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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 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Separation of the rare-earth fission product poisons from spent nuclear fuel

A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Experimental Test Plan for PWR Sister Rods in the High Burnup Spent Fuel Data Project

The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.

dry storage↗
DOE OSTI2017

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project team is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component sizing, and timing of events. These alternative options arise due to technological, economic, or policy considerations. As new developments regularly emerge, the operational concepts will be periodically updated. This paper gives an overview of the different potential alternatives identified in the Concept of Operations document at a conceptual level.

Cumberland, Riley M↗
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. 241Am is a major gateway nuclide for most of minor actinides and thus important nuclide for core physics and fuel-cycle calculations. The 241Am(n,?) branching ratio (BR) is in fact the energy dependent (see Fig. 1), 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 241Am(n,?) in ENDF/B-VII.0 library is relatively small and flat in thermal energy range, gradually increases within the intermediate energy range, and even becomes larger at 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 are 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. Most of currently available depletion codes use a pre-determined single value BR for each library. However, ideally it should be determined on-the-fly basis like that of one-group cross sections. These issues provide a strong incentive to investigate the effect of different 241Am(n,?) BRs on core and spent fuel parameters. This paper investigates the impact of the 241Am(n,?) BR on the results of 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).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Spent Nuclear Fuel Canister Capacity: Have the Limits Been Reached?

The commercial US nuclear utilities have been transferring spent nuclear fuel (SNF) from spent fuel pools (SFPs) to dry storage systems since 1986. SFPs were initially constructed for temporary storage, but without SNF reprocessing options in the United States and with no final disposal pathway yet available,SFPs started to reach capacity. Hence, to allow continued operation of the nation’s commercial nuclearreactor fleet, utilities started transferring SNF from SFPs (wet storage) to dry storage. Dry storage systems have evolved from storage-only systems first loaded in 1986 to the high-capacity dual-purpose storage and transport canisters used today. A question has arisen as to whether maximum canister capacity has been reached at the current capacity of ~50 kW heat load for storage, with 37-assembly pressurized water reactor and 89-assembly boiling water reactor canisters. This paper presents a holistic overview of commercial US SNF management practices over the years, the current SFP status in terms of filled (occupied) and empty (available) locations, the expected future reactor discharges, and the decay heat load distribution of SFP inventories as of June 2013. Finally, the paper makes a case that further increase in dry storage system capacity both in terms of number of assemblies and decay heat (1) is not needed, and (2) may not be practical from the cost-benefit perspective. b

Banerjee, Kaushik↗
DOE OSTI2017

Calculations vs. Measurements for Remnant Dose Rates from SNS Spent Structures

The Spallation Neutron Source (SNS) in Oak Ridge, Tennessee, is an accelerator driven neutron scattering facility for materials research. Presently SNS is capable to operate at 1.4 MW proton beam power incident on a mercury target with a proton beam energy of 1 GeV and 60 Hz repetition rate. SNS target system components are periodically replaced because they reach their end-of-life due to radiation induced material damage. Target vessel, which houses mercury target, is exchanged about two-three times per year and the proton beam window (PBW) is exchanged every two – three years.Each spent structure that leaves the SNS site requires supporting documentation with radionuclide inventory and dose rate prediction for the time of the transportation. Neutronics analyses are performed, assuming realistic irradiation history and decay case to ensure that the container/package, housing the structure, is compliant with the waste management regulations. Analyses are complex due to geometry, multi-code usage and following data treatment.To validate analyses, measurements of dose rates from the spent target vessel # 13 and PBW module #5 were performed. Neutronics analyses were performed to calculate residual dose rates from both structures for the time of measurements.

Popova, Irina I.↗
DOE OSTI2018

Modeling Passive Fast-Neutron Emission Tomography of Spent Nuclear Fuel

For the past year Oak Ridge National Laboratory has been developing a passive fast-neutron emission tomography capability. Like the passive gamma emission tomography capability presently being evaluated by the International Atomic Energy Agency, the goal of this capability is to detect the removal or substitution of individual fuel pins in spent nuclear fuel assemblies for international safeguards applications, such as verifying the integrity of an assembly before transfer to difficult-to-access storage. The present imaging technique is based on neutron emissions originating primarily from curium-244, which is produced predominantly at the end of the exposure cycle. As a result, this technique may be sensitive not only to fuel pins that are removed or substituted subsequent to all irradiation, but also to fuel pins that are substituted and subsequently irradiated. At present, a laboratory prototype imager is under construction. The purpose of the prototype is to demonstrate an imaging capability sufficient to resolve individual fuel pins using commercially available boron straw detectors that can withstand the high gamma ray dose from a typical spent fuel assembly. To achieve both practical measurement durations and sufficient resolution to discern individual pins, the prototype imager employs a novel collimator design where the slits of a parallel-slit collimator are rearranged to distribute them around a ring surrounding the target fuel assembly. This presentation will report on simulations to understand the neutron signal-to-noise ratio and gamma dose rate as a function of the collimator design, the neutron detector’s response to gamma rays as a function of dose rate, and image reconstruction challenges associated with this collimator design.

Hausladen, Paul↗
DOE OSTI2019

A Misload Analysis Methodology Supporting Criticality Analysis of Spent Nuclear Fuel Canisters Using As-Loaded Configurations

This paper presents an assembly misload analysis methodology developed to support criticality safety analysis of spent nuclear fuel (SNF) dual-purpose canisters (DPCs) using as-loaded configurations. The misload analysis approach is based on Interim Staff Guidance 8 “Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks” (ISG 8 rev. 3), but extended to support as-loaded configurations. The two misload scenarios analyzed included; (1) the correct assemblies are selected from the pool but placed incorrectly into the most reactive configuration inside the canister, and (2) the incorrect, most reactive assembly/assemblies in the pool is/are placed into the most reactive position(s) in the canister. Results from this misload analysis approach can be combined with misload probability to support criticality safety assessment of SNF during storage, transportation, and disposal. The misload analysis has been applied to 67 loaded DPCs at two sites using a disposal scenario. The analyzed sites include both PWR and BWR reactors and two canister variants—MPC-32 and MPC-68.

Liljenfeldt, Henrik↗
DOE OSTI2012

Delayed Gamma-Ray Spectroscopy for Spent Nuclear Fuel Assay

High-energy, beta-delayed gamma-ray spectroscopy is investigated as a nondestructive assay technique for the determination of plutonium mass in spent nuclear fuel. This approach exploits the unique isotope-specific signatures contained in the delayed gamma-ray emission spectra detected following active interrogation with an external neutron source. A high fidelity modeling approach that couples radiation transport, analytical decay/depletion, and a newly developed gamma-ray emission source reconstruction code is described. Initially simulated and analyzed was a “one-pass” delayed gamma-ray assay that focused on the long-lived signatures. Also presented are the results of an independent study that investigated “pulsed mode” measurements to capture the more isotope-specific, short-lived signatures. The initial modeling results outlined in this paper suggest that delayed gamma-ray assay of spent nuclear fuel assemblies can be accomplished with a neutron generator of sufficient strength and currently available gamma-ray detectors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

CISCC Countermeasures for Spent Fuel Canisters in Concrete Casks

Aging management of spent fuel is an important issue worldwide. Chloride-induced stress corrosion cracking (CISCC) of welded canisters is particularly important for concrete casks, where it occurs in response to a combination of environmental, material, and residual stress factors. In this paper, we introduce a countermeasure to prevent the occurrence of CISCC by adjusting the flow rate of the cooling air into the concrete cask so that the canister surface temperature does not decrease below 100°C, which is a conservative deliquesce temperature threshold for CISCC occurrence. The flow adjustment device was installed in a cask model, and two tests were conducted under the heat rate conditions corresponding to those of an actual canister. In the first test, the device was operated under a condition where the canister surface temperature was below 100°C to verify whether it could increase the canister surface temperature to above 100°C. In the second test, the heat rate was decreased from a state where the canister surface temperature was above 100°C, to verify whether the device could maintain the canister surface temperature above 100°C. Good results were obtained in both cases. However, in the case of a low decay heat rate of spent fuel after extended dry storage, the temperature of the canister surface may decrease below 100°C even if no air supply is provided. Therefore, in this case, it may be necessary to take measures such as blocking air outlets.

Takeda, Hirofumi↗
DOE OSTI2022

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

The United States Department of Energy (DOE) has long term goals to develop solutions for managing the nation’s spent nuclear fuel (SNF) and high-level waste (HLW) inventory. The Integrated Waste Management (IWM) program under the DOE office of Nuclear Energy (DOE-NE) is employing system-level engineering and analysis principles to inform potential future waste management system architectures. Managing the spent nuclear waste requires the use of system-level analysis software that takes various aspects of the fuel cycle into account like waste generation, on-site/centralized storage, transportation infrastructure, and long-term disposal. The Next Generation System Analysis Model (NGSAM) is an agent-based model that was developed to simulate the transportation and storage of SNF and HLW. As an agent-based model, NGSAM has the capability to detail the interaction and movement of individual components and groups, such as rail cars and casks. The SNF inventory from commercial nuclear reactors is currently in temporary storage at multiple locations spread across the United States. Shipping of SNF from these locations relies on one of three transportation modes: rail, heavy-haul truck, or barge. Out of the three modes identified, rail is generally the most preferred due to the size of the canisters and casks the SNF would be shipped in. However, under some scenarios, a direct rail route might not be readily available to a reactor site or improving the rail infrastructure at shutdown sites might be too cost-prohibitive for utilities to opt for a direct rail transfer. Under such scenarios, using a barge or heavy haul truck to de-inventory the site and transfer the SNF to a nearby intermodal transfer site with adequate rail infrastructure where the payload could be transferred to a rail car might prove to be an attractive option. This work initially presents the various intermodal transportation options that could be used to transfer SNF from reactor sites to rail cars. This is followed by exploring the operational steps in each of these modes to move the SNF from a reactor site and transfer it to a rail car. This work also presents the procedure of implementing the intermodal transfer methodology in NGSAM using various Java methods. Finally, the process times for accomplishing each of the individual steps are furnished. The implementation ideology, assumptions, and future steps are presented in this work.

Gadey, Harish Reddy↗
DOE OSTI2013

Experience and Lessons Learned from Conditioning of Spent Sealed Sources in Singapore - 13107

In 2010, IAEA requested KAERI (Korea Atomic Energy Research Institute) to support Singapore for conditioning spent sealed sources. Those that had been used for a lightning conductor, check source, or smoke detector, various sealed sources had been collected and stored by the NEA (National Environment Agency) in Singapore. Based on experiences for the conditioning of Ra-226 sources in some Asian countries since 2000, KAERI sent an expert team to Singapore for the safe management of spent sealed sources in 2011. As a result of the conditioning, about 575.21 mCi of Am-241, Ra-226, Co-60, and Sr-90 were safely conditioned in 3 concrete lining drums with the cooperation of the KAERI expert team, the IAEA supervisor, the NEA staff and local laborers in Singapore. Some lessons were learned during the operation: (1) preparations by a local authority are very helpful for an efficient operation, (2) a preliminary inspection by an expert team is helpful for the operation, (3) brief reports before and after daily operation are useful for communication, and (4) a training opportunity is required for the sustainability of the expert team. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Conditioning and Repackaging of Spent Radioactive Cs-137 and Co-60 Sealed Sources in Egypt - 13490

Radioactive Sealed sources (RSSs) are widely use all over the world in medicine, agriculture, industry, research, etc. The accidental misuse and exposure to RSSs has caused significant environmental contamination, serious injuries and many deaths. The high specific activity of the materials in many RSSs means that the spread of as little as microgram quantities can generate significant risk to human health and inhibit the use of buildings and land. Conditioning of such sources is a must to protect humans and environment from the hazard of ionizing radiation and contamination. Conditioning is also increase the security of these sources by decreasing the probability of stolen and/or use in terrorist attacks. According to the law No.7/2010, Egyptian atomic energy authority represented in the hot laboratories and waste management center (centralized waste facility, HLWMC) has the responsibility of collecting, conditioning, storing and management of all types of radioactive waste from all Egyptian territory including spent radioactive sealed sources (SRSSs). This paper explains the conditioning procedures for two of the most common SRSSs, Cs{sup 137} and Co{sup 60} sources which make up more than 90% of the total spent radioactive sealed sources stored in our centralized waste facility as one of the major activities of hot laboratories and waste management center. Conditioning has to meet three main objectives, be acceptable for storage, enable their safe transport, and comply with disposal requirements. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Spent Sealed Sources Management in Switzerland - 12011

Information is provided about the international recommendations for the safe management of disused and spent sealed radioactive sources wherein the return to the supplier or manufacturer is encouraged for large radioactive sources. The legal situation in Switzerland is described mentioning the demand of minimization of radioactive waste as well as the situation with respect to the interim storage facility at the Paul Scherrer Institute (PSI). Based on this information and on the market situation with a shortage of some medical radionuclides the management of spent sealed sources is provided. The sources are sorted according to their activity in relation to the nuclide-specific A2-value and either recycled as in the case of high active sources or conditioned as in the case for sources with lower activity. The results are presented as comparison between recycled and conditioned activity for three selected nuclides, i.e. Cs-137, Co-60 and Am-241. (author)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Management of Spent and Disused Sealed Radioactive Sources in the Czech Republic - 12124

The Czech Republic is a country with a well-developed peaceful utilization of nuclear energy and ionizing radiation. Sealed Radioactive Sources (further also SRS) are broadly used in many areas in the Czech Republic, e.g. in research, industry, medicine, education, agriculture, etc. Legislation in the field of ionizing radiation source utilization has been fully harmonized with European Community legislation. SRS utilization demands a proper system which must ensure the safe use of SRS, including the management of disused (spent) and orphaned SRS. In the Czech Republic, a comprehensive system of SRS management has been established that is comparable with systems in other developed countries. The system covers both legal and institutional aspects. The Central Register of Ionizing Radiation Sources is an important part of the system. It is a tracking system that covers all activities related to SRS, from their production or import to the end of their use (recycling or disposal). Many spent SRS are recycled and can be used for other purposes after inspection, repacking or reprocessing. When the disused SRS are not intended for further use, they are managed as radioactive waste (RAW). The system of SRS management also ensures the suitable resolution of situations connected with improper SRS handling (in the case of orphaned sources, accidents, etc.). (author)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Pyrolysis of Spent Ion Exchange Resins - 12210

Organic ion exchangers (IEX) play a major and increasing role in the reactor coolant and other water purification processes. During their operation time they receive significant amounts of radioactivity, making their disposal, together with their organic nature, as medium active waste challenging. Processes applied so far do not eliminate the organic matter, which is unwanted in disposal facilities, or, if high temperatures are applied, raise problems with volatile radionuclides. NUKEM Technologies offers their well introduces process for the destruction of spent solvent (TBP), the pebble bed pyrolysis, now for the treatment of spent IEX (and other problematic waste), with the following benefits: the pyrolysis product is free of organic matter, and the operation temperature with approx. 500 deg. C keeps Cs radionuclides completely in the solid residue. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Prairie Island Nuclear Station Spent Filter Processing for Direct Disposal - 12333

This paper will discuss WMG's filter processing experience within the commercial nuclear power industry, specifically recent experience processing high activity spent filters generated by Xcel Energy's Prairie Island Nuclear Station (Prairie Island), located in Welch, MN. WMG processed for disposal eighty-four 55-gallon drums filled with varying types of high activity spent filters. The scope of work involved characterization, packaging plan development, transport to the WMG's Off-Site Processing location, shredding the filter contents of each drum, cement solidifying the shredded filter material, and finally shipping the solidified container of shredded filter material to Clive, Utah where the container was presented to EnergySolutions Disposal site for disposal in their Containerised Waste Facility. This sequence of events presented in this paper took place a total of nine (9) times over a period of four weeks. All 1294 filters were successfully solidified into nine (9) -WMG 142 steel liners, and each was successfully disposed of as Class A Waste at EnergySolutions Disposal Site in Clive, Utah. Prairie Island's waste material was unique in that all its filters were packaged in 55-gallon drums; and since the station packaged its filters in drums it was much easier to develop packaging plans for such a large volume of legacy filters. For this author, having over 20-years of waste management experiences, storing and shipping waste material in 55-gallon drums is not immediately thought of as a highly efficient method of managing its waste material. However, Prairie Island's use of 55-gallon drums to store and package its filters provided a significant advantage. Drums could be mixed and matched to provide the most efficient processing method while still meeting the Waste Class A limits required for disposal. (author)

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

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

The United States Department of Energy (DOE) Office of Nuclear Energy is applying knowledge and understanding in the areas of systems level engineering, analysis, and decision making to better inform the waste management pathways for U.S. spent nuclear fuel (SNF). Currently there are 93 operating and 23 shutdown commercial nuclear reactors in the United States. SNF at most of these locations is being stored in spent fuel pools, dry storage, or both [1, 2]. This paper initially goes over the basics of the agent-based simulation tool known as the Next Generation System Analysis Model (NGSAM) that has the capability to model interaction and movements of individual components or groups like casks, railcars etc. The next section covers some of the methods implemented in the Java Transportation Operations Model (JTOM) to facilitate the movement of assemblies, casks, railcars etc. and the various intermodal transfer options. This section also covers the current cask loading and intermodal transfer times implemented in NGSAM. The last section goes over the proposed values for intermodal transportation, and cask transfer/loading operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗