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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.

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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A Preliminary Analysis on the Applicability of Deep Borehole Disposal for Spent Fuels in Korea - 15496

A deep geological disposal system, the disposal depth is about 500 m below ground, is considered as the safest method to isolate the spent fuels (SF) or high-level radioactive waste from the human environment with the best available technology at present time. The disposal safety of this system has been demonstrated with underground research laboratory and some advanced countries such as Finland and Sweden are implementing their disposal project on commercial stage. However, if these high-level radioactive wastes can be disposed of in deeper and more stable rock formation than deep geological disposal depth, it has several advantages. Therefore, as an alternative disposal concept, i.e., deep borehole disposal technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept for deep borehole disposal of spent fuels or high level radioactive wastes was reviewed. And the key technologies and challenges in development of this disposal method with the nuclear environment were analyzed. Also, very preliminary deep borehole disposal concept including disposal canister concept was developed according to the nuclear environment. These results will be used as an input for the analyses of applicability for DBD in Korea. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Spent Nuclear Fuel Vibration Integrity Study - 16332

The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: - SNF system interface bonding plays an important role in SNF vibration performance, - Fuel structure contributes to the SNF system stiffness, - There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and - SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Media Effects on the Recycling of Rhodium from Spent Nuclear Fuel - 16364

As byproducts of U-235 fission, precious metals (e.g., Ag, Pd, Ru, Rh) are produced in quantities greater than naturally occurring ores. The complex nature of spent nuclear fuel hinders the application of classical methods for this precious metal recovery. Rhodium has particular interest in energy applications and increasing demand for this metal has led to the exploration of alternative resources. To evaluate the recovery of rhodium from spent nuclear fuel, a benzoyl-thiourea ligand was examined to determine the efficiency of rhodium partitioning in liquid-liquid extraction. The effects of nitrate and chloride media were studied to understand the influence of these anions on rhodium speciation and extraction. Near quantitative extraction of rhodium was observed from both nitrate and chloride media at long equilibration times. However, slope analysis experiments indicated different extracted species based on the anion present. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

The UK's National Policy and Programme for the Management of Spent Fuel and Radioactive Waste - 16574

This paper provides a broad overview of the United Kingdom's (UK's) national policies and national programme for the management of spent fuel and radioactive waste. The aim of the paper is to ensure a clear understanding of the approach taken in the UK so that opportunities for learning lessons and joint working, where they exist, can be identified and to ensure a general understanding of the interdependencies between the various approaches taken in the UK. With a legacy that dates back to the late 1940's/early 1950's the UK have a range of complex and difficult issues to resolve that have evolved over time. What was considered to be acceptable in the past is not thought of in the same way today. We therefore need to be forward thinking, innovative and above all ready to adapt our practices and policies to address previously unknown issues and complications. Due to the wide range of types of radioactive wastes and materials to be managed in the UK we have needed to establish a set of high level principles to cover all radioactive waste types while at the same time ensure that specific approaches for specific issues are developed within that framework. In the UK we have also had to consider the issue of integration to ensure that an approach that addresses one issue does not inadvertently cause a problem elsewhere - for example: in disposing of the high volumes of very low level waste we needed an approach to ensure that other low level waste has a disposal route. Financing of decommissioning and future waste management is clearly one of the biggest challenges to be addressed. As mentioned above ensuring that the legacy left for future generation is not one of a burden - we need to ensure that measures are in place to fund and manage long lived radioactive wastes. Finally, it is clear that not all radioactive wastes are the same and therefore different approaches are needed for managing the different types of material. This requires a proportionate and prioritised approach so while the same ultimate goal is applied (i.e.: the safe and responsible management of radioactive waste) some radioactive wastes are more equal than others and need to be treated accordingly. In the UK we believe that we have a broad programme in place that allows the right balance to be struck in managing all our radioactive waste and spent fuel in a proportionate, safe and responsible manner.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Application of peaking factor credit in spent fuel pool criticality analysis

Higher moderator temperature (lower density) and higher fuel temperatures used in the depletion analysis of spent fuel increases the reactivity. Modern criticality analyses for spent fuel pools typically determine the highest power assembly and use the temperatures from this high power assembly. The high power assembly is determined by finding the maximum assembly averaged relative power or peaking factor. For the depletion analysis it is appropriate to use the average peaking factor over the life of the assembly. This can be determined by dividing the assembly burnup by the sum of the core cycle burnups for the cycles the assembly was in the core. A bounding assembly averaged peaking factor is typically about 1.4. Clearly this is 40% higher than the average assembly power. The reactivity limiting fuel assembly generally has low burnup and thus a low peaking factor. Rather than assume a limiting peaking factor to determine the temperatures in the depletion analysis, the depletion analysis can be performed multiple times with different temperature sets so the depleted isotopic content is a function of the peaking factor. This would allow more fuel assemblies to satisfy the loading criteria while maintaining the same criticality margin. The peaking factor credit can be about 2% in k or about 3.5 GWd/MTU change in the burnup requirements. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Cosmic Ray Muon Radiography of Spent Fuel in Dry Storage Casks

In the absence of a permanent repository or reprocessing facilities, highly radioactive spent nuclear fuel bundles are typically placed in long-term storage in above ground casks. These dry storage casks are heavily shielded to prevent the escape of radiation from the contents to the environment. Casks under international safeguards are monitored to detect diversion of spent fuel. The heavy shielding of the cask walls also precludes inspection of cask contents with typical radiographic probes. When international safeguards inspectors need to verify cask contents (which may be necessary to recover from a loss of continuity of knowledge), the casks must be moved back into a cooling pool to be safely opened and visually inspected. This time-consuming procedure is invasive and potentially dangerous. A non-destructive radiographic method to verify a cask's contents in situ would therefore be a useful safeguards tool. Highly energetic cosmic ray muons are an alternative tomographic probe. Since muons are colorless leptons, they do not participate in strong interactions with nucleons, and their large mass suppresses radiative energy loss while passing through materials. These properties allow muons to penetrate dense shielding that defeats more typical probes like photons and neutrons. Measurements of the scattering of muons that pass through objects provide information on the object's internal structure. This summary describes the first measurements using cosmic ray muons to inspect the contents of a partially loaded dry storage cask. We show that partial defects in the cask loading can be identified with this technique. This data shows that cosmic ray muons are an effective radiographic probe of dry storage casks. This technique may be useful for international safeguards inspectors wishing to verify a cask's contents without opening it. The analysis techniques described here allow the column of missing fuel bundles to be determined, but not the row. Recording data from several positions around the cask would allow ambiguities in x-y position to be resolved. Alternatively, a detector could be designed such that it surrounds the cask and provides a full image of the casks interior. Improved measurements and the design of optimized cask imaging instruments are being explored. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

New Approach to Eccentric Positioning of Fuel Assemblies in a Spent Fuel Pool

Criticality analyses of a spent fuel pool (SFP) typically assume the fuel assemblies are centered in the cells. In reality fuel assemblies are not constrained to be centered but can be placed anywhere within the available space. For SFPs with absorber panels it has been shown that the most reactive condition is with the fuel assemblies centered. However, for SFPs without absorber panels eccentric positioning of assemblies increases reactivity. Historically, it was common practice to analyze eccentric positioning by modeling a 2x2 arrangement of the fuel where each assembly is positioned toward the center (with reflective or periodic bounding conditions). The delta k between the eccentric model and the center model was taken as an uncertainty which was then statistically combined with other uncertainties. This historical approach had two problems. First, arrangement of more than 4 assemblies placed toward the center of the model produced higher k's. Second, in a large pool it was likely that 4 assemblies collocated toward each other would exist so asymmetry should be treated as a bias rather than an uncertainty. The number of collocated assemblies is important to the magnitude of the effect. Since spent fuel pools are large it is expected that some collocated eccentrically placed assemblies will occur so the eccentric placement reactivity should be accounted for by a bias. The magnitude of this bias depends greatly on the number of eccentrically placed assemblies that are collocated. A simple probability model assuming random placement can be used to establish the number of collated eccentric assemblies. It has been shown that for a SFP with 800 cells there is less than a 5% probability that more than 16 eccentric assemblies will be collocated. The eccentricity bias using 16 eccentric assemblies collocated is small (on the order of 0.001 in delta k or less) and should be easy to add as a bias in the SFP criticality analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Understanding Public Support for Transportation of Commercial Spent Nuclear Fuel - 19205

Commercially generated spent nuclear fuel (CSNF) in the US is currently stored in cooling pools and dry storage systems at or near nuclear power plants. These cooling pools and dry storage systems provide only a temporary solution, and as CSNF continues to accumulate an integrated program for transport and permanent disposal will need to be developed. Efforts to site the necessary facilities will require that we consider the transportation of spent fuel from reactor sites across the country to one or more centralized locations. This will likely generate interest and possible resistance among US residents who live along potential transportation routes. In this study, we use a recent survey of US adults to explore the nature of public support/opposition to CSNF transportation and the primary drivers of support/opposition such as risk perceptions. We also explore factors that may attenuate the effect of these drivers on support/opposition, such as public trust in agencies that are involved in CSNF management. (authors)

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

The TRANSCEND Consortium - In-situ Identification of Surface Corrosion Products on Spent Nuclear Fuels - 20276

The management of spent nuclear fuel is a major ongoing concern for the UK owing to the cessation of reprocessing operations at Sellafield and the large, complex inventory arising from Magnox, AGR, PWR and prototype reactors. Retrieval and relocation operations for legacy fuels are imminent and therefore, any models that enhance our understanding of fuel evolution will help mitigate the risks associated with fuel storage and disposal. The TRANSCEND Consortium on nuclear waste management comprises four work packages, within this current paper we provide a summary overview of progress to date and illustrative results from Theme 3: Spent Nuclear Fuels. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Impact of Americium-241 (n,γ) Branching Ratio on SFR Core Reactivity and Spent Fuel Characteristics

An accurate prediction of core physics and fuel cycle parameters largely depends on the order of details and accuracy in nuclear data taken into account for actual calculations. {sup 241}Am is a major gateway nuclide for most of the minor actinides and thus an important nuclide for core physics and fuel-cycle calculations. The {sup 241}Am(n,γ) branching ratio (BR) is in fact energy dependent, therefore, it is necessary to taken into account the spectrum effect on the calculation of the average BR for the full-core depletion calculations. Moreover, the accuracy of the BR used in the depletion calculations could significantly influence the core physics performance and post irradiated fuel compositions. The BR of {sup 241}Am(n,γ) in the ENDF/B-VII.0 library is relatively small and flat in the thermal energy range, gradually increases within the intermediate energy range, and even becomes larger in the fast energy range. This indicates that the properly collapsed BR for fast reactors could be significantly different from that of thermal reactors. The evaluated BRs also differ from one evaluation to another. As seen in Table I, average BRs for several evaluated libraries calculated by means of a fast spectrum are similar but have some differences. While some of widely available and used depletion codes, such as ORIGEN in SCALE, have a capability to calculate a variable BR based on the system neutron spectrum, most of the currently available depletion codes still use a pre-determined single value BR for each library. However, ideally it should be determined on an on-the-fly basis like that of one-group cross sections. These issues provide a strong incentive to investigate the effect of different {sup 241}Am(n,γ) BRs on core and spent fuel parameters. This paper investigates the impact of the {sup 241}Am(n,γ) BR on the results of sodium-cooled fast reactor (SFR) full-core based fuel-cycle calculations. The analysis is performed by gradually increasing the value of BR from 0.15 to 0.25 and studying its impact on the core reactivity and characteristics of SFR spent fuels over extended storage times (∼10,000 years). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Ultra-Long-Life Fast Reactor Cores having Axial Blanket-Driver-Blanket Burnup Strategy with Thorium and PWR Spent Fuels

The sustainable energy over a long-term period can be supplied by ultra-long-life fast reactor cores that are designed by fully utilizing breeding by fast neutrons to be operated over several tens of years without refueling. But the fast reactor cores of this type usually have high positive coolant void reactivity worth resulted from the high TRU (Transuranics) contents in the high burnup, which can degrade the inherent safety through high positive coolant expansion reactivity coefficient. Recently, the authors have designed new small ultra-long-life cores that can be operated over 50{approx}60 EFPYs (Effective Full Power Years) and satisfy the self-controllability conditions under the unprotected accidents. The cores use axial blanket-driver-blanket configuration to achieve ultra-long operational cycle length where the fuel burnup propagates from the outer core to the inner core and from the axially central driver to the outer blanket regions. The objective of this work is to suggest the different design options with thorium addition and TRU and uranium from PWR spent fuels for effectively utilizing the large PWR spent fuel stocks and to analyze the performances of the cores. (authors)

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

Nuclear data uncertainty quantification for the nuclide inventory of a Calvert Cliffs spent fuel sample

The impact of nuclear data cross section uncertainties and covariance matrices on the nuclide vector of spent nuclear fuel was investigated; This exercise was carried out for the Calvert Cliffs fuel assembly D047 benchmark available in the SFCOMPO database. Sample P irradiated in rod MKP109 for 4 cycles up to a burnup of approximately 44 GWd/MTU was selected for the analysis. Nuclear data uncertainties were taken from the most recent libraries released by evaluation projects JEFF, ENDF/B and JENDL, and were propagated using the SANDY code via a stochastic sampling approach. This paper provides a quantification of the uncertainty on the concentration of several actinides and fission products relevant for spent fuel management. Uncertainties generally below 5 % were predicted for the concentrations of most of the uranium, neptunium and plutonium isotopes relevant for SNF applications. Curium isotopes carry larger uncertainties that might exceed 10 %. The contribution of cross section uncertainties on the concentrations of fission products was found to be marginal with the exception of a few nuclides. These results can be significantly affected by the lack of evaluated covariance matrices for the capture cross section of several fission products. Burnup tracers such as {sup 148}Nd and {sup 137}Cs have negligible uncertainties because of the power normalisation imposed in every stochastic calculation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

[Presentation] Long time scale Multiphysics simulation of spent nuclear fuel canister in MOOSE

Pebble-bed reactors are an important class of advanced reactors under consideration for various applications where their fuel would give a significant advantage in siting and high-quality heat production. However, the disposal of their fuel is not as thoroughly studied as other fuel forms. This article provides an example of evaluating advanced reactor spent nuclear fuel in MOOSE. In this study, pebble fuel is analyzed in a well-known spent fuel canister design to characterize the behavior of this fuel form over a long time scale. The results indicate that after approximately 100 years, decay heat is significantly reduced and the maximum temperature in the canister equalizes with the external temperature. The simulation goes on to an end time of one million years, demonstrating the capability of efficiently dealing with long time scales efficiently. We conclude that the canister temperatures seem manageable even with very aggressive loading times and while there are several improvements to be implemented in the future, MOOSE is currently capable of simulating the required scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Reference Fuel Development for Non-Aluminum Spent Nuclear Fuel Management

The Savannah River Site (SRS) L Area Facility provides for the safe receipt, storage, handling, and shipping of spent nuclear fuel (SNF) and has received more than 47,000 SNF assemblies since 1964. In order to consolidate receipt and storage analysis, L Area criticality safety has historically used several aluminum-clad “reference fuels” to establish bounding storage, handling, and cask loading limits and then applied a reactivity comparison approach to demonstrate that candidate fuels may be processed under the reference fuel limits. Currently, only aluminum-clad SNF from off-site research reactors are transferred on-site. The Accelerated Basin De-Inventory (ABD) program will begin the removal of bundled Non-Aluminum Spent Nuclear Fuel (NASNF) from the L Area disassembly basin for dissolution and disposition, which has driven the need for a non-aluminum reference fuel. This paper discusses the process and results of creating a new fictional homogenous NASNF highly enriched uranium reference fuel, “MITZ,” to be used in nuclear criticality safety evaluations for upcoming SNF disposition operations. Data demonstrating the relationship between neutron multiplication behavior and fuel spacing is generated for the new reference fuel MITZ as well as existing reference fuels representing several types of SNF assemblies.

Reference Fuel↗
DOE OSTI2025

Reference Fuel Development for Non-Aluminum Spent Nuclear Fuel Management

The Savannah River Site (SRS) L Area Facility provides for the safe receipt, storage, handling, and shipping of spent nuclear fuel (SNF) and has received more than 47,000 SNF assemblies since 1964. In order to consolidate receipt and storage analysis, L Area criticality safety has historically used several aluminum-clad “reference fuels” to establish bounding storage, handling, and cask loading limits and then applied a reactivity comparison approach to demonstrate that candidate fuels may be processed under the reference fuel limits. Currently, only aluminum-clad SNF from off-site research reactors are transferred on-site. The Accelerated Basin De-Inventory (ABD) program will begin the removal of bundled Non-Aluminum Spent Nuclear Fuel (NASNF) from the L Area disassembly basin for dissolution and disposition, which has driven the need for a non-aluminum reference fuel. This paper discusses the process and results of creating a new fictional homogenous NASNF highly enriched uranium reference fuel, “MITZ,” to be used in nuclear criticality safety evaluations for upcoming SNF disposition operations. Data demonstrating the relationship between neutron multiplication behavior and fuel spacing is generated for the new reference fuel MITZ as well as existing reference fuels representing several types of SNF assemblies.

Reference Fuel↗
DOE OSTI2018

High burn-up spent nuclear fuel transport reliability investigation

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which has provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗