Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “SPENT”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Preliminary Evaluation of Removing Spent Nuclear Fuel from Shutdown Nuclear Power Plant Sites - 2018 - 19101

This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. The U.S. Department of Energy Office of Integrated Waste Management (DOE-IWM) continues to conduct evaluations of removing spent nuclear fuel (SNF) from shutdown nuclear power plant sites. The 14 shutdown sites included in the evaluation to date have nuclear power reactors that have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The 14 shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun. The Oyster Creek Nuclear Generating Station ceased operations in September 2018 and will be included in future evaluations. The evaluations of the shutdown sites contain four components: - A characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; - A description of the on-site infrastructure at the shutdown sites; - An evaluation of the near-site transportation infrastructure and experience at the shutdown sites - An evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. The primary sources for the inventory of SNF were the DOE GC-859 and RW-859 spent nuclear fuel inventory databases. This inventory has been augmented with data obtained from the shutdown sites for SNF discharges that occurred after June 2013. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included field observations and information collected during visits to the 14 shutdown sites, information provided by managers and staff at the shutdown sites, and Google Earth imagery. Extensive photographs taken during the shutdown site visits have proven to be particularly useful in documenting the current conditions at and near the shutdown sites. Tribal and state representatives, State and Regional Group representatives, and Federal Railroad Administration representatives have participated in 11 of the shutdown site visits. In addition to visiting the shutdown site, site visits often include activities such as meeting with local railroads to discuss the near-site railroad infrastructure and meeting with community engagement panels. Oyster Creek permanently ceased power operations on September 17, 2018 and fuel was removed from the reactor vessel on September 25, 2018. Planning for the Oyster Creek site visit has begun and is conducted in three stages. In the first stage, data and documents that are applicable to Oyster Creek are collected. In the second stage, questions are developed based on the data and documents collected during the first stage. These questions are submitted to the Oyster Creek site approximately one month before the site visit. The third stage involves establishing specific schedules and participants. Participants will likely include tribal and state representatives, the Council of State Governments - Eastern Regional Conference, and the Federal Railroad Administration. State participation will likely include a representative of the State Rail Safety Participation Program, the New Jersey Department of Transportation, and the New Jersey Bureau of Nuclear Engineering. The site visit is tentatively expected to occur in April or May of 2019. (authors)

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

Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 20466

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating and testing prototype railcars that will comply with the safety standard S-2043. This standard was developed by the Association of American Railroads (AAR) specifically for railcars used to transport High-Level Radioactive Material (HLRM). AAR defines the term HLRM to include both spent nuclear fuel (SNF) and high-level radioactive waste (HLW). DOE is in the process of developing and testing prototype railcars that will satisfy Standard S-2043. This is a technical paper that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this paper in no manner supersedes, overrides, or amends the Standard Contract. This paper reflects technical work which could support future decision making by the Department of Energy (DOE or Department). No inferences should be drawn from this paper regarding future actions by DOE, which are limited both by the terms of the Standard Contract and a lack of Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository. (authors)

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

Preliminary Analysis of Advanced Reactor Spent Nuclear Fuel Storage, Transportation, and Disposal

Due to increased interest in advanced reactor deployment and their associated potential new fuel cycles, the U.S. Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program has begun to evaluate the possible implications of long term management and final disposition of the spent nuclear fuel (SNF) generated. Safely managing and dispositioning this SNF, along with any other associated radioactive waste, is the primary focus of this initial preliminary assessment. This paper summarizes efforts to evaluate the characteristics and packaging options for three types of advanced reactor SNF forms: (1) tristructural isotropic (TRISO), (2) metallic, and (3) irradiated fuel salt presented in the report titled “Storage, Transportation, and Disposal of Advanced Reactor Spent Nuclear Fuel and High-Level Waste”. TRISO and metallic SNF and their associated waste streams were emphasized because of the near-term anticipated demonstrations of X-energy’s Xe-100 and TerraPower and GE Hitachi’s Natrium advanced reactors. Preliminary information on spent fuel salts discharged from molten-salt reactors (MSRs) was also examined to provide a baseline for future efforts. All calculations and assumptions were based on publicly available information. This paper identifies several different reactors that produce either TRISO or metallic SNF as well as a few of the reactor and fuel characteristics used for the assessments. Based on these characteristics, calculations were performed to determine the applicability of packaging SNF into existing canister designs. The evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, and as-loaded radiation, thermal, and criticality analyses to identify constraints on storage, transportation, and disposal. Gaps in publicly available data pertaining to reactor operation and/or fuel composition provide increased uncertainty in some evaluations. Additionally, uncertainty in packaging and SNF management operations provide additional uncertainty. However, preliminary conclusions can still be assessed through this work and are presented in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Thermoelectric powered wireless sensors for spent fuel monitoring

This paper describes using thermoelectric generators to power wireless sensors to monitor spent nuclear fuel during dry-cask storage. OrigenArp was used to determine the decay heat of the spent fuel at different times during the service life of the dry-cask. The Engineering Equation Solver computer program modeled the temperatures inside the spent fuel storage facility during its service life. The temperature distribution in a thermoelectric generator and heat sink was calculated using the computer program Finite Element Heat Transfer. From these temperature distributions the power produced by the thermoelectric generator was determined as a function of the service life of the dry-cask. In addition, an estimation of the path loss experienced by the wireless signal can be made based on materials and thickness of the structure. Once the path loss is known, the transmission power and thermoelectric generator power requirements can be determined. This analysis estimates that a thermoelectric generator can produce enough power for a sensor to function and transmit data from inside the dry-cask throughout its service life. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Electrochemical performance studies of MnO{sub 2} nanoflowers recovered from spent battery

Highlights: • MnO{sub 2} is recovered from spent zinc–carbon batteries as nanoflowers structure. • Recovered MnO{sub 2} nanoflowers show high specific capacitance. • Recovered MnO{sub 2} nanoflowers show stable electrochemical cycling up to 900 cycles. • Recovered MnO{sub 2} nanoflowers show low resistance in EIS data. - Abstract: The electrochemical performance of MnO{sub 2} nanoflowers recovered from spent household zinc–carbon battery is studied by cyclic voltammetry, galvanostatic charge/discharge cycling and electrochemical impedance spectroscopy. MnO{sub 2} nanoflowers are recovered from spent zinc–carbon battery by combination of solution leaching and electrowinning techniques. In an effort to utilize recovered MnO{sub 2} nanoflowers as energy storage supercapacitor, it is crucial to understand their structure and electrochemical performance. X-ray diffraction analysis confirms the recovery of MnO{sub 2} in birnessite phase, while electron microscopy analysis shows the MnO{sub 2} is recovered as 3D nanostructure with nanoflower morphology. The recovered MnO{sub 2} nanoflowers exhibit high specific capacitance (294 F g{sup −1} at 10 mV s{sup −1}; 208.5 F g{sup −1} at 0.1 A g{sup −1}) in 1 M Na{sub 2}SO{sub 4} electrolyte, with stable electrochemical cycling. Electrochemical data analysis reveal the great potential of MnO{sub 2} nanoflowers recovered from spent zinc–carbon battery in the development of high performance energy storage supercapacitor system.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Recovery of manganese oxides from spent alkaline and zinc–carbon batteries. An application as catalysts for VOCs elimination

Highlights: • Manganese oxides were synthesized using spent batteries as raw materials. • Spent alkaline and zinc–carbon size AA batteries were used. • A biohydrometallurgical process was employed to bio-lixiviate batteries. • Manganese oxides were active in the oxidation of VOCs (ethanol and heptane). - Abstract: Manganese, in the form of oxide, was recovered from spent alkaline and zinc–carbon batteries employing a biohydrometallurgy process, using a pilot plant consisting in: an air-lift bioreactor (containing an acid-reducing medium produced by an Acidithiobacillus thiooxidans bacteria immobilized on elemental sulfur); a leaching reactor (were battery powder is mixed with the acid-reducing medium) and a recovery reactor. Two different manganese oxides were recovered from the leachate liquor: one of them by electrolysis (EMO) and the other by a chemical precipitation with KMnO{sub 4} solution (CMO). The non-leached solid residue was also studied (RMO). The solids were compared with a MnO{sub x} synthesized in our laboratory. The characterization by XRD, FTIR and XPS reveal the presence of Mn{sub 2}O{sub 3} in the EMO and the CMO samples, together with some Mn{sup 4+} cations. In the solid not extracted by acidic leaching (RMO) the main phase detected was Mn{sub 3}O{sub 4}. The catalytic performance of the oxides was studied in the complete oxidation of ethanol and heptane. Complete conversion of ethanol occurs at 200 °C, while heptane requires more than 400 °C. The CMO has the highest oxide selectivity to CO{sub 2}. The results show that manganese oxides obtained using spent alkaline and zinc–carbon batteries as raw materials, have an interesting performance as catalysts for elimination of VOCs.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Significance of a Solid Electrolyte Interphase on Separation of Anode and Cathode Materials from Spent Li-Ion Batteries by Froth Flotation

The separation of electrode active materials from spent Li-ion batteries (LIBs) by froth flotation is challenging due to the changes in surface properties of electrode active materials from cycling as well as the presence of organic binders. In this work, the froth flotation separation of aged anode and cathode composite materials from spent LIBs was systematically investigated after the materials were heat treated. The results show that aged anode and cathode materials from spent LIBs can be well separated from each other after a heating process in air at 400 degrees C and at which some of the PVDF binder remains intact. The underlying mechanism was investigated by X-ray photoelectron spectroscopy (XPS), contact angle measurements, and scanning transmission electron microscopy (STEM) coupled with energy-dispersive X-ray spectroscopy (EDX). The results from the XPS and contact angle measurements show that there is a hydrophilic and oxygen-rich layer on the surface of aged anode materials. This hydrophilic surface, associated with the solid electrolyte interface (SEI) layer, impacts the froth flotation process significantly. The results also show that both the SEI layers and PVDF binder residues on the surface are removed at 400 degrees C for an hour, restoring the hydrophobicity of the anode materials, which, in turn, benefits the separation of anode and cathode materials. The STEM/EDX elemental analysis data confirms that there are 20 nm-thick oxygen-rich SEI layers on the surfaces, which can be removed after a heating process. The present result illustrates the significance of the SEI layers in flotation separation of electrode materials and sheds new lights into the future development of the recycling processes for the separation of anode and cathode composite materials from spent Li-ion batteries.

25 ENERGY STORAGE↗
DOE OSTI2014

The characteristic assessment of spent ion exchange resin from PUSPATI TRIGA REACTOR (RTP) for immobilization process

In this paper, spent ion exchange resin generated from PUSPATI TRIGA reactor (RTP) in Malaysian Nuclear Agency were characterized based on the water content, radionuclide content and radionuclide leachability. The result revealed that the water content in the spent resin is 48%. Gamma spectrometry analysis indicated the presence of {sup 134}Cs, {sup 137}Cs, {sup 152}Eu, {sup 54}Mn, {sup 58}Co, {sup 60}Co and {sup 65}Zn. The leachability test shows a small concentrations (<1 Bq/l) of {sup 152}Eu and {sup 134}Cs were leached out from the spent resin while {sup 60}Co activity concentrations slightly exceeded the limit generally used for industrial wastewater i.e. 1 Bq/l. Characterization of spent ion exchange resin sampled from RTP show that this characterization is important as a basis to immobilize this radioactive waste using geopolymer technology.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Structural dynamics modeling of spent nuclear fuel during hypothetical package drop events

The response of spent nuclear fuel (SNF) to hypothetical package drop events is of particular interest in the scope of spent fuel storage and transportation because of the mechanical shock encountered in such scenarios. Previous testing and modeling by the U.S. Department of Energy has demonstrated that the shock and vibration environment of normal shipping and handling conditions (excluding package drop events) is relatively benign and does not challenge the integrity of spent nuclear fuel. Cask drop events are worth considering because SNF packages are required to withstand free drops onto unyielding surfaces as part of their licensing basis. The acceleration experienced during drop events can be orders of magnitude higher, and thus more advanced models are needed to encompass potential nonlinear behavior of the fuel, such as spacer grid buckling and rod-to-rod impact. This work describes a number of finite element models developed to calculate the response of spent nuclear fuel to various hypothetical drop events that have been validated by package and fuel assembly drop tests conducted in the last decade. Sensitivity of the model response to factors such as package drop orientation, secondary impacts, and irradiated material properties as well as their potential impacts to fuel cladding integrity, was also investigated. Cask drops are not expected as a regular occurrence during SNF transportation, but this work helps raise the understanding of SNF mechanical loads to the point of consistency with the package design requirements.

Kadooka, Kevin↗
DOE OSTI2021

Spent Nuclear Fuel Mechanical Loads in the General Package Drop Scenario

The US Department of Energy Spent Fuel and Waste Science and Technology program is performing research to determine the mechanical loading conditions applied to spent nuclear fuel (SNF) during normal conditions of transportation to inform mechanical tests of SNF and close an important knowledge gap related to the practical disposition of SNF in the US. A recent multi-national collaborative test campaign measured SNF assembly impact response to the 30 cm horizontal package drop scenario, which is a common regulatory basis test of SNF package design. Researchers at Pacific Northwest National Laboratory (PNNL) are using the test data to validate explicit finite element models to calculate the mechanical loads and structural response of spent nuclear fuel assemblies in the as-tested 30 cm horizontal package drop scenario. Once the as-tested package drop model is validated, the next step is to apply the model to the general 30 cm drop scenario, which includes all impact angles, all fuel assembly types, and all burnup conditions. PNNL is developing a damage model that will incorporate the results of finite element parametric studies to establish trends in SNF mechanical loading to various input parameters, like impact orientation and burnup. The damage model will have the capability to estimate mechanical loads for any single set of input parameters, but it will first be used to describe the upper bounds of potential SNF loading in the 30 cm drop scenario. This paper describes PNNL’s progress toward developing the general solution to the problem of spent nuclear fuel mechanical loads in the 30 cm package drop scenario and it describes the next steps in closing the knowledge gap.

Klymyshyn, Nicholas A.↗
DOE OSTI2017

Gamma-ray mirror technology for NDA of spent fuel

Direct measurements of gamma rays emitted by fissile material have been proposed as an alternative to measurements of the gamma rays from fission products. From a safeguards applications perspective, direct detection of uranium (U) and plutonium (Pu) K-shell fluorescence emission lines and specific lines from some of their isotopes could lead to improved shipper-receiver difference or input accountability at the start of Pu reprocessing. However, these measurements are difficult to implement when the spent fuel is in the line-of-sight of the detector, as the detector is exposed to high rates dominated by fission product emissions. To overcome the combination of high rates and high background, grazing incidence multilayer mirrors have been proposed as a solution to selectively reflect U and Pu hard X-ray and soft gamma rays in the 90 to 420 keV energy into a high-purity germanium (HPGe) detector shielded from the direct line-of-sight of spent fuel. Several groups demonstrated that K-shell fluorescence lines of U and Pu in spent fuel could be detected with Ge detectors. In the field of hard X-ray optics the performance of reflective multilayer coated reflective optics was demonstrated up to 645 keV at the European Synchrotron Radiation Facility. Initial measurements conducted at Oak Ridge National Laboratory with sealed sources and scoping experiments conducted at the ORNL Irradiated Fuels Examination Laboratory (IFEL) with spent nuclear fuel further demonstrated the pass-band properties of multilayer mirrors for reflecting specific emission lines into 1D and 2D HPGe detectors, respectively.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Structural Health Monitoring of Nuclear Spent Fuel Storage Facilities

Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. To ensure that nuclear power remains clean energy, monitoring has been identified by DOE as a high priority cross-cutting need, necessary to determine and predict the degradation state of the systems, structures, and components (SSCs) important to safety (ITS). Therefore, nondestructive structural condition monitoring becomes a need to be installed on existing or to be integrated into future storage system to quantify the state of health or to guarantee the safe operation of nuclear power plants (NPPs) during their extended life span. In this project, the lead university and the collaborating national laboratory teamed to develop a nuclear structural health monitoring (n-SHM) system based on in-situ piezoelectric sensing technologies that can monitor structural degradation and aging for nuclear spent fuel DCSS and similar structures. We also aimed to identify and quantify possible influences of nuclear spent fuel environment (temperature and radiation) to the piezoelectric sensor system and come up with adequate solutions and guidelines therefore. We have therefore developed analytical model for piezoelectric based n-SHM methods, with considerations of temperature and irradiation influence on the model of sensing and algorithms in acoustic emission (AE), guided ultrasonic waves (GUW), and electromechanical impedance spectroscopy (EMIS). On the other side, experimentally the temperature and irradiation influence on the piezoelectric sensors and sensing capabilities were investigated. Both short-term and long-term irradiation investigation with our collaborating national laboratory were performed. Moreover, we developed multi-modal sensing, validated in laboratory setup, and conducted the testing on the We performed multi-modal sensing development, verification and validation tests on very complex structures including a medium-scale vacuum drying chamber and a small-scale mockup canister available for the desired testing. Our work developed the potential candidate for long term structural health monitoring of spent fuel canister through piezoelectric wafer sensors and provided the sensing methodologies based on AE and GUW methodologies. It overall provides an innovative system and methodology for enhancing the safe operation of nuclear power plant. All major accomplishments planned in the original proposal were successfully achieved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Oxyhydroxides on aluminum spent nuclear fuel: formation studies and removal practices to prevent radiolytic gas production

Aluminum-clad nuclear fuel has been at the forefront in research and commercial reactors. After being removed from the reactor, the spent nuclear fuel is placed in large-volume water basins (wet storage) that are constantly monitored for water quality to minimize corrosion/deterioration. The water provides both radiation shielding and cooling for the freshly removed fuel. It is highly desired to have a pathway for fuel to be transitioned to dry storage in sealed casks, so that spent nuclear fuel can be safely transported and dispositioned in a permanent repository, and basin capacity can be made available for temporary wet storage of fresher spent fuel. A challenge for safe, road-ready dry storage is the hydrated aluminum oxides (oxyhydroxides) that are prevalent on aluminum spent nuclear fuel that has been stored under water for years. If water is released from the oxyhydroxides during dry storage, it can create corrosion within the storage canister, and hydrogen can be generated through radiolysis of the oxyhydroxides, causing pressurization and flammability risks. The focus of this project is to develop knowledge of the oxyhydroxide growth and production of gases through radiolysis and to apply empirical models for predicting these phenomena. The expertise developed will be used to formulate conditions and parameters for oxyhydroxide removal/dehydration from fuel cladding through vacuum or forced-air gas drying, to prevent problems during permanent dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

High Fidelity Forward Model Development for Nuclear Reactor Spent Fuel Technical Nuclear Forensics

The fidelity of the forward model within a spent fuel forensic analysis system was improved by using two unique methodologies. The first consisted of developing a system to create accurate one-group neutron cross-section libraries for any user specified reactor system. In such, a detailed model is developed using the depletion code MONTEBURNS. During MONTEBURNS execution, cross-section libraries are generated at every user specified burnup step in time. These libraries could be developed for many reactor systems, then housed in a database and used for analyzing unknown fuel samples. The forensic analysis system for spent fuel resulted in higher accuracy at predicting the initial uranium isotopic compositions and burnup from spent fuel samples. Using this method, the error in results was reduced from the order of 1-6% down to less than 1% when recovering a fuel sample's burnup and initial uranium isotopic composition. The second method consisted of implementing 2D/3D reactor depletion codes as the forward model within the system's framework. This method would allow the usage of potentially recoverable geometric information from an unknown sample. No predetermined cross-section library is required for the system using this method, therefore potentially reducing model error associated with the neutron flux spectrum. The accuracy of the recovered initial uranium isotopic compositions and burnup from spent fuel samples was also improved using this method, even more so than the first. For MTR reactors, the error using this method was significantly reduced and was driven to below 0.5%. However, additional research may be required to determine the ideal fission yield and recoverable energy per fission for cases where significant amounts of 239 PU are bred and burned throughout the life of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Potential Research and Development Opportunities for Light Water Reactor Spent Nuclear Fuel at INL

This report documents the commercial spent nuclear fuel currently in storage at Idaho National Laboratory with a specific focus on the potential research and development (R&D) opportunities for this material. Commercial spent nuclear fuel is stored in dry casks on the INL site at the Idaho Nuclear Technology and Engineering Center (INTEC, formerly the Idaho Chemical Processing Plant or ICPP). This spent fuel was part of the DOE-industry cooperative dry cask demonstration project that began in 1983 and from historical fuel testing and treatment activities that had been performed at the Test Area North (TAN)-607 facility. Most of the fuel is “low burnup,” but given that it was removed from the respective reactors between 1970 and 1985, it was representative for its time. The majority of the fuel is of the 15×15 Westinghouse PWR variety, though additional fuel types (e.g., TN-BRP) and configurations are also stored. Potential R&D opportunities identified are (1) the REA-2023 cask, (2) the TN-BRP cask, and (3) the Castor V/21 cask. Fuel from the TN-24P may be an alternative to the intact assemblies in the Castor V/21. Specifically, the REA-2023 cask contains 9 rods that have been exposed to air since 2005. These rods have a detailed pre-loading characterization and thus any changes in corrosion or geometry can most likely be attributed to the storage environment. In addition, similar rods to the nine in REA-2023 reside in the Castor V/21 cask whose environment has been maintained with inert nitrogen. The Castor V/21 cask also contains intact PWR assemblies that could be used to perform limited confirmation of the pyroprocessing approach. An alternative to the use of the Castor assemblies for pyroprocessing is the consolidated fuel that is stored in the TN-24P, which would not require removal from the assembly grid since it has already been packaged in accessible two-piece shell strongback containers. The TN-BRP cask contains a number of Gd-doped, low burnup assemblies that could be of use to validate code packages used to justify burnup-up credit for boiling water reactors (BWRs). The INL dry casks are stored on the CPP-2707 pad at INTEC, adjacent to the CPP-603 Irradiated Fuel Storage Facility, which incorporates a dry hot cell that has the potential to receive a full scale commercial cask of the types identified. Several evaluations have been made regarding this task, including full scale mockup testing of the physical clearances required to remove and handle fuel assemblies and components. A proposed schedule of 14 months and cost of $5.3 million was the outcome of analysis performed by Fluor Idaho and is included as Appendix A & B. In addition to the spent fuel at INL, there is also a full-scale mockup assembly that was fabricated for the Protypical Consolidation Demonstration Project to validate fuel rod consolidation approaches. The assembly was fabricated by Westinghouse as a 15 x 15 PWR, using Zircaloy-4 tubing and steel end fittings in accordance with product specifications. The only departure from specification is the use of copper metal in lieu of uranium oxide pellets to achieve a weight that is within 10% of the standard.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Review of Experimental Assay Data for PWR Spent Fuel

The primary objective of this study is to review existing experimental assay data for pressurized water reactor (PWR) spent fuel, identify gaps in existing data, and asses the potential value of new isotopic measurements for nuclides relevant to burnup credit, shielding, and decay heat validation. The impact of adding new, high-quality experimental assay data on uncertainties associated with PWR depletion validation is discussed. In particular, the discussion focuses on potential benefits of additional isotopic measurements for spent fuel rods available at Oak Ridge National Laboratory. Selected measurements for these spent fuel rods are ongoing in Fiscal Year 2019 (FY19) as funded by the US Department of Energy (DOE) under the Spent Fuel and Waste Science and Technology (SFWST) Program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Maintaining Continuity of Knowledge of Spent Fuel Pools: Field Testing (Final Report)

Maintaining Continuity of Knowledge (CoK) of spent fuel throughout its life cycle is a key objective of the International Atomic Energy Agency (IAEA). This report describes a project in which two specific technologies, cameras with infrared (IR) imaging capabilities and imaging sonar, were tested to determine their effectiveness in augmenting optical surveillance cameras currently used for maintaining CoK for spent nuclear fuel (SNF) staged in spent fuel pools. Furthermore, the project identifies any modifications required to the technologies to facilitate deployment for international safeguards purposes. This report presents results for the imaging sonar tested at Sandia National Laboratories (SNL) in a mockup spent nuclear fuel pool and for the IR-capable cameras tested by Pacific Northwest National Laboratory (PNNL) in the Oregon State University Training, Research, Isotopes, General Atomics (TRIGA) research reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MELCOR Input Model for Spent Fuel Transportation and Storage Canister

This report documents the progress and current results of the MELCOR spent fuel cask input model. The MELCOR model is being developed to investigate aerosol transport and deposition given the aerosol physical phenomena models within MELCOR. To perform the analyses, a general portrayal of the MAGNASTOR® cask system has been employed; however, this system was selected based on available information to provide a reasonable representation of a spent fuel cask. The analytical results are not intended to characterize the performance of the MAGNASTOR® cask. Instead, the provided results are intended to enhance our general understanding of the aerosol behavior within casks and the validity of current models. The current model efforts are being performed to investigate hypothetical UO 2 release from failed fuel pins within a spent fuel cask. The existing MELCOR model of the MAGNASTOR® cask system has been adapted to permit future comparative analyses with the GOTHIC representation of the MAGNASTOR® cask. To support this comparison, the PNNL model characteristics that are unrelated to the aerosol modeling were applied to the MELCOR model. These characteristics included improved comparability of the axial fidelity, total spent fuel power, fuel pin axial power profile, and heat losses from cannister. The thermal-hydraulic solutions are improved within the capability of the MELCOR code and will permit better overall agreement with the GOTHIC results. Detailed results are presented on the thermal-hydraulic analysis of the MELCOR cask as well as characterization of UO 2 aerosol dispersion and deposition within the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗