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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Passive methods for spent fuel characterisation at the Finnish geological repository

Development of the Passive Gamma Emission Tomography (PGET) and Passive Neutron Albedo Reactivity (PNAR) methods in the context of the Finnish geological repository for spent nuclear fuel has shown that they provide, for BWR fuel assemblies, the comprehensive verification needed to meet the nuclear safeguards objectives of the repository. The principles of the PGET and PNAR methods and the design and operation of the respective instruments are presented. A few results from measurements at the spent fuel storage pools at the Finnish nuclear power plants are discussed. The directions of ongoing and future developments are indicated.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2013

Analysis of Dose Consequences Arising from the Release of Spent Nuclear Fuel from Dry Storage Casks

The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level – 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

Durbin, Samuel G.↗
DOE OSTI2013

The Use of Non Destructive Assay Measurements to Estimate Burnup and Plutonium Inventory in Pressurized Water Reactor Spent Fuel Assemblies

One of the overarching goals of the Next Generation of Safeguards Initiative (NGSI) Spent Fuel (SF) Non-Destructive Assay (NDA) Project is to determine the Pu mass in spent fuel assemblies with NDA instruments.1 Other goals include verifying initial enrichment, burn-up and cooling time of declaration, detecting diversion or replacement of pins, and/or determining heat emitted from an assembly (not a safeguards goal but important nonetheless). In pursuit of these goals, various NDA instruments of interest need to be built, tested, and then calibrated. Ideally simple calibration sources can be used such as isotopic sources or fresh assemblies. However, given the fact that many of the measured signals are influenced by the complex, energy-dependent interaction of neutrons with a range of isotopes, the connection between the measured signals and the Pu mass is difficult to make.

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DOE OSTI2014

Spent Fuel Assembly Modeling for Unattended Gamma Emission Tomography (UGET) Project [Slides]

NGSI-SF project provided good basis for spent fuel assemblies. UGET assemblies have both DU pins and water replacing fuel in the center of the bundle. Both MCNP geometry files and CINDER90 isotopic compositions files were created and used by LLNL and/or others. Four different spent fuel assembly types were represented: PWR, BWR, VVER-440, and the VVER-1000. Not enough information was known about irradiation of assemblies in Forsmark or Olkiluoto (JNT 1510) benchmarks for detailed calculations.

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DOE OSTI2014

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

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

Spent Nuclear Fuel and Reprocessing Waste Inventory

This report provides information on the inventory of spent nuclear fuel (SNF) in the United States located at Nuclear Power Reactor (NPR) and Independent Spent Fuel Storage Installation (ISFSI) sites, as well as SNF and reprocessing waste located at U.S. Department of Energy (DOE) sites and other research and development (R&D) centers. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future nuclear power generation scenarios. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).

Peters, Shan↗
DOE OSTI2012

Automated Characterization of Spent Fuel through the Multi-Isotope Process (MIP) Monitor

This research developed an algorithm for characterizing spent nuclear fuel (SNF) samples based on simulated gamma spectra. The gamma spectra for a variety of light water reactor fuels typical of those found in the United States were simulated. Fuel nuclide concentrations were simulated in ORIGEN-ARP for 1296 fuel samples with a variety of reactor designs, initial enrichments, burn ups, and cooling times. The results of the ORIGEN-ARP simulation were then input to SYNTH to simulate the gamma spectrum for each sample. These spectra were evaluated with partial least squares (PLS)-based multivariate analysis methods to characterize the fuel according to reactor type (pressurized or boiling water reactor), enrichment, burn up, and cooling time. Characterizing some of the features in series by using previously estimated features in the prediction greatly improves the performance. By first classifying the spent fuel reactor type and then using type-specific models, the prediction error for enrichment, burn up, and cooling time improved by a factor of two to four. For some features, the prediction was further improved by including additional information, such as including the predicted burn up in the estimation of cooling time. The optimal prediction flow was determined based on the simulated data. A PLS discriminate analysis model was developed which perfectly classified SNF reactor type. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE.

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DOE OSTI2016

ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.

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DOE OSTI2016

Evaluation of Pad 18 Spent Mercury Gold Trap Stainless Steel Container Failure

Failure of the Pad 18 spent mercury gold trap stainless steel waste container is principally attributed to corrosion induced by degradation of plasticized polyvinyl chloride (pPVC) waste packaging material. Dehydrochlorination of pPVC polymer by thermal and/or radiolytic degradation is well-known to evolve HCl gas, which is highly corrosive to stainless steel and other metals in the presence of moisture. Degradation of the pPVC packaging material was likely caused by radiolysis in the presence of tritium gas within the waste container, though other degradation mechanisms (aging, thermo-oxidation, plasticizer migration) over 30 years storage may have contributed. Corrosion was also likely enhanced by the crevice in the container weld design, and may have been enhanced by the presence of tritiated water. Similar non-failed spent mercury gold trap waste containers did not show radiographic evidence of plastic packaging or trapped free liquid within the container. Therefore, those containers are not expected to exhibit similar failures. Halogenated polymers such as pPVC subject to degradation can evolve halide gases such as HCl, which is corrosive in the presence of moisture and can generate pressure in sealed systems.

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DOE OSTI2017

Analyses of Field Test Data at the Atucha-1 Spent Fuel Pools

A field test was conducted at the Atucha-1 spent nuclear fuel pools to validate a software package for gross defect detection that is used in conjunction with the inspection tool, Spent Fuel Neutron Counter (SFNC). A set of measurements was taken with the SFNC and the software predictions were compared with these data and analyzed. The data spanned a wide range of cooling times and a set of burnup levels leading to count rates from the several hundreds to around twenty per second. The current calibration in the software using linear fitting required the use of multiple calibration factors to cover the entire range of count rates recorded. The solution to this was to use power regression data fitting to normalize the predicted response and derive one calibration factor that can be applied to the entire set of data. The resulting comparisons between the predicted and measured responses were generally good and provided a quantitative method of detecting missing fuel in virtually all situations. Since the current version of the software uses the linear calibration method, it would need to be updated with the new power regression method to make it more user-friendly for real time verification and fieldable for the range of responses that will be encountered.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2018

Neutron Residual Stress Mapping for Spent Nuclear Fuel Storage Canister Weldment

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 that have no current outlet than storage. These materials are often held in long-term storage as an interim stage within their lifecycle. Lifecycles can include reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. Stress corrosion cracking of interim storage containers has been identified as a high priority data gap by the Department of Energy. However, little has been done with regards to the canister weld material properties and their impact on stress corrosion, until recently from Sandia National Laboratory canister mockup program. Furthermore, because no post-weld heat treatment was required, the associated high tensile residual stress within these canister welds can drive the initiation and growth of stress corrosion cracking (SCC) cracks. In this work, we carried out in-situ neutron residual stress mapping on a weld plate sectioned from Sandia mock-up canister wall. Significant residual stress profiles were identified from the received longitudinal weld sample. The details of the weldment residual stress program development at ORNL are presented in this progress report. The generated residual stress profile from neutron diffraction technique is similar to that of Sandia residual stress data obtained from the deep hole drilling approach.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Spent Fuel NDA Project: Data Authentication Considerations

The overall spent fuel nondestructive assay project seeks to develop improved measurement capability for the verification of spent nuclear fuel, especially before its disposal or movement to hard-to-access storage. Various systems are being considered, employing neutron and/or gamma radiation detection with either passive or active methods; their use scenarios are not yet well defined. In a practical deployment, the measurement system would likely need to operate in unattended mode. The output results may also need to be shared between multiple recipients with various interests. The data authentication task considers what issues are important in being able to trust the measurement results. By defining and analyzing a generic (3z(Bbaseline(3y (Bsystem scenario, we have identified five key factors needing specific attention: application use-case details, equipment tamper indication, supporting (ancillary) instruments, systems implementation, and instrument state-of-health reporting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Public Views about Storage and Disposal Options for Spent Nuclear Fuel: Energy and Environment Survey, 2017

The US currently has about 80,000 metric tons of uranium in 279,000 used fuel assemblies from commercial spent nuclear fuel (SNF), most of which is stored “on-site” at or near power plants where it was produced. On-site storage facilities were not designed to provide a permanent solution for the disposal of spent nuclear fuel, and building a permanent disposal facility will likely take decades. Therefore, it is important to consider how the US public views options for constructing one or more storage facilities for safely consolidating and storing SNF in the interim. The 2017 iteration of the Energy and Environment survey (EE17) by the Center for Energy, Security, & Society (CES&S) included a battery of questions that measure public views about SNF storage and disposal options. The questions gauge general support for continued on-site storage, interim storage, and permanent disposal. EE17 also measured support for several of the specific sites under consideration, including the two private initiatives for interim storage of SNF in New Mexico and Texas. In addition, EE17 respondents provide insight into the factors likely to affect broader public support for these initiatives as the siting process unfolds, including public views about the importance of support for a prospective facility by host communities and host state residents.

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DOE OSTI2018

Development of a Visualization Concept for Robotic Operations to Consolidate Commercial Spent Nuclear Fuel

Sandia and NAC are advancing a new conceptual initiative associated with integrating the back end of the nuclear fuel cycle. Sandia has robust capability in robotics and spent fuel disposal science and engineering. NAC has demonstrated rod consolidation packaging technology that potentially enables more efficient nuclear fuel cycle back end integration. NAC also has a robust capability in spent fuel packaging.

Freeze, Geoffrey↗
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 OSTI2019

Gap Analysis to Guide DOE R&D in Supporting Extended Storage and Transportation of Spent Nuclear Fuel: An FY2019 Assessment

The Department of Energy(DOE), Office of Nuclear Energy (NE), Spent Fuel and Waste Science and Technology (SFWST) program is performing research and development in the area of commercial spent nuclear fuel (SNF) long term storage and transportation. This program is being conducted under the provisions of the Nuclear Waste Policy Act (NWPA) of 1982 and its amendments that require the DOE to take title to and manage SNF after storage at the utility reactor site. This report is a condensed version of previous gap reports (Hanson 2012 and Hanson 2019) with up-dated gap priority assessments. The gap priorities have been updated from Hanson 2019 because 2019 is based on R&D performed through 2017. Much important work has been done since 2017 that requires a change in a few of the priority rankings to better focus the near-term R&D program. Background material, regulatory positions, operational and inventory status, and prioritization schemes are discussed in detail in Hanson 2012/2019, and are not repeated in this report. One exception is an overview of the prioritization criteria for reference. This is meant to give the reader an appreciation of the framework for prioritization of the identified gaps. A complete discussion of the prioritization scheme is provided in Hanson 2019.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Cost estimation inputs for spent nuclear fuel geologic disposal concepts (Revision 1)

A set of 16 geologic disposal concepts is described in sufficient detail for rough-order-of-magnitude repository cost estimates, for disposal of spent nuclear reactor fuel in generic crystalline, argillaceous, and salt host geologic media. The description includes total length, diameter, and volume for all underground shafts, ramps, drifts and large-diameter borings. Basic types of ground support are specified. Total repository capacity is assumed to be approximately 140,000 MT of spent fuel, but concepts are described in terms of modular panels each containing 10,000 MT. Waste packaging is described, and the materials and outer dimensions for disposal overpacks are given. The manner of emplacement is specified, with any additional fixturing, lining, buffer materials, and backfill needed. Thermal limits for waste package emplacement or repository closure are given, as appropriate for enclosed and open emplacement modes, respectively. The 16 disposal concepts are based on disposal concept studies performed for the Used Fuel Disposition R&D program between 2011 and 2014, and the accompanying engineering analyses. Revision 1 results from review and checking of repository spacings, drift length estimates, repository plan area estimates, closure power limits, and estimates of fuel age at closure.

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