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At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

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

Development of flaw acceptance criteria for aging management of spent nuclear fuel multiple-purpose canisters

A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic In-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Validation Study for Crediting Chlorine in Criticality Analyses for Spent Nuclear Fuel Disposition

Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10 -5 k eff ). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.

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

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k eff ) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup credit at peak reactivity requires a different set of experiments than for pressurized-water reactor burnup credit analysis because of differences in actinide compositions, presence of residual gadolinium absorber, and lower fission product concentrations. A survey of available critical experiments is presented along with a sample criticality code validation and determination of undercoverage penalties for some nuclides. The validation of depleted fuel compositions at peak reactivity presents many challenges which largely result from a lack of radiochemical assay data applicable to BWR fuel in this burnup range. In addition, none of the existing low burnup measurement data include residual gadolinium measurements. An example bias and uncertainty associated with validation of actinide-only fuel compositions is presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Isotopic Analysis of Spent Nuclear Fuel with an Ultra-High Rate HPGe Spectrometer

A longstanding challenge is the assay of spent nuclear fuel (SNF). Determining the isotopic content of SNF requires gamma-ray spectroscopy. PNNL has developed new digital filtering and analysis techniques to produce an ultra high-rate gamma-ray spectrometer from a standard coaxial high-purity germanium (HPGe) crystal. This ~40% efficient detector has been operated for SNF measurements at a throughput of about 400k gamma-ray counts per second (kcps) at an input rate of 1.3 Mcps. Optimized filtering algorithms preserve the spectroscopic capability of the system even at these high rates. This talk will present the results of a SNF measurement with aged SNF pellets at PNNL’s Radiochemical Processing Laboratory, first results with a FPGA front end processor capable of processing the data in real time, and the development path toward a multi-element system to assay fuel assemblies.

Fast, James E.↗
DOE OSTI2016

Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may find other safety and security or safeguards applications, such as arms control verification.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2014

ON-LINE MONITORING FOR PROCESS CONTROL AND SAFEGUARDING OF RADIOCHEMICAL STREAMS AT SPENT FUEL REPROCESSING PLANTS

The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent nuclear fuel reprocessing plants to ensure that significant quantities of weapons-grade nuclear material are not diverted from these facilities. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resource-intensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) techniques in conjunction with the traditional and highly precise DA methods may provide a more timely, cost-effective and resource-efficient means for MC&A verification at such facilities. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies, including a spectroscopy-based monitoring system, to potentially reduce the time and resource burden associated with current techniques. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major non-radioactive flowsheet chemicals using UV-vis-near infrared and Raman spectroscopy. This paper provides an overview of the methods and reports our on-going efforts to develop and demonstrate the technologies. Our ability to identify material intentionally diverted from a liquid-liquid extraction contactor system was successfully tested using on-line process monitoring as a means to detect the amount of material diverted. A chemical diversion and detection of that diversion, from a solvent extraction scheme was demonstrated using a centrifugal contactor system operating a tributyl phosphate based extraction. A portion of the feed from a counter-current extraction system was diverted while a continuous extraction experiment was underway; the spectroscopic on-line process monitoring system was simultaneously measuring the feed, raffinate and organic products streams. The amount observed to be diverted by on-line spectroscopic process monitoring was in excellent agreement with values based from the known mass of sample directly diverted from system feed solution.

Np↗
DOE OSTI2017

Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification

The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup, under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes. The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.

Jacobsson-Svard, Staffan↗
DOE OSTI2017

Standardized Canisters for Spent Nuclear Fuel: Their Potential Impact and a Proposed Path Forward

This paper describes an evaluation completed in Fiscal Year 2016 (FY16) that investigates using standardized triple-purpose (storage, transportation, and disposal) canisters within an integrated waste management system (IWMS). The IWMS in this particular evaluation is focused on management of commercial spent nuclear fuel (SNF). Incorporation of a standardized canister system into the IWMS provides an opportunity to develop an integrated approach to address storage, transportation, and disposal issues. Regardless of implementation timing and method, deployment of such a canister system could cause significant system-wide impacts. Over the past few years, a number of system analyses have been performed for the US Department of Energy (DOE), Office of Nuclear Energy (NE) using logistic simulations to evaluate the impacts of incorporating these smaller canisters into the IWMS. Furthermore, these simulations predict the amount of resources (e.g., staff, canisters/casks, railcars, facilities), the size of facilities, the timing of operations, and rough-order-of-magnitude (ROM) cost estimates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Method for energy recovery of spent ERL beams

A method for recovering energy from spent energy recovered linac (ERL) beams. The method includes adding a plurality of passive decelerating cavities at the beam dump of the ERL, adding one or more coupling waveguides between the passive decelerating cavities, setting an adequate external Q (Qext) to adjust to the beam loading situation, and extracting the RF energy through the coupling waveguides.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2018

Detailed SCALE Dose Rate Evaluations for a Consolidated Interim Spent Nuclear Fuel Storage Facility

This paper presents a confirmatory dose rate calculation for a proposed consolidated interim spent nuclear fuel (SNF) storage facility (CISF). The proposed CISF was modeled in detail using SCALE 6.2.1 to determine the radiation dose map of the surrounding areas of the facility and determine the location of the controlled area boundary based on 10 CFR 72.104 requirements. The spatial dose rate distribution was determined as the total dose rate contributions by each of the 467 storage casks comprising the CISF. A rectangular boundary outside of which no dose rate exceeds 25 mrem/yr has the dimensions of 1.25 km x 1.25 km. It denotes the smallest rectangle that could serve as the controlled area boundary. The results (dose map) show the unique capability of the SCALE code system for a high-resolution radiation dose simulation of a large SNF storage facility.

Radulescu, Georgeta↗
DOE OSTI2018

Muon-Computed Tomography using POCA Trajectory for Imaging Spent Nuclear Fuel in Dry Storage Casks

Cosmic ray muon-computed tomography (μCT) is a new imaging modality with unique characteristics that can be particularly important for applications in nuclear proliferation detection and international treaty verification. Using cosmic ray muons for nuclear security presents several potential advantages. Among others, muons are generated naturally in the atmosphere, can penetrate high-density materials, and are freely available. No radiological sources are required and consequently there is no associated radiological dose. Recently, the feasibility of using muons for imaging spent nuclear fuel stored in shielded casks has been explored and has been proved beneficial. However, challenges in μCT imaging include low muon flux of ~10,000 muons/m2/min, the effects of multiple Coulomb scattering (MCS) blurring the image, and inefficiency in being able to use all recorded muons for imaging. In this paper, we argue that the use of muon tracing should produce tomographic muon images with improved quality – or more quickly for the same image quality – compared to the case where conventional methods are used. In our paper, we report on the development and assessment of a novel muon tracing method for μCT. The proposed method back projects the muon’s scattering angle into each pixel crossed by its PoCA trajectory then forward projects the variance of the scattering angle in each pixel to detector bins along the muon’s incident horizontal direction. Two scenarios were simulated to assess the expected detection capability of this proposed method. GEANT4 was used to model the main characteristics of 1-60 GeV muons through matter. The simulated images showed an expected improvement in resolution and a reduced the reliance on the muon momentum information compared to a more conventional muon tomography method.

Liu, Zhengzhi↗
DOE OSTI2018

DOSE RATE VERIFICATION FOR SNS SPENT STRUCTURES

Residual dose rate measurements were conducted on two Spallation Neutron Source spent structures, target vessel #13 and proton beam window #5, after extraction from their service locations. These measurements were used to verify calculation methods for radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation. Neutronics analyses for predicting residual dose rates were carried out using complex and rigorous geometric models of the structures and their surroundings. Calculational results were analyzed and compared to the measured dose rates

Popova, Irina I.↗
DOE OSTI2019

Neutron Residual Stress Mapping of Repaired Spent Nuclear Fuel Welded Stainless-Steel Canisters

Stress corrosion cracking (SCC) caused by atmospheric deposition and deliquescence of aggressive compounds such as chloride-containing salts or oxidized sulfur species may pose a potential issue to the extended service life of spent nuclear fuel (SNF) storage canisters. When a chemically susceptible material is exposed to a hostile environment (e.g., coastal salt air) for a sufficient length of time, SCC may occur at locations on the canister which have undergone high tensile residual stress. These locations are often found in weld zones and the heat-affected zones (HAZ) of welded joints. Ideally, these cracks could be repaired by traditional fusion welding techniques, but the highly localized heat input of the repair welding may introduce additional high tensile residual stress and could increase the risk for future SCCs. This paper presents the results of neutron residual stress mapping of as-welded and repaired stainless steel specimens. The as-welded specimens were provided by Sandia National Laboratories from a canister mock-up manufactured using procedures similar to those used for actual canister production. ASME-qualified gas-tungsten arc welding (GTAW) was used to repair the specimens, and post-repair residual stresses were measured using neutron diffraction at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). As-welded and repaired specimen residual stress distributions are evaluated and compared.

Chatzidakis, Stylianos↗
DOE OSTI2019

Detailed Radiation Dose Rate Evaluations of Commercial Spent Nuclear Fuel Canisters

Electronic devices that are being designed for continuous monitoring of spent nuclear fuel (SNF) canister internal integrity may be subjected to very high radiation levels depending on their placement. Components of electronic devices (e.g., silicon semiconductors, optical glasses, insulators) have a limited tolerance to ionizing radiation. For selected locations on the inner and outer surfaces of a representative dry storage pressurized water reactor (PWR) SNF canister and a representative dry storage boiling water reactor (BWR) SNF canister, this paper provides (1) the dose rate in silicon as a function of decay time and (2) the time-integrated dose in silicon. The analyzed decay time interval is from 5 to 70 years after fuel discharge. Each representative canister contained identical SNF assemblies with bounding radiation sources. Surface dose rates varied significantly as a function of location and SNF assembly decay time. At a 5-year decay time, the estimated maximum dose rate values were 480 Gy/h (PWR canister) and ~366 Gy/h (BWR canister), and the estimated minimum dose rate was ~0.2 Gy/h for both canisters. Over a 65-year period, the dose rate decreased by a factor between 13 and 580, depending on location. The total dose over a time interval could be very large (e.g., estimated maximum value over a 5-year interval is 1.5E+07 Gy) depending on location and the time interval over which dose is accumulated.

Radulescu, Georgeta↗
DOE OSTI2019

Dose to Cementitious Material for Moderator Exclusion in a Spent Nuclear Fuel Canister During a Repository Timeframe

One of the concerns regarding direct disposal of spent nuclear fuel (SNF) dual-purpose canisters (DPCs) currently in dry storage is criticality during disposal time frames (e.g., 10,000 years or more). Preconditioning DPCs with appropriate filler materials (e.g., cementitious materials, low-melting metals) before disposal could prevent post-closure criticality by means of moderator displacement. However, potential radiolysis of the filler material by ionizing radiation and release of reactive chemical species over repository time frames may negatively impact the mechanical and chemical properties of the filler materials. Evaluation of the energy deposited in candidate filler materials by SNF radiation is essential for the calculation of radiolytic species. This paper provides an evaluation of the total energy deposited by gamma and neutron radiation in a candidate filler material over 10,000 years for a representative pressurized water reactor SNF DPC. The canister is assumed to contain identical SNF assemblies with a burnup value of 55 GWd/MTU, which is the bounding burnup value for more than 95% of the SNF inventory discharged from 1968 to 2013. The evaluated energy deposited per unit volume of filler material between 10 years and 10,000 years after fuel discharge is approximately 9.3E+23 eV/cm3. Gas production from radiolysis was not evaluated and may be revisited in future work. Radiolysis may be the principal gas generation mechanism for a cementitious system.

Radulescu, Georgeta↗
DOE OSTI2019

Criticality Safety Analysis of Spent Nuclear Fuel Canisters Using As-Loaded Configurations

Dual purpose canisters (DPCs) used for storage and transportation of spent nuclear fuel (SNF) are typically designed and evaluated using bounding (enveloping) fuel characteristics such as fuel type, fuel dimensions, initial enrichment, discharge burnup, and cooling time. This is a design basis, bounding licensing approach for SNF storage and transportation systems, as licensing and supporting safety analysis reviews are performed prior to the actual fuel loading. The bounding fuel characteristics for a system are developed by fully utilizing the regulatory and regulators recommended safety limits such as neutron multiplication factor (keff) approaching 0.95 to maximize the system utilizations. In reality, there are wide variations in SNF assembly burnups, initial enrichments, and cooling times. Therefore, dry storage systems are typically loaded with assemblies that satisfy the bounding fuel characteristics defined in the certificate of compliance (CoC), with some amount of unquantified and uncredited margin. This paper presents an as-loaded canister-specific criticality analysis approach for quantifying inherent uncredited margins in already loaded DPCs. The as-loaded analysis approach has been implemented in a new SNF management and analysis tool - The Used Nuclear Fuel-Storage, Transportation, and Disposal Analysis Resource and Data System (UNF ST&DARDS). The paper presents as-loaded criticality analysis of 616 canisters at 28 US reactor sites. Additionally, the paper discusses as-loaded analysis methodology validation approach using detailed reactor operational and fuel design data.

Banerjee, Kaushik↗
DOE OSTI2017

Nanoscale Compositional Mapping of Spent HZSM-5: An Atom Probe Tomography Study

Zeolites are highly crystalline and porous materials possessing a well-defined one- to three-dimensional pore structure. Zeolites play a major role in the petrochemical industry, exploitation of non-conventional petroleum resources and nonpetroleum resources like gas and coal. Recently, zeolite has been researched heavily in the valorization of biomass for renewable fuel and chemical generation. It has been reported widely that zeolite deactivation can occur via two main mechanisms. The first one is leaching of alumina from the unit structure of the zeolite in the presence of steam; this mechanism causes the irreversible deactivation. The second mechanism is the formation of coke during the transformation of organic compounds in the pores and/or the outer surface and the poisoning of the active sites and/or blocking of the pores. The catalyst deactivation from this mode can be regenerated by burning the catalyst under oxidative conditions. Repeated thermal treatment for spent catalyst regeneration is also shown to cause permanent modification of the zeolite structure. So identifying the nature and the location of the coking material and changes in Al distribution in catalyst after catalytic reactions becomes imperative. Recently, a relatively new technique called atom probe tomography (APT) has been increasingly used to perform 3D quantitative analysis of element distribution with sub-nanometer scale spatial resolution and high mass sensitivity up to the few ppm level of concentration of elements in materials. In the study we used a multimodal approach including C K-edge x-ray absorption along with 13C cross polarization magic angle spinning nuclear magnetic resonance (CP-MAS NMR) and APT to investigate element distribution within a HZSM-5 as-synthesized condition and after ethanol conversion reaction.

Devaraj, Arun↗