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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Lead Slowing-Down Spectrometry Time Spectral Analysis for Spent Fuel Assay: FY11 Status Report

Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration, of which PNNL is a part, to study the feasibility of Lead Slowing Down Spectroscopy (LSDS). This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10% typical of today's confirmatory assay methods. This document is a progress report for FY2011 PNNL analysis and algorithm development. Progress made by PNNL in FY2011 continues to indicate the promise of LSDS analysis and algorithms applied to used fuel. PNNL developed an empirical model based on calibration of the LSDS to responses generated from well-characterized used fuel. The empirical model, which accounts for self-shielding effects using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the true self-shielding functions of the used fuel assembly models. The potential for the direct and independent assay of the sum of the masses of 239Pu and 241Pu to within approximately 3% over a wide used fuel parameter space was demonstrated. Also, in FY2011, PNNL continued to develop an analytical model. Such efforts included the addition of six more non-fissile absorbers in the analytical shielding function and the non-uniformity of the neutron flux across the LSDS assay chamber. A hybrid analytical-empirical approach was developed to determine the mass of total Pu (sum of the masses of 239Pu, 240Pu, and 241Pu), which is an important quantity in safeguards. Results using this hybrid method were of approximately the same accuracy as the pure empirical approach. In addition, total Pu with much better accuracy with the hybrid approach than the pure analytical approach. In FY2012, PNNL will continue efforts to optimize its empirical model and minimize its reliance on calibration data. In addition, PNNL will continue to develop an analytical model, considering effects such as neutron-scattering in the fuel and cladding, as well as neutrons streaming through gaps between fuel pins in the fuel assembly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Storage and Reprocessing of Spent Nuclear Fuel

Addressing the problem of waste, especially high-level waste (HLW), is a requirement of the nuclear fuel cycle that cannot be ignored. We explore the two options employed currently, long-term storage and reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

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DOE OSTI2016

Integrated Systems Analysis for a Regional Deep Borehole Repository for Commercial Spent Nuclear Fuel - 16291

Some nuclear waste stakeholders have suggested private interim storage facilities (ISF), but little work has been done to consider the cost or logistics for nuclear waste management or disposal facilities smaller than the national scale. This paper performs an integrated system analysis for a deep borehole repository system which would accept waste from a regional subset of the national commercial reactor fleet. The regional subset is comprised of a centrally-located host state with a large SNF inventory surrounded by states with small SNF inventories. The chosen region will contain between one-fifth and one-quarter of the national commercial SNF inventory. Using best estimates of past and future SNF discharges in the sample region, the SNF inventory in the chosen region is characterized by reactor of origin. Basic parameters are then chosen for the regional repository, such as the annual rate of waste emplacement and the time necessary to process borehole-specific waste canisters. Scenarios chosen for analysis test the sensitivity of a variety of system parameters, including the annual rate of waste acceptance, the division of annual waste acceptance among sites, the use of bare fuel transportation canisters, and the size of the regional subset of reactors. Estimated costs for the facility are $13 billion for around 27,000 MT of SNF and $16 billion for around 35,000 MT. This translates to $490,000 per MT and $466,000 per MT, respectively. These unit costs are comparable to the $370,000 per MT estimate for the full 140,000 MT national inventory. (authors)

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DOE OSTI2019

Virtual Simulation of Spent Nuclear Fuel Repackaging Facility Design Concepts - 19479

This paper presents recent implementation of virtual simulation of SNF repackaging facility operation. The development was based on recent wet and dry facility design concepts. The development involved creating equipment and facility models and implementation of discrete event simulation of large-scale facility-level operation. Lastly, it presented how a hardware-in-the-loop simulation is possible and that a good amount of accuracy was retained from virtual simulation to hardware simulation for the tested configuration. It is expected that this type of simulation will act as a proof of concept for system engineers and help them make more informed decisions when considering potential solutions for SNF handling and disposal. (authors)

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DOE OSTI2017

Classification and imaging of spent nuclear fuel dry casks using cosmic ray muons - 132

The present work explores alternative approaches for in situ monitoring with near real-time data analysis and imaging of dry casks using muons. Monitoring can provide useful signals about the cask contents in a yes/no decision format faster and with fewer muons than traditional imaging. Imaging can be performed as a second step in this process if specific information about the location of the fuel assemblies is needed. To explore and analyze these two modalities, simulations of muon scattering and displacement were performed for various scenarios, including vertical and horizontal fully loaded dry casks, half-loaded dry casks, dry casks with one row of fuel assemblies missing, dry casks with one fuel assembly missing, and empty dry casks. This work demonstrates monitoring can provide fast and reliable indication about the contents of the cask. For 100,000 muons, a few minutes of measurement time with a 3.6 m{sup 2} detector at a 45 deg. zenith angle, 90% detection probability for missing fuel assemblies can be achieved with a false alarm rate lower than 1%. If a positive signal is detected, then longer measurements can be performed and coupled with muon imaging algorithms to determine the location of missing nuclear material. In these cases, transmission muon radiography can produce images of acceptable resolution for all scenarios. In the horizontal dry cask case, all scenarios can be identified and separated adequately. In the vertical dry cask case, similar conclusions can be drawn, but because of shadowing from other fuel assemblies, the location of a missing fuel assembly cannot be identified in the 2-D transmission map. However, views from different directions may resolve this. (authors)

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DOE OSTI2025

Fabrication of surrogate oxide spent fuel with various cracking patterns and design of an axial gas transport apparatus

In this article, understanding gas transport behavior in nuclear fuel rods is important for the design, performance, and safety of nuclear fuels. Surrogate materials help enable efficient research by reducing both the costs and the amount of time required. A parametric study using Darcy’s law is completed that demonstrates the feasibility of observing pressure decay over short 13-to-15-cm specimens to enable full characterization of the fabricated specimens using x-ray computed tomography. This paper demonstrates that thermally shocked and mechanically compressed alumina pellets produce surrogate samples whose various cracking patterns are representative of the severity of cracking observed as a function of burnup in irradiated nuclear fuels. Furthermore, image analyses of the cracking patterns—in conjunction with gas transport testing using surrogate samples—affords a valuable accelerated basis for developing gas transport simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Spent Nuclear Fuel Disposition

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNF is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Radioactivity of spent TRIGA fuel

Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Lead Slowing-Down Spectrometry for Spent Fuel Assay: FY11 Status Report

Executive Summary Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration to study the feasibility of Lead Slowing Down Spectroscopy (LSDS). This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10% typical of today’s confirmatory assay methods. This document is a progress report for FY2011 collaboration activities. Progress made by the collaboration in FY2011 continues to indicate the promise of LSDS techniques applied to used fuel. PNNL developed an empirical model based on calibration of the LSDS to responses generated from well-characterized used fuel. The empirical model demonstrated the potential for the direct and independent assay of the sum of the masses of 239Pu and 241Pu to within approximately 3% over a wide used fuel parameter space. Similar results were obtained using a perturbation approach developed by LANL. Benchmark measurements have been successfully conducted at LANL and at RPI using their respective LSDS instruments. The ISU and UNLV collaborative effort is focused on the fabrication and testing of prototype fission chambers lined with ultra-depleted 238U and 232Th, and uranium deposition on a stainless steel disc using spiked U3O8 from room temperature ionic liquid was successful, with improving thickness obtained. In FY2012, the collaboration plans a broad array of activities. PNNL will focus on optimizing its empirical model and minimizing its reliance on calibration data, as well continuing efforts on developing an analytical model. Additional measurements are planned at LANL and RPI. LANL measurements will include a Pu sample, which is expected to provide more counts at longer slowing-down times to help identify discrepancies between experimental data and MCNPX simulations. RPI measurements will include the assay of an entire fresh fuel assembly for the study of self-shielding effects as well as the ability to detect diversion by detecting a missing fuel pin in the fuel assembly. The development of threshold neutron sensors will continue, and UNLV will calibrate existing ultra-depleted uranium deposits at ISU.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Developments of Spent Nuclear Fuel Pyroprocessing Technology at Idaho National Laboratory

This paper summarizes research in used fuel pyroprocessing that has been published by Idaho National Laboratory over the last decade. It includes work done both on treatment of Experimental Breeder Reactor-II and development of advanced technology for potential scale-up and commercialization. Collaborations with universities and other laboratories is included in the cited work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Nondestructive Assay Options for Spent Fuel Encapsulation

This report describes the role that nondestructive assay (NDA) techniques and systems of NDA techniques may have in the context of an encapsulation and deep geological repository. The potential NDA needs of an encapsulation and repository facility include safeguards, heat content, and criticality. Some discussion of the facility needs is given, with the majority of the report concentrating on the capability and characteristics of individual NDA instruments and techniques currently available or under development. Particular emphasis is given to how the NDA techniques can be used to determine the heat production of an assembly, as well as meet the dual safeguards needs of 1) determining the declared parameters of initial enrichment, burn-up, and cooling time and 2) detecting defects (total, partial, and bias). The report concludes with the recommendation of three integrated systems that might meet the combined NDA needs of the encapsulation/repository facility.

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