizes the second quarter of fiscal year (FY) 2014 low-level radioactive waste (LLW) and mixed low-level radioactive waste (MLLW) shipments. This report also includes annual summaries for FY 2014 in Tables 4 and 5. Tabular summaries are provided which include the following: Sources of and carriers for LLW and MLLW shipments to and from the NNSS; Number and external volume of LLW and MLLW shipments; Highway routes used by carriers; and Incident/accident data applicable to LLW and MLLW shipments. In this report shipments are accounted for upon arrival at the NNSS, while disposal volumes are accounted for upon waste burial. The disposal volumes presented in this report do not include minor volumes of non-radioactive materials that were approved for disposal. Volume reports showing cubic feet (ft3) generated using the Low-Level Waste Information System may vary slightly due to differing rounding conventions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Numerical Assessment of SNF Parameter Bandwidths and Muon Tomography Assay for Large Volume Casks - 18185

For the characterization of spent fuel and for the safe direct disposal of SNF we present: a) a method of numerical assessment of their radionuclides inventory that is derived from the initial fuel characteristics and individual burn-up / cooling time history of that particular fuel assembly, and b) how the numerical information can be experimentally assessed using muon tomography as a nondestructive assay of large and heterogeneous SNF casks. We combine both complementary methods, i.e. numerical assessment and muon tomography (MT) experimental assay, to provide a powerful tool for verifying the nuclear inventory without compromising the casks or radioactive material containers. Here, we shall present the numerical assessment technique and preliminary but very promising experimental results, and explain the scope of a versatile and mobile MT-detection system for the identification of nuclear material inside large casks containing spent fuel assemblies and/or other large volume heterogeneous material containers for which the European Commission has recently granted R and D funding within the CHANCE project to be developed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Chemo-chromic Pd-V{sub 2}O{sub 5} Sensors for Passive Hydrogen Detection in Nuclear Containments - 18268

The ability to detect and monitor hydrogen gas efficiently in process and storage facilities, handling nuclear material, is crucial to ensuring their safety. The accumulation of hydrogen gas, above the lower flammable limit (LFL), in a nuclear waste containment is a concern since it creates the potential for a hydrogen-air explosion to occur, which could lead to a loss of containment and result in the uncontrolled release of radioactive material into the surrounding environment. The events that took place at Fukushima Daiichi Nuclear Power Plant highlighted the vulnerability of conventional hydrogen detection to extreme events, where power may be lost. In the present work, chemo-chromic hydrogen sensors have been fabricated, using transition metal oxide thin films, to provide eye-readable detection systems that would be resilient to plant power failure. Vanadium oxide (V{sub 2}O{sub 5}) films were prepared on quartz glass substrates by sol-gel deposition and sensitized with a palladium (Pd) catalyst, deposited by electron beam evaporation. When exposed to hydrogen, the Pd catalyst dissociates H{sub 2} to H atoms, which diffuse into the V{sup (V)}{sub 2}O{sub 5} forming a hydrogen-vanadium metal bronze, H{sub 2}V{sup (III)}{sub 2}O{sub 5}, resulting in a noticeable colour change from orange to dark green. To assess their viability for nuclear safety applications, these sensors have been irradiated to total doses between 5 and 250 kGy using a Co-60 gamma isotope irradiator. The results suggest that gamma irradiation, at the levels examined, has an effect on the initial colour of the V{sub 2}O{sub 5} and Pd-V{sub 2}O{sub 5} thin films with decreased transmittance above 540 nm. The orange starting colour darkened and developed a green tone, with the degree of colour change depending on the applied total dose. Changes in surface morphology and characteristics have been examined by using Scanning Electron Microscopy (SEM) and Raman spectroscopy. High level (250 kGy) gamma radiation exposure begins to produce surface degradation on V{sub 2}O{sub 5} thin films; however this behaviour is not observed for films that are also coated with palladium. Chemo-chromic properties of both un-irradiated and irradiated Pd-V{sub 2}O{sub 5} thin films were determined by examining their optical transmittance, using UV-vis spectroscopy, under exposure to a 4% H{sub 2}-N{sub 2} gas mixture. Exposure to gamma radiation has been found to have negligible effect upon colour change behaviour after 30 minutes exposure to hydrogen gas. The results suggest that the thin film V{sub 2}O{sub 5} sensitised with Pd is a plausible technique for application in the monitoring of hydrogen gas in low-level gamma radiation environments. (authors)

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

Identification of Distorted Gamma-Ray Signature Patterns Using Digital Filtering and Auto-Associative Memory Implemented with a Hopfield Neural Network

The detection and identification of radioactive sources in search applications involve analyzing passive gamma-ray emissions from high-level radioactive materials. This process uses a mobile detector-spectrometer in a complex field test environment. Recently, the use of artificial intelligence for gamma-ray spectrum analysis has shown promising results. However, challenges persist in identifying isotopic signatures from spectral measurements that may be distorted due to source shielding, random variations in natural radioactive background, or insufficient measurement time to obtain clear spectral lines. Here, this paper presents a novel intelligent signature recognition method that combines digital filtering techniques with an artificial Hopfield Neural Network (HNN). The HNN leverages auto-associative memory to store training sample patterns and match them with incoming gamma spectra from distorted sources. It restores the testing sources’ measurements by finding the closest matching signature patterns in the spectral library. Before HNN recognition, the measured spectrum undergoes preprocessing with a digital image filter to reduce fluctuations. Performance of the proposed method is evaluated using a set of gamma-ray spectra measured with a sodium iodide detector. The data collected include measurements from six pure samples: 241 Am, 60 Co, 137 Cs, 192 Ir, 239 Pu, and 235 U, which are used for training and validation (i.e. six cases). Additionally, the data set contains 24 distorted synthesized sources with various fluctuating backgrounds. Test results demonstrate the potential of the proposed method to accurately recognize the correct isotope with high precision, achieving an accuracy rate exceeding 85%. Furthermore, the proposed method exhibits superior performance compared to the conventional multiple regression fitting and simple feedforward neural network methods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Enterprise SRS: Leveraging Ongoing Operations To Advance Nuclear Fuel Cycles Research And Development Programs

The Savannah River Site (SRS) is repurposing its vast array of assets to solve future national issues regarding environmental stewardship, national security, and clean energy. The vehicle for this transformation is Enterprise SRS which presents a new, radical view of SRS as a united endeavor for ''all things nuclear'' as opposed to a group of distinct and separate entities with individual missions and organizations. Key among the Enterprise SRS strategic initiatives is the integration of research into facilities in conjunction with on-going missions to provide researchers from other national laboratories, academic institutions, and commercial entities the opportunity to demonstrate their technologies in a relevant environment and scale prior to deployment. To manage that integration of research demonstrations into site facilities, The Department of Energy, Savannah River Operations Office, Savannah River Nuclear Solutions, the Savannah River National Laboratory (SRNL) have established a center for applied nuclear materials processing and engineering research (hereafter referred to as the Center). The key proposition of this initiative is to bridge the gap between promising transformational nuclear fuel cycle processing discoveries and large commercial-scale-technology deployment by leveraging SRS assets as facilities for those critical engineering-scale demonstrations necessary to assure the successful deployment of new technologies. The Center will coordinate the demonstration of R&D technologies and serve as the interface between the engineering-scale demonstration and the R&D programs, essentially providing cradle-to-grave support to the research team during the demonstration. While the initial focus of the Center will be on the effective use of SRS assets for these demonstrations, the Center also will work with research teams to identify opportunities to perform research demonstrations at other facilities. Unique to this approach is the fact that these SRS assets will continue to accomplish DOE's critical nuclear material missions (e.g., processing in H-Canyon and plutonium storage in K-Area). Thus, the demonstration can be accomplished by leveraging the incremental cost of performing demonstrations without needing to cover the full operational cost of the facility. Current Center activities have been focused on integrating advanced safeguards monitoring technologies demonstrations into the SRS H-Canyon and advanced location technologies demonstrations into K-Area Materials Storage. These demonstrations are providing valuable information to researchers and customers as well as providing the Center with an improved protocol for demonstration management that can be exercised across the entire SRS (as well as to offsite venues) so that future demonstrations can be done more efficiently and provide an opportunity to utilize these unique assets for multiple purposes involving national laboratories, academia, and commercial entities. Key among the envisioned future demonstrations is the use of H-Canyon to demonstrate new nuclear materials separations technologies critical for advancing the mission needs DOE-Nuclear Energy (DOE-NE) to advance the research for next generation fuel cycle technologies. The concept is to install processing equipment on frames. The frames are then positioned into an H-Canyon cell and testing in a relevant radiological environment involving prototypic radioactive materials can be performed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗