Method for shearing spent nuclear fuel assemblies

DOEpatents

Weil, Bradley S.; Watson, Clyde D.

1977-01-01

A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

Radiation induced corrosion of copper for spent nuclear fuel storage

NASA Astrophysics Data System (ADS)

Björkbacka, Åsa; Hosseinpour, Saman; Johnson, Magnus; Leygraf, Christofer; Jonsson, Mats

2013-11-01

The long term safety of repositories for radioactive waste is one of the main concerns for countries utilizing nuclear power. The integrity of engineered and natural barriers in such repositories must be carefully evaluated in order to minimize the release of radionuclides to the biosphere. One of the most developed concepts of long term storage of spent nuclear fuel is the Swedish KBS-3 method. According to this method, the spent fuel will be sealed inside copper canisters surrounded by bentonite clay and placed 500 m down in stable bedrock. Despite the importance of the process of radiation induced corrosion of copper, relatively few studies have been reported. In this work the effect of the total gamma dose on radiation induced corrosion of copper in anoxic pure water has been studied experimentally. Copper samples submerged in water were exposed to a series of total doses using three different dose rates. Unirradiated samples were used as reference samples throughout. The copper surfaces were examined qualitatively using IRAS and XPS and quantitatively using cathodic reduction. The concentration of copper in solution after irradiation was measured using ICP-AES. The influence of aqueous radiation chemistry on the corrosion process was evaluated based on numerical simulations. The experiments show that the dissolution as well as the oxide layer thickness increase upon radiation. Interestingly, the evaluation using numerical simulations indicates that aqueous radiation chemistry is not the only process driving the corrosion of copper in these systems.

Proliferation resistance assessment of various methods of spent nuclear fuel storage and disposal

NASA Astrophysics Data System (ADS)

Kollar, Lenka

Many countries are planning to build or already are building new nuclear power plants to match their growing energy needs. Since all nuclear power plants handle nuclear materials that could potentially be converted and used for nuclear weapons, they each present a nuclear proliferation risk. Spent nuclear fuel presents the largest build-up of nuclear material at a power plant. This is a proliferation risk because spent fuel contains plutonium that can be chemically separated and used for a nuclear weapon. The International Atomic Energy Agency (IAEA) safeguards spent fuel in all non-nuclear weapons states that are party to the Non-Proliferation Treaty. Various safeguards methods are in use at nuclear power plants and research is underway to develop safeguards methods for spent fuel in centralized storage or underground storage and disposal. Each method of spent fuel storage presents different proliferation risks due to the nature of the storage method and the safeguards techniques that are utilized. Previous proliferation resistance and proliferation risk assessments have mainly compared nuclear material through the whole fuel cycle and not specifically focused on spent fuel storage. This project evaluates the proliferation resistance of the three main types of spent fuel storage: spent fuel pool, dry cask storage, and geological repository. The proliferation resistance assessment methodology that is used in this project is adopted from previous work and altered to be applicable to spent fuel storage. The assessment methodology utilizes various intrinsic and extrinsic proliferation-resistant attributes for each spent fuel storage type. These attributes are used to calculate a total proliferation resistant (PR) value. The maximum PR value is 1.00 and a greater number means that the facility is more proliferation resistant. Current data for spent fuel storage in the United States and around the world was collected. The PR values obtained from this data are 0.49 for

Update and evaluation of decay data for spent nuclear fuel analyses

NASA Astrophysics Data System (ADS)

Simeonov, Teodosi; Wemple, Charles

2017-09-01

Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

Systems for the Intermodal Routing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Steven K; Liu, Cheng

The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable systemmore » for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the

Rack for storing spent nuclear fuel elements

DOEpatents

Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

1978-06-20

A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

DOEpatents

Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

2013-11-05

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

DOEpatents

Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

2014-06-10

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

DOE Office of Scientific and Technical Information (OSTI.GOV)

N /A

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign andmore » domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.« less

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required tomore » complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.« less

Air Shipment of Spent Nuclear Fuel from Romania to Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Igor Bolshinsky; Ken Allen; Lucian Biro

Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities formore » shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.« less

Study of Compton suppression for use in spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Bender, Sarah

The focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any additional peaks, due to its relatively low resolution. Samples of Approved Test Material (ATM) commercial fuel elements were obtained from Pacific Northwest National Laboratory. The samples had been processed using the beginning stages of the PUREX method and represented the unseparated feed solution from a reprocessing facility. Compton suppressed measurements of the ATM fuel samples were recorded inside the guard detector annulus, to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Photopeak losses were observed in the measurements of the dissolved ATM

Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

NASA Astrophysics Data System (ADS)

Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

2014-08-01

Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

Pyroprocess for processing spent nuclear fuel

DOEpatents

Miller, William E.; Tomczuk, Zygmunt

2002-01-01

This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Rochman, D.; Leray, O.; Vasiliev, A.; Pecchia, M.; Ferroukhi, H.; Caruso, S.

2017-09-01

In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

Spent fuel cask handling at an operating nuclear power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pal, A.C.

1988-01-01

The importance of spent fuel handling at operating nuclear power plants cannot be overstated. Because of its highly radioactive nature, however, spent fuel must be handled in thick, lead-lined containers or casks. Thus, all casks for spent fuel handling are heavy loads by the US Nuclear Regulatory Commission's definition, and any load-drop must be evaluated for its potential to damage safety-related equipment. Nuclear Regulatory Guide NUREG-0612 prescribes the regulatory requirements of alternative heavy-load-handling methodologies such as (a) by providing cranes that meet the requirements of NUREG-0554, which shall be called the soft path, or (b) by providing protective devices atmore » all postulated load-drop areas to prevent any damage to safety-related equipment, which shall be called the hard path. The work reported in this paper relates to cask handling at New York Power Authority's James A. FitzPatrick (JAF) plant.« less

Molten tin reprocessing of spent nuclear fuel elements

DOEpatents

Heckman, Richard A.

1983-01-01

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.

Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additionalmore » microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.« less

Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.

Radioactive iodine is the Achilles’ heel in the design for the safe geological disposal of spent UO2 nuclear fuel. Iodine’s high solubility and anticipated instant release during waste package compromise jeopardize performance assessment calculations. However, dissolution studies have indicated that the instant release fraction (IRF) of radioiodine (I) does not correlate with increasing fuel burn-up. In fact, there is a peak in the release iodine at around 50-60 Mwd/kgU and with increasing burn-up the instant release of iodine decreases. Detailed electron microscopy analysis of high burn-up fuel (~80 MWd/kgU) has revealed the presence of (Pd,Ag)(I,Br) nano-particles. As UO2 fuels aremore » irradiated, the Ag and Pd content increases, from 239Pu fission, enabling radioiodine to be retained. The occurrence of these phases in nuclear fuels may have significant implications for the long-term behavior of iodine.« less

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been storedmore » on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

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 UNFmore » 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).« less

Spent Nuclear Fuel Disposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C.

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 UNFmore » 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).« less

Nanoparticle Precipitation in Irradiated and Annealed Ceria Doped with Metals for Emulation of Spent Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Weilin; Conroy, Michele A.; Kruska, Karen

Epsilon alloy precipitates have been observed with varied compositions and sizes in spent nuclear fuels, such as UO2. Presence of the inclusions, along with other oxide precipitates, gas bubbles and irradiation-induced structural defects, can significantly degrade the physical properties of the fuel. To predict fuel performance, a fundamental study of the precipitation processes is needed. This study uses ceria (CeO2) as a surrogate for UO2. Polycrystalline CeO2 films doped with Mo, Ru, Rh, Pd and Re (surrogate for Tc) were grown at 823 K using pulsed laser deposition, irradiated at 673 K with He+ ions, and subsequently annealed at highermore » temperatures. A number of methods, including transmission electron microscopy and atom probe tomography, were applied to characterize the samples. The results indicate that there is a uniform distribution of the doped metals in the as-grown CeO2 film. Pd particles of ~3 nm in size appear near dislocation edges after He+ ion irradiation to ~13 dpa. Thermal annealing at 1073 K in air leads to formation of precipitates with Mo and Pd around grain boundaries. Further annealing at 1373 K produces 70 nm sized precipitates with small grains at cavities.« less

78 FR 61401 - Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-03

... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-155; 72-43 and NRC-2013-0218] Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory... the Big Rock Point (BRP) Independent Spent Fuel Storage Installation (ISFSI). ADDRESSES: Please refer...

77 FR 37937 - License Renewal Application for Prairie Island Nuclear Generating Plant Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-25

... Prairie Island Nuclear Generating Plant Independent Spent Fuel Storage Installation AGENCY: Nuclear... INFORMATION CONTACT: Pamela Longmire, Ph.D., Project Manager, Licensing Branch, Division of Spent Fuel Storage... February 29, 2012 (ADAMS Accession number ML12065A073), by Prairie Island Nuclear Generating Plant (PINGP...

AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christopher Landers; Igor Bolshinsky; Ken Allen

2010-07-01

In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments weremore » transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.« less

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE PAGES

Hu, Jianwei; Gauld, Ian C.

2014-12-01

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Gauld, Ian C.

The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-07-01

Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horiamore » Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.« less

Preparation for Testing, Safe Packing and Shipping of Spent Nuclear Fuel from IFIN-HH, Bucharest-Magurele to Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dragolici, C.A.; Zorliu, A.; Popa, V.

2007-07-01

The Russian Research Reactor Fuel Return (RRRFR) program is promoted by IAEA and DOE in order to repatriate of irradiated research reactor fuel originally supplied by Russia to facilities outside the country. Developed under the framework of the Global Threat Reduction Initiative (GTRI) the take-back program [1] common goal is to reduce both proliferation and security risks by eliminating or consolidating inventories of high-risk material. The main objective of this program is to support the return to Russian Federation of fresh or irradiated HEU and LEU fuel. Being part of this project, Romania is fulfilling its tasks by examining transportmore » and transfer cask options, assessment of transport routes, and providing cost estimates for required equipment and facility modifications. Spent Nuclear Fuel (SNF) testing, handling, packing and shipping are the most common interests on which the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei' (IFIN-HH) is focusing at the moment. (authors)« less

Hanford Spent Nuclear Fuel Project recommended path forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fulton, J.C.

The Spent Nuclear Fuel Project (the Project), in conjunction with the U.S. Department of Energy-commissioned Independent Technical Assessment (ITA) team, has developed engineered alternatives for expedited removal of spent nuclear fuel, including sludge, from the K Basins at Hanford. These alternatives, along with a foreign processing alternative offered by British Nuclear Fuels Limited (BNFL), were extensively reviewed and evaluated. Based on these evaluations, a Westinghouse Hanford Company (WHC) Recommended Path Forward for K Basins spent nuclear fuel has been developed and is presented in Volume I of this document. The recommendation constitutes an aggressive series of projects to construct andmore » operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. The overall processing and storage scheme is based on the ITA team`s proposed passivation and vault storage process. A dual purpose staging and vault storage facility provides an innovative feature which allows accelerated removal of fuel and sludge from the basins and minimizes programmatic risks beyond any of the originally proposed alternatives. The projects fit within a regulatory and National Environmental Policy Act (NEPA) overlay which mandates a two-phased approach to construction and operation of the needed facilities. The two-phase strategy packages and moves K Basins fuel and sludge to a newly constructed Staging and Storage Facility by the year 2000 where it is staged for processing. When an adjoining facility is constructed, the fuel is cycled through a stabilization process and returned to the Staging and Storage Facility for dry interim (40-year) storage. The estimated total expenditure for this Recommended Path Forward, including necessary new construction, operations, and deactivation of Project facilities through 2012, is approximately $1,150 million (unescalated).« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christian, Jerry D.; Sterbentz, James W.

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.2more » 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.« less

Separator assembly for use in spent nuclear fuel shipping cask

DOEpatents

Bucholz, James A.

1983-01-01

A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

3DD - Three Dimensional Disposal of Spent Nuclear Fuel - 12449

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dvorakova, Marketa; Slovak, Jiri

2012-07-01

Three dimensional disposal is being considered as a way in which to store long-term spent nuclear fuel in underground disposal facilities in the Czech Republic. This method involves a combination of the two most common internationally recognised disposal methods in order to practically apply the advantages of both whilst, at the same time, eliminating their weaknesses; the method also allows easy removal in case of potential re-use. The proposed method for the disposal of spent nuclear fuel will reduce the areal requirements of future deep geological repositories by more than 30%. It will also simplify the container handling process bymore » using gravitational forces in order to meet requirements concerning the controllability of processes and ensuring operational and nuclear safety. With regard to the issue of the efficient potential removal of waste containers, this project offers an ingenious solution which does not disrupt the overall stability of the original disposal complex. (authors)« less

Dissolution of spent nuclear fuel in carbonate-peroxide solution

NASA Astrophysics Data System (ADS)

Soderquist, Chuck; Hanson, Brady

2010-01-01

This study shows that spent UO2 fuel can be completely dissolved in a room temperature carbonate-peroxide solution apparently without attacking the metallic Mo-Tc-Ru-Rh-Pd fission product phase. In parallel tests, identical samples of spent nuclear fuel were dissolved in nitric acid and in an ammonium carbonate, hydrogen peroxide solution. The resulting solutions were analyzed for strontium-90, technetium-99, cesium-137, europium-154, plutonium, and americium-241. The results were identical for all analytes except technetium, where the carbonate-peroxide dissolution had only about 25% of the technetium that the nitric acid dissolution had.

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2012-05-10

2009. 143 Abdul Mannan, “Preventing Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in...Program.” Some analysts argue that spent nuclear fuel is more vulnerable when being transported . 144 Martellini, 2008. Pakistan’s Nuclear Weapons...urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment technology, which it

Measurement and analysis of gamma-rays emitted from spent nuclear fuel above 3 MeV.

PubMed

Rodriguez, Douglas C; Anderson, Elaina; Anderson, Kevin K; Campbell, Luke W; Fast, James E; Jarman, Kenneth; Kulisek, Jonathan; Orton, Christopher R; Runkle, Robert C; Stave, Sean

2013-12-01

The gamma-ray spectrum of spent nuclear fuel in the 3-6 MeV energy range is important for active interrogation since gamma rays emitted from nuclear decay are not expected to interfere with measurements in this energy region. There is, unfortunately, a dearth of empirical measurements from spent nuclear fuel in this region. This work is an initial attempt to partially fill this gap by presenting an analysis of gamma-ray spectra collected from a set of spent nuclear fuel sources using a high-purity germanium detector array. This multi-crystal array possesses a large collection volume, providing high energy resolution up to 16 MeV. The results of these measurements establish the continuum count-rate in the energy region between 3 and 6 MeV. Also assessed is the potential for peaks from passive emissions to interfere with peak measurements resulting from active interrogation delayed emissions. As one of the first documented empirical measurements of passive emissions from spent fuel for energies above 3 MeV, this work provides a foundation for active interrogation model validation and detector development. © 2013 Elsevier Ltd. All rights reserved.

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE PAGES

Vaccaro, S.; Gauld, I. C.; Hu, J.; ...

2018-01-31

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false

Advancing the Fork detector for quantitative spent nuclear fuel verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Gauld, I. C.; Hu, J.

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations.more » A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This study describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false

Advancing the Fork detector for quantitative spent nuclear fuel verification

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Gauld, I. C.; Hu, J.; De Baere, P.; Peterson, J.; Schwalbach, P.; Smejkal, A.; Tomanin, A.; Sjöland, A.; Tobin, S.; Wiarda, D.

2018-04-01

The Fork detector is widely used by the safeguards inspectorate of the European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) to verify spent nuclear fuel. Fork measurements are routinely performed for safeguards prior to dry storage cask loading. Additionally, spent fuel verification will be required at the facilities where encapsulation is performed for acceptance in the final repositories planned in Sweden and Finland. The use of the Fork detector as a quantitative instrument has not been prevalent due to the complexity of correlating the measured neutron and gamma ray signals with fuel inventories and operator declarations. A spent fuel data analysis module based on the ORIGEN burnup code was recently implemented to provide automated real-time analysis of Fork detector data. This module allows quantitative predictions of expected neutron count rates and gamma units as measured by the Fork detectors using safeguards declarations and available reactor operating data. This paper describes field testing of the Fork data analysis module using data acquired from 339 assemblies measured during routine dry cask loading inspection campaigns in Europe. Assemblies include both uranium oxide and mixed-oxide fuel assemblies. More recent measurements of 50 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel are also analyzed. An evaluation of uncertainties in the Fork measurement data is performed to quantify the ability of the data analysis module to verify operator declarations and to develop quantitative go/no-go criteria for safeguards verification measurements during cask loading or encapsulation operations. The goal of this approach is to provide safeguards inspectors with reliable real-time data analysis tools to rapidly identify discrepancies in operator declarations and to detect potential partial defects in spent fuel assemblies with improved reliability and minimal false positive alarms

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koji Shirai

2006-04-01

The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and

Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dilger, Fred; Halstead, Robert J.; Ballard, James D.

Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very largemore » dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980

78 FR 45575 - Duke Energy Carolinas, LLC; Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-29

... will allow Oconee to effectively manage its spent fuel inventory to meet decay heat zoning requirements... thermal stresses, including potential elongation from decay heat and irradiation. In addition, the NRC...] system provides for the horizontal dry storage of canisterized spent fuel assemblies in a concrete...

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

1998-06-30

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

Dry halide method for separating the components of spent nuclear fuels

DOEpatents

Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

1998-01-01

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

Determination of plutonium in spent nuclear fuel using high resolution X-ray

DOE PAGES

McIntosh, Kathryn G.; Reilly, Sean D.; Havrilla, George J.

2015-05-30

Characterization of Pu is an essential aspect of safeguards operations at nuclear fuel reprocessing facilities. A novel analysis technique called hiRX (high resolution X-ray) has been developed for the direct measurement of Pu in spent nuclear fuel dissolver solutions. hiRX is based on monochromatic wavelength dispersive X-ray fluorescence (MWDXRF), which provides enhanced sensitivity and specificity compared with conventional XRF techniques. A breadboard setup of the hiRX instrument was calibrated using spiked surrogate spent fuel (SSF) standards prepared as dried residues. Samples of actual spent fuel were utilized to evaluate the performance of the hiRX. The direct detection of just 39more » ng of Pu is demonstrated. Initial quantitative results, with error of 4–27% and precision of 2% relative standard deviation (RSD), were obtained for spent fuel samples. The limit of detection for Pu (100 s) within an excitation spot of 200 μm diameter was 375 pg. This study demonstrates the potential for the hiRX technique to be utilized for the rapid, accurate, and precise determination of Pu. Moreover, the results highlight the analytical capability of hiRX for other applications requiring sensitive and selective nondestructive analyses.« less

Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

2018-01-01

China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process

DOEpatents

Heckman, R.A.

1980-12-19

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support te liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Method for calculating the duration of vacuum drying of a metal-concrete container for spent nuclear fuel

NASA Astrophysics Data System (ADS)

Karyakin, Yu. E.; Nekhozhin, M. A.; Pletnev, A. A.

2013-07-01

A method for calculating the quantity of moisture in a metal-concrete container in the process of its charging with spent nuclear fuel is proposed. A computing method and results obtained by it for conservative estimation of the time of vacuum drying of a container charged with spent nuclear fuel by technologies with quantization and without quantization of the lower fuel element cluster are presented. It has been shown that the absence of quantization in loading spent fuel increases several times the time of vacuum drying of the metal-concrete container.

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

NASA Astrophysics Data System (ADS)

Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A. A.

2017-01-01

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matthew Bunn; Steve Fetter; John P. Holdren

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recyclingmore » to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.« less

76 FR 40403 - R.E. Ginna Nuclear Power Plant, LLC, R.E. Ginna Nuclear Power Plant, R.E. Ginna Independent Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-08

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-244; Docket No. 72-67] R.E. Ginna Nuclear Power Plant, LLC, R.E. Ginna Nuclear Power Plant, R.E. Ginna Independent Spent Fuel Storage Installation; Notice of... Facility Operating License No. DPR-18, for the R.E. Ginna Nuclear Power Plant (Ginna), currently held by R...

Performance assessment of self-interrogation neutron resonance densitometry for spent nuclear fuel assay

NASA Astrophysics Data System (ADS)

Hu, Jianwei; Tobin, Stephen J.; LaFleur, Adrienne M.; Menlove, Howard O.; Swinhoe, Martyn T.

2013-11-01

Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the 239Pu and 235U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of 239Pu and 235U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that 235U content has a major impact on the SINRD signal in addition to the 239Pu content. Plutonium-241 and 241Am are the two main nuclides responsible for the variation in the count rate ratio with cooling time. In short, this work

78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-28

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 73 [NRC-2010-0340; NRC-2009-0163] RIN 3150-AI64 Physical..., ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This revised document sets forth means... physical protection of spent nuclear fuel (SNF) during transportation by road, rail, and water; and for...

Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

PubMed

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

Characterization of the radiation environment for a large-area interim spent-nuclear-fuel storage facility

NASA Astrophysics Data System (ADS)

Fortkamp, Jonathan C.

Current needs in the nuclear industry and movements in the political arena indicate that authorization may soon be given for development of a federal interim storage facility for spent nuclear fuel. The initial stages of the design work have already begun within the Department of Energy and are being reviewed by the Nuclear Regulatory Commission. This dissertation addresses the radiation environment around an interim spent nuclear fuel storage facility. Specifically the dissertation characterizes the radiation dose rates around the facility based on a design basis source term, evaluates the changes in dose due to varying cask spacing configurations, and uses these results to define some applicable health physics principles for the storage facility. Results indicate that dose rates from the facility are due primarily from photons from the spent fuel and Co-60 activation in the fuel assemblies. In the modeled cask system, skyshine was a significant contribution to dose rates at distances from the cask array, but this contribution can be reduced with an alternate cask venting system. With the application of appropriate health physics principles, occupation doses can be easily maintained far below regulatory limits and maintained ALARA.

Spent nuclear fuel discharges from US reactors 1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-05-05

This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactivemore » Waste Management.« less

Spent nuclear fuel canister storage building conceptual design report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swenson, C.E.

This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

Local negotiation on compensation siting of the spent nuclear fuel repository in Finland

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kojo, Matti

The aim of the paper is to analyse the local negotiation process between the Municipality of Eurajoki and the nuclear power company Teollisuuden Voima (TVO) and the nuclear waste management company Posiva Oy. The aim of the negotiations was to find an acceptable form of compensation for siting a spent nuclear fuel repository in Olkiluoto, Finland. The paper includes background information on the siting process in Finland, the local political setting in the Municipality of Eurajoki and a description of the negotiation process. The analysis of the negotiations on compensation is important for better understanding the progress of the Finnishmore » siting process. The paper describes the picture of the contest to host the spent nuclear fuel repository. It also provides more information on the relationship between the Municipality of Eurajoki and the power company TVO. The negotiations on compensation and the roles of various players in the negotiations have not been studied in detail because the minutes of the Vuojoki liaison group were not available before the decision of the Supreme Administrative Court in May 2006. (author)« less

Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Newsom, H.C.

This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted.

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

DOE PAGES

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena; ...

2016-10-22

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poulson, Daniel Cris; Durham, J. Matthew; Guardincerri, Elena

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casksmore » is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.« less

Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

DOE PAGES

Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; ...

2015-03-04

Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetricallymore » burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.« less

Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martinik, Tomas; Henzl, Vladimir; Grape, Sophie

Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetricallymore » burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.« less

Depleted uranium startup of spent-fuel treatment operations at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Mariani, R.D.; Bonomo, N.L.

1995-12-31

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of Experimental Breeder Reactor II (EBR-II) spent nuclear fuel. This fuel will be treated using an electrometallurgical process in the fuel conditioning facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The process equipment is undergoing testing with depleted uranium in preparation for irradiated fuel operations during the summer of 1995.

76 FR 34007 - Draft Regulatory Basis for a Potential Rulemaking on Spent Nuclear Fuel Reprocessing Facilities

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

... processes are more akin to fuel cycle processes. This framework was established in the 1970's to license the... nuclear power globally and close the nuclear fuel cycle through reprocessing spent fuel and deploying fast... Accounting;'' and a Nuclear Energy Institute white [[Page 34009

76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-16

... High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Public meeting... Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,'' and 73... Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage facilities. The draft regulatory...

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.

Chemical reactivity testing for the National Spent Nuclear Fuel Program. Revision 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koester, L.W.

This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, Y60-101PD, Quality Program Description, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will bemore » noted. The project consists of conducting three separate series of related experiments, ''Passivation of Uranium Hydride Powder With Oxygen and Water'', '''Passivation of Uranium Hydride Powder with Surface Characterization'', and ''Electrochemical Measure of Uranium Hydride Corrosion Rate''.« less

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and eDNA was extracted for pyrosequencing of bacterial rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of dissolved carbonmore » were co-localized with the precipitant relative to bulk water. Bacterial densities were highly variable between sampling locations and with depth; cell numbers (log scale) ranged from 5.6 to 4.89 cells / mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs / amplicon library (3% cutoff) and the greatest percent majority of sequences were aligned to the families Burkholderiales (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across diverse environmental landscapes were performed and microbiological signatures unique to the spent nuclear fuel storage basin environment were revealed. These insights could spur the development of tractable bioindicators that are specific of and diagnostic for water quality at

Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

NASA Astrophysics Data System (ADS)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2017-10-01

Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

Characterizing nuclear and mitochondrial DNA in spent embryo culture media: genetic contamination identified.

PubMed

Hammond, Elizabeth R; McGillivray, Brent C; Wicker, Sophie M; Peek, John C; Shelling, Andrew N; Stone, Peter; Chamley, Larry W; Cree, Lynsey M

2017-01-01

To characterize nuclear and mitochondrial DNA (mtDNA) in spent culture media from normally developing blastocysts to determine whether it could be used for noninvasive genetic assessment. Prospective embryo cohort study. Academic center and private in vitro fertilization (IVF) clinic. Seventy patients undergoing intracytoplasmic sperm injection (ICSI) and 227 blastocysts. Culture media assessment, artificial blastocoele fluid collapse and DNA analysis using digital polymerase chain reaction (dPCR), long-range PCR, quantitative PCR (qPCR), and DNA fingerprinting. Presence of nuclear and mtDNA in three different commercial culture media from Vitrolife and Irvine Scientific, spent embryo media assessment at the cleavage and blastocyst stages of development, and analysis of the internal media controls for each patient that had been exposed to identical conditions as embryo media but did not come into contact with embryos. Higher levels of nuclear and mtDNA were observed in the culture media that had been exposed to embryos compared with the internal media controls. Nuclear DNA (∼4 copies) and mtDNA (∼600 copies) could be detected in spent media, and the levels increased at the blastocyst stage. No increase in DNA was detected after artificial blastocoele fluid collapse. Mixed sex chromosome DNA was detected. This originated from contamination in the culture media and from maternal (cumulus) cells. Due to the limited amount of template, the presence of embryonic nuclear DNA could not be confirmed by DNA fingerprinting analysis. Currently DNA from culture media cannot be used for genetic assessment because embryo-associated structures release DNA into the culture medium and the DNA is of mixed origin. Copyright © 2016 American Society for Reproductive Medicine. Published by Elsevier Inc. All rights reserved.

Radiolytic degradation of a new diglycol-diamide ligand for actinide and lanthanide co-extraction from spent nuclear fuel

NASA Astrophysics Data System (ADS)

Ossola, Annalisa; Macerata, Elena; Tinonin, Dario A.; Faroldi, Federica; Giola, Marco; Mariani, Mario; Casnati, Alessandro

2016-07-01

Within the Partitioning and Transmutation strategies, great efforts have been devoted in the last decades to the development of lipophilic ligands able to co-extract trivalent Lanthanides (Ln) and Actinides (An) from spent nuclear fuel. Because of the harsh working conditions these ligands undergo, it is important to prove their chemical and radiolytic stability during the counter-current multi-stage extraction process. In the present work the hydrolytic and radiolytic resistance of the freshly prepared and aged organic solutions containing the new ligand (2,6-bis[(N-methyl-N-dodecyl)carboxamide]-4-methoxy-tetrahydro-pyran) were investigated in order to evaluate the impact on the safety and efficiency of the process. Liquid-liquid extraction tests with spiked solutions showed that the ligand extracting performances are strongly impaired by storing the samples at room temperature and in the light. Moreover, the extracting efficiency of the irradiated samples resulted to be influenced by gamma irradiation, while selectivity remains unchanged. Preliminary mass spectrometric data showed that degradation is mainly due to the acid-catalysed reaction of the ligand carboxamide and ether groups with the 1-octanol present in the diluent.

Spent Fuel Working Group Report. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

O`Toole, T.

1993-11-01

The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety.more » To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary`s initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group`s Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities.« less

Modeling of molecular and particulate transport in dry spent nuclear fuel canisters

NASA Astrophysics Data System (ADS)

Casella, Andrew M.

2007-09-01

The transportation and storage of spent nuclear fuel is one of the prominent issues facing the commercial nuclear industry today, as there is still no general consensus regarding the near- and long-term strategy for managing the back-end of the nuclear fuel cycle. The debate continues over whether the fuel cycle should remain open, in which case spent fuel will be stored at on-site reactor facilities, interim facilities, or a geologic repository; or if the fuel cycle should be closed, in which case spent fuel will be recycled. Currently, commercial spent nuclear fuel is stored at on-site reactor facilities either in pools or in dry storage containers. Increasingly, spent fuel is being moved to dry storage containers due to decreased costs relative to pools. As the number of dry spent fuel containers increases and the roles they play in the nuclear fuel cycle increase, more regulations will be enacted to ensure that they function properly. Accordingly, they will have to be carefully analyzed for normal conditions, as well as any off-normal conditions of concern. This thesis addresses the phenomena associated with one such concern; the formation of a microscopic through-wall breach in a dry storage container. Particular emphasis is placed on the depressurization of the canister, release of radioactivity, and plugging of the breach due to deposition of suspended particulates. The depressurization of a dry storage container upon the formation of a breach depends on the temperature and quantity of the fill gas, the pressure differential across the breach, and the size of the breach. The first model constructed in this thesis is capable of determining the depressurization time for a breached container as long as the associated parameters just identified allow for laminar flow through the breach. The parameters can be manipulated to quantitatively determine their effect on depressurization. This model is expanded to account for the presence of suspended particles. If

Spent nuclear fuel recycling with plasma reduction and etching

DOEpatents

Kim, Yong Ho

2012-06-05

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF₆ and PuF₄. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF₆ (gas) that is formed to decompose back to PuF₄ (solid), and b) to maintain stability of the UF₆. Uranium (in the form of gaseous UF₆) is easily extracted and separated from the plutonium (in the form of solid PuF₄). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF₆ (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF₄ or NF₃ as the fluorine sources instead of F₂ or HF.

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

NASA Astrophysics Data System (ADS)

Durham, J. M.; Poulson, D.; Bacon, J.; Chichester, D. L.; Guardincerri, E.; Morris, C. L.; Plaud-Ramos, K.; Schwendiman, W.; Tolman, J. D.; Winston, P.

2018-04-01

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. Here we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. This application of technology and methods commonly used in high-energy particle physics provides a potential solution to this long-standing problem in international nuclear safeguards.

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

Method of increasing the deterrent to proliferation of nuclear fuels

DOEpatents

Rampolla, Donald S.

1982-01-01

A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy

2013-07-01

Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Doherty, A.L.

1997-06-01

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release frommore » the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.« less

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-12-09

Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008...gave additional urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment...technology, which it mastered by the mid-1980s. Highly-enriched uranium (HEU) is one of two types of fissile material used in nuclear weapons; the other

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-10-15

and technical measures to prevent unauthorized or accidental use of nuclear weapons, as well as contribute to physical security of storage ...Talks On Nuclear Security,” The Boston Globe, May 5, 2009. 79 Abdul Mannan, “Preventing Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or...a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008. 80 Martellini, 2008. 81 For more information

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Durham, J. M.; Poulson, D.; Bacon, J.

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE PAGES

Durham, J. M.; Poulson, D.; Bacon, J.; ...

2018-04-10

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, D.J.; McTaggart, D.R.

1983-08-31

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Separation of uranium from technetium in recovery of spent nuclear fuel

DOEpatents

Pruett, David J.; McTaggart, Donald R.

1984-01-01

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc.sup.+7 therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

NDE of copper canisters for long-term storage of spent nuclear fuel from the Swedish nuclear power plants

NASA Astrophysics Data System (ADS)

Stepinski, Tadeusz

2003-07-01

Sweden has been intensively developing methods for long term storage of spent fuel from the nuclear power plants for twenty-five years. A dedicated research program has been initiated and conducted by the Swedish company SKB (Swedish Nuclear Fuels and Waste Management Co.). After the interim storage SKB plans to encapsulate spent nuclear fuel in copper canisters that will be placed at a deep repository located in bedrock. The canisters filled with fuel rods will be sealed by an electron beam weld. This paper presents three complementary NDE techniques used for assessing the sealing weld in copper canisters, radiography, ultrasound, and eddy current. A powerful X-ray source and a digital detector are used for the radiography. An ultrasonic array system consisting of a phased ultrasonic array and a multi-channel electronics is used for the ultrasonic examination. The array system enables electronic focusing and rapid electronic scanning eliminating the use of a complicated mechanical scanner. A specially designed eddy current probe capable of detecting small voids at the depth up to 4 mm in copper is used for the eddy current inspection. Presently, all the NDE techniques are verified in SKB's Canister Laboratory where full scale canisters are welded and examined.

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2013 CFR

2013-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2014 CFR

2014-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE PAGES

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.; ...

2017-04-28

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Analysis of new measurements of Calvert Cliffs spent fuel samples using SCALE 6.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Giaquinto, J. M.; Gauld, I. C.

High quality experimental data for isotopic compositions in irradiated fuel are important to spent fuel applications, including nuclear safeguards, spent fuel storage, transportation, and final disposal. The importance of these data has been increasingly recognized in recent years, particularly as countries like Finland and Sweden plan to open the world’s first two spent fuel geological repositories in 2020s, while other countries, including the United States, are considering extended dry fuel storage options. Destructive and nondestructive measurements of a spent fuel rod segment from a Combustion Engineering 14 × 14 fuel assembly of the Calvert Cliffs Unit 1 nuclear reactor havemore » been recently performed at Oak Ridge National Laboratory (ORNL). These ORNL measurements included two samples selected from adjacent axial locations of a fuel rod with initial enrichment of 3.038 wt% 235U, which achieved burnups close to 43.5 GWd/MTU. More than 50 different isotopes of 16 elements were measured using high precision measurement methods. Various investigations have assessed the quality of the new ORNL measurement data, including comparison to previous measurements and to calculation results. Previous measurement data for samples from the same fuel rod measured at ORNL are available from experiments performed at Pacific Northwest National Laboratory in the United States and the Khoplin Radium Institute in Russia. Detailed assembly models were developed using the newly released SCALE 6.2 code package to simulate depletion and decay of the measured fuel samples. Furthermore, results from this work show that the new ORNL measurements provide a good quality radiochemical assay data set for spent fuel with relatively high burnup and long cooling time, and they can serve as good benchmark data for nuclear burnup code validation and spent fuel studies.« less

Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

DOE PAGES

Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; ...

2015-09-10

Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities,more » the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.« less

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pond, R.B.; Matos, J.E.

1996-05-01

As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, aremore » not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.« less

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2010-02-04

Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008...measures to prevent unauthorized or accidental use of nuclear weapons, as well as contribute to physical security of storage facilities and personnel...strategic nuclear assets could be obtained by terrorists, or used by elements in the Pakistani government. Chair of the Joint Chiefs of Staff Admiral

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-07-30

Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008. 79...that Pakistan’s strategic nuclear assets could be obtained by terrorists, or used by elements in the Pakistani government. Chair of the Joint Chiefs...that gave additional urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2010-10-07

Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008; Martellini, 2008. 99...prevent unauthorized or accidental use of nuclear weapons, as well as contribute to physical security of storage facilities and personnel reliability... nuclear assets could be obtained by terrorists, or used by elements in the Pakistani government. Chair of the Joint Chiefs of Staff Admiral Michael

Inhibition of EGFR nuclear shuttling decreases irradiation resistance in HeLa cells.

PubMed

Wei, Hong; Zhu, Zijie; Lu, Longtao

2017-01-01

Cervical cancer is a leading cause of mortality in women worldwide. The resistance to irradiation at the advanced stage is the main reason for the poor prognosis and high mortality. This work aims to elucidate the molecular mechanism underlying the radio-resistance. In this study, we determined the pEGFR-T654 and pDNA-PK-T2609 expression level changes in irradiated HeLa cells treated with T654 peptide, a nuclear localization signal (NLS) inhibitor, to inhibit EGFR nuclear transport. Cell viability, cell cycle and migratory capacity were analyzed. Xenograft animal model was used to evaluate the effect of EGFR nuclear transport inhibition on the tumor growth in vivo. The enhanced translocation of nuclear EGFR in the irradiated HeLa cells correlated with the increasing level of pEGFR-T654 and pDNA-PK-T2609. Inhibition of EGFR nuclear translocation by NLS peptide inhibitor attenuated DNA damage repair in the irradiated HeLa cells, decreased cell viability and promoted cell death through arrest at G0 phase. NLS peptide inhibitor impaired the migratory capacity of irradiated HeLa cells, and negatively affected tumorigenesis in xenograft mice. This work puts forward a potential molecular mechanism of the irradiation resistance in cervical cancer cells, providing a promising direction towards an efficient therapy of cervical cancer.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE PAGES

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael; ...

2016-11-17

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sharp, Nicholas; Ticknor, Brian W.; Bronikowski, Michael

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

Development of U-frame bending system for studying the vibration integrity of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Wang, Hong; Wang, Jy-An John; Tan, Ting; Jiang, Hao; Cox, Thomas S.; Howard, Rob L.; Bevard, Bruce B.; Flanagan, Michelle

2013-09-01

A bending fatigue system developed to evaluate the response of spent nuclear fuel rods to vibration loads is presented. A U-frame testing setup is used for imposing bending loads on the fuel rod specimen. The U-frame setup consists of two rigid arms, side connecting plates to the rigid arms, and linkages to a universal testing machine. The test specimen's curvature is obtained through a three-point deflection measurement method. The tests using surrogate specimens with stainless steel cladding revealed increased flexural rigidity under unidirectional cyclic bending, significant effect of cladding-pellets bonding on the response of surrogate rods, and substantial cyclic softening in reverse bending mode. These phenomena may cast light on the expected response of a spent nuclear fuel rod. The developed U-frame system is thus verified and demonstrated to be ready for further pursuit in hot-cell tests.

NEUTRON CHARACTERIZATION OF ENSA-DPT TYPE SPENT FUEL CASK AT TRILLO NUCLEAR POWER PLANT.

PubMed

Méndez-Villafañe, Roberto; Campo-Blanco, Xandra; Embid, Miguel; Yéboles, César A; Morales, Ramón; Novo, Manuel; Sanz, Javier

2018-04-23

The Neutron Standards Laboratory of CIEMAT has conducted the characterization of the independent spent fuel storage installation at the Trillo Nuclear Power Plant. At this facility, the spent fuel assemblies are stored in ENSA-DPT type dual purpose casks. Neutron characterization was performed by dosimetry measurements with a neutron survey meter (LB6411) inside the facility, around an individual cask and between stored casks, and outside the facility. Spectra measurements were also performed with a Bonner sphere system in order to determine the integral quantities and validate the use of the neutron monitor at the different positions. Inside the facility, measured neutron spectra and neutron ambient dose equivalent rate are consistent with the casks spatial distribution and neutron emission rates, and measurements with both instruments are consistent with each other. Outside the facility, measured neutron ambient dose equivalent rates are well below the 0.5 μSv/h limit established by the nuclear regulatory authority.

Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.; Coops, M.S.

1982-01-19

A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A nonoxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel.

An allowable cladding peak temperature for spent nuclear fuels in interim dry storage

NASA Astrophysics Data System (ADS)

Cha, Hyun-Jin; Jang, Ki-Nam; Kim, Kyu-Tae

2018-01-01

Allowable cladding peak temperatures for spent fuel cladding integrity in interim dry storage were investigated, considering hydride reorientation and mechanical property degradation behaviors of unirradiated and neutron irradiated Zr-Nb cladding tubes. Cladding tube specimens were heated up to various temperatures and then cooled down under tensile hoop stresses. Cool-down specimens indicate that higher heat-up temperature and larger tensile hoop stress generated larger radial hydride precipitation and smaller tensile strength and plastic hoop strain. Unirradiated specimens generated relatively larger radial hydride precipitation and plastic strain than did neutron irradiated specimens. Assuming a minimum plastic strain requirement of 5% for cladding integrity maintenance in interim dry storage, it is proposed that a cladding peak temperature during the interim dry storage is to keep below 250 °C if cladding tubes are cooled down to room temperature.

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2012 CFR

2012-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the...

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa

2006-04-01

In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstrationmore » project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.

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

Nuclear Nonproliferation: Concerns With U.S. Delays in Accepting Foreign Research Reactors’ Spent Fuel

DTIC Science & Technology

1994-03-01

transport or storage plans. The return of some of the spent fuel will also depend on the readiness of dry storage . One expert told us that...enriched uranium fuel (HEU), a material that can be used to make nuclear bombs, in civilian nuclear programs worldwide. Research reactors are of...address the environmental impact of transporting the fuel and storing it in both existing and new storage units, possibly by June 1995. Under the

Strategic Minimization of High Level Waste from Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, Michael F.; Benedict, Robert W.

The pyroprocessing of spent nuclear fuel results in two high-level waste streams--ceramic and metal waste. Ceramic waste contains active metal fission product-loaded salt from the electrorefining, while the metal waste contains cladding hulls and undissolved noble metals. While pyroprocessing was successfully demonstrated for treatment of spent fuel from Experimental Breeder Reactor-II in 1999, it was done so without a specific objective to minimize high-level waste generation. The ceramic waste process uses “throw-away” technology that is not optimized with respect to volume of waste generated. In looking past treatment of EBR-II fuel, it is critical to minimize waste generation for technologymore » developed under the Global Nuclear Energy Partnership (GNEP). While the metal waste cannot be readily reduced, there are viable routes towards minimizing the ceramic waste. Fission products that generate high amounts of heat, such as Cs and Sr, can be separated from other active metal fission products and placed into short-term, shallow disposal. The remaining active metal fission products can be concentrated into the ceramic waste form using an ion exchange process. It has been estimated that ion exchange can reduce ceramic high-level waste quantities by as much as a factor of 3 relative to throw-away technology.« less

Method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

1982-01-19

Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

Systems and methods for processing irradiation targets through a nuclear reactor

DOEpatents

Dayal, Yogeshwar; Saito, Earl F.; Berger, John F.; Brittingham, Martin W.; Morales, Stephen K.; Hare, Jeffrey M.

2016-05-03

Apparatuses and methods produce radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets may be inserted and removed from instrumentation tubes during operation and converted to radioisotopes otherwise unavailable during operation of commercial nuclear reactors. Example apparatuses may continuously insert, remove, and store irradiation targets to be converted to useable radioisotopes or other desired materials at several different origin and termination points accessible outside an access barrier such as a containment building, drywell wall, or other access restriction preventing access to instrumentation tubes during operation of the nuclear plant.

Cosmic Ray Muon Imaging of Spent Nuclear Fuel in Dry Storage Casks

DOE PAGES

Durham, J. Matthew; Guardincerri, Elena; Morris, Christopher L.; ...

2016-04-29

In this paper, cosmic ray muon radiography has been used to identify the absence of spent nuclear fuel bundles inside a sealed dry storage cask. The large amounts of shielding that dry storage casks use to contain radiation from the highly radioactive contents impedes typical imaging methods, but the penetrating nature of cosmic ray muons allows them to be used as an effective radiographic probe. This technique was able to successfully identify missing fuel bundles inside a sealed Westinghouse MC-10 cask. This method of fuel cask verification may prove useful for international nuclear safeguards inspectors. Finally, muon radiography may findmore » other safety and security or safeguards applications, such as arms control verification.« less

Transportation of spent MTR fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raisonnier, D.

1997-08-01

This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Head, C.R.

1997-08-01

This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

Calculation of the process of vacuum drying of a metal-concrete container with spent nuclear fuel

NASA Astrophysics Data System (ADS)

Karyakin, Yu. E.; Lavrent'ev, S. A.; Pavlyukevich, N. V.; Pletnev, A. A.; Fedorovich, E. D.

2012-01-01

An algorithm and results of calculation of the process of vacuum drying of a metal-concrete container intended for long-term "dry" storage of spent nuclear fuel are presented. A calculated substantiation of the initial amount of moisture in the container is given.

Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

DOEpatents

Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

1982-01-19

Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in the reaction region of a separation vessel which includes a reflux region positioned above the molten tin solvent. The reflux region minimizes loss of evaporated solvent during the separation of the actinide fuels from the volatile fission products. Additionally, inclusion of the reflux region permits the separation of the more volatile fission products (noncondensable) from the less volatile ones (condensable).

Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository--Volume 2: Methodology and Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor, L.L.; Wilson, J.R.; Sanchez, L.C.

1998-10-01

The US Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality duringmore » the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).« less

Active Interrogation for Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; Dougan, Arden

2015-11-05

The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

Irradiation of nuclear track emulsions with thermal neutrons, heavy ions, and muons

NASA Astrophysics Data System (ADS)

Artemenkov, D. A.; Bradnova, V.; Zaitsev, A. A.; Zarubin, P. I.; Zarubina, I. G.; Kattabekov, R. R.; Mamatkulov, K. Z.; Rusakova, V. V.

2015-07-01

Exposures of test samples of nuclear track emulsion were analyzed. Angular and energy correlations of products originating from the thermal-neutron-induced reaction n th +10 B → 7 Li + (γ)+ α were studied in nuclear track emulsions enriched in boron. Nuclear track emulsions were also irradiated with 86Kr+17 and 124Xe+26 ions of energy about 1.2 MeV per nucleon. Measurements of ranges of heavy ions in nuclear track emulsionsmade it possible to determine their energies on the basis of the SRIM model. The formation of high-multiplicity nuclear stars was observed upon irradiating nuclear track emulsions with ultrarelativistic muons. Kinematical features studied in this exposure of nuclear track emulsions for events of the muon-induced splitting of carbon nuclei to three alpha particles are indicative of the nucleardiffraction interaction mechanism.

Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald

1999-09-01

The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was frommore » a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.« less

Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Janney, D. E.; Keiser, D. D.

2003-09-01

Argonne National Laboratory has developed a pyroprocessing-based technique for conditioning spent sodium-bonded nuclear-reactor fuel in preparation for long-term disposal. The technique produces a metallic waste form whose nominal composition is stainless steel with 15 wt.% Zr (SS-15Zr), up to ˜ 11 wt.% actinide elements (primarily uranium), and a few percent metallic fission products. Actual and simulated waste forms show similar eutectic microstructures with approximately equal proportions of iron solid solution phases and Fe-Zr intermetallics. This article reports on an analysis of simulated waste forms containing uranium, neptunium, and plutonium.

Irradiation of nuclear track emulsions with thermal neutrons, heavy ions, and muons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Artemenkov, D. A., E-mail: artemenkov@lhe.jinr.ru; Bradnova, V.; Zaitsev, A. A.

Exposures of test samples of nuclear track emulsion were analyzed. Angular and energy correlations of products originating from the thermal-neutron-induced reaction n{sub th} +{sup 10} B → {sup 7} Li + (γ)+ α were studied in nuclear track emulsions enriched in boron. Nuclear track emulsions were also irradiated with {sup 86}Kr{sup +17} and {sup 124}Xe{sup +26} ions of energy about 1.2 MeV per nucleon. Measurements of ranges of heavy ions in nuclear track emulsionsmade it possible to determine their energies on the basis of the SRIM model. The formation of high-multiplicity nuclear stars was observed upon irradiating nuclear track emulsionsmore » with ultrarelativistic muons. Kinematical features studied in this exposure of nuclear track emulsions for events of the muon-induced splitting of carbon nuclei to three alpha particles are indicative of the nucleardiffraction interaction mechanism.« less

Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, Steven C.; Madey,Theodore E.; Haustein, Peter E.

2000-06-01

The purpose of this project is to deliver pertinent information that can be used to make rational decisions about the safety and treatment issues associated with dry storage of spent nuclear fuel materials. In particular, we will establish an understanding of: (1) water interactions with failed-fuel rods and metal-oxide materials; (2) the role of thermal processes and radiolysis (solid-state and interfacial) in the generation of potentially explosive mixtures of gaseous H2 and O2; and (3) the potential role of radiation-assisted corrosion during fuel rod storage.

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, Charles W.

1992-01-01

A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

DOE PAGES

Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.; ...

2016-01-01

Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

Resolving Past Liabilities for Future Reduction in Greenhouse Gases; Nuclear Energy and the Outstanding Federal Liability of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Donohue, Jay

This thesis will: (1) examine the current state of nuclear power in the U.S.; (2) provide a comparison of nuclear power to both existing alternative/renewable sources of energy as well as fossil fuels; (3) dissect Standard Contracts created pursuant to the National Waste Policy Act (NWPA), Congress' attempt to find a solution for Spent Nuclear Fuel (SNF), and the designation of Yucca Mountain as a repository; (4) the anticipated failure of Yucca Mountain; (5) explore WIPP as well as attempts to build a facility on Native American land in Utah; (6) examine reprocessing as a solution for SNF used by France and Japan; and, finally, (7) propose a solution to reduce GHG's by developing new nuclear energy plants with financial support from the U.S. government and a solution to build a storage facility for SNF through the sitting of a repository based on a "bottom-up" cooperative federalism approach.

Gamma-ray mirror technology for NDA of spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Descalle, M. A.; Ruz-Armendariz, J.; Decker, T.

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 ofmore » 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.« less

High burn-up spent nuclear fuel transport reliability investigation

DOE PAGES

Wang, Jy-An; Wang, Hong; Jiang, Hao; ...

2018-04-15

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 hasmore » 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.« less

High burn-up spent nuclear fuel transport reliability investigation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An; Wang, Hong; Jiang, Hao

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 hasmore » 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.« less

Spent fuel data base: commercial light water reactors. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

Analysis of Transportation Options for Commercial Spent Fuel in the U.S.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kalinina, Elena; Busch, Ingrid Karin

.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage The U.S. Department of Energy (DOE) is laying the groundwork for implementing interim storage and associated transportation of spent nuclear fuel (SNF) highand associated transportation of spent nuclear fuel (SNF) and high and

Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions

NASA Astrophysics Data System (ADS)

Åberg Lindell, M.; Andersson, P.; Grape, S.; Håkansson, A.; Thulin, M.

2018-07-01

In addition to verifying operator declared parameters of spent nuclear fuel, the ability to experimentally infer such parameters with a minimum of intrusiveness is of great interest and has been long-sought after in the nuclear safeguards community. It can also be anticipated that such ability would be of interest for quality assurance in e.g. recycling facilities in future Generation IV nuclear fuel cycles. One way to obtain information regarding spent nuclear fuel is to measure various gamma-ray intensities using high-resolution gamma-ray spectroscopy. While intensities from a few isotopes obtained from such measurements have traditionally been used pairwise, the approach in this work is to simultaneously analyze correlations between all available isotopes, using multivariate analysis techniques. Based on this approach, a methodology for inferring burnup, cooling time, and initial fissile content of PWR fuels using passive gamma-ray spectroscopy data has been investigated. PWR nuclear fuels, of UOX and MOX type, and their gamma-ray emissions, were simulated using the Monte Carlo code Serpent. Data comprising relative isotope activities was analyzed with decision trees and support vector machines, for predicting fuel parameters and their associated uncertainties. From this work it may be concluded that up to a cooling time of twenty years, the 95% prediction intervals of burnup, cooling time and initial fissile content could be inferred to within approximately 7 MWd/kgHM, 8 months, and 1.4 percentage points, respectively. An attempt aiming to estimate the plutonium content in spent UOX fuel, using the developed multivariate analysis model, is also presented. The results for Pu mass estimation are promising and call for further studies.

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

A Specific Long-Term Plan for Management of U.S. Nuclear Spent Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levy, Salomon

2006-07-01

A specific plan consisting of six different steps is proposed to accelerate and improve the long-term management of U.S. Light Water Reactor (LWR) spent nuclear fuel. The first step is to construct additional, centralized, engineered (dry cask) spent fuel facilities to have a backup solution to Yucca Mountain (YM) delays or lack of capacity. The second step is to restart the development of the Integral Fast Reactor (IFR), in a burner mode, because of its inherent safety characteristics and its extensive past development in contrast to Acceleration Driven Systems (ADS). The IFR and an improved non-proliferation version of its pyro-processingmore » technology can burn the plutonium (Pu) and minor actinides (MA) obtained by reprocessing LWR spent fuel. The remaining IFR and LWR fission products will be treated for storage at YM. The radiotoxicity of that high level waste (HLW) will fall below that of natural uranium in less than one thousand years. Due to anticipated increased capital, maintenance, and research costs for IFR, the third step is to reduce the required number of IFRs and their potential delays by implementing multiple recycles of Pu and Neptunium (Np) MA in LWR. That strategy is to use an advanced separation process, UREX+, and the MIX Pu option where the role and degradation of Pu is limited by uranium enrichment. UREX+ will decrease proliferation risks by avoiding Pu separation while the MIX fuel will lead to an equilibrium fuel recycle mode in LWR which will reduce U. S. Pu inventory and deliver much smaller volumes of less radioactive HLW to YM. In both steps two and three, Research and Development (R and D) is to emphasize the demonstration of multiple fuel reprocessing and fabrication, while improving HLW treatment, increasing proliferation resistance, and reducing losses of fissile material. The fourth step is to license and construct YM because it is needed for the disposal of defense wastes and the HLW to be generated under the proposed plan

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, C.W.

1992-03-24

A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

The effect of fission-energy Xe ion irradiation on the structural integrity and dissolution of the CeO2 matrix

NASA Astrophysics Data System (ADS)

Popel, A. J.; Le Solliec, S.; Lampronti, G. I.; Day, J.; Petrov, P. K.; Farnan, I.

2017-02-01

This work considers the effect of fission fragment damage on the structural integrity and dissolution of the CeO2 matrix in water, as a simulant for the UO2 matrix of spent nuclear fuel. For this purpose, thin films of CeO2 on Si substrates were produced and irradiated by 92 MeV 129Xe23+ ions to a fluence of 4.8 × 1015 ions/cm2 to simulate fission damage that occurs within nuclear fuels along with bulk CeO2 samples. The irradiated and unirradiated samples were characterised and a static batch dissolution experiment was conducted to study the effect of the induced irradiation damage on dissolution of the CeO2 matrix. Complex restructuring took place in the irradiated films and the irradiated samples showed an increase in the amount of dissolved cerium, as compared to the corresponding unirradiated samples. Secondary phases were also observed on the surface of the irradiated CeO2 films after the dissolution experiment.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

DOEpatents

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel and Transuranic Radioactive Wastes (40 CFR Part 191)

EPA Pesticide Factsheets

This regulation sets environmental standards for public protection from the management and disposal of spent nuclear fuel, high-level wastes and wastes that contain elements with atomic numbers higher than uranium (transuranic wastes).

77 FR 76952 - Rescinding Spent Fuel Pool Exclusion Regulations

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-31

... [Docket No. PRM-51-29; NRC-2012-0215] Rescinding Spent Fuel Pool Exclusion Regulations AGENCY: Nuclear... NRC institute a rulemaking to rescind the regulations excluding consideration of spent fuel pool...

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; D. Vaden; S.X. Li

During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontallymore » displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor

Default operational intervention levels (OILs) for severe nuclear power plant or spent fuel pool emergencies.

PubMed

McKenna, T; Kutkov, V; Vilar Welter, P; Dodd, B; Buglova, E

2013-05-01

Experience and studies show that for an emergency at a nuclear power plant involving severe core damage or damage to the fuel in spent fuel pools, the following actions may need to be taken in order to prevent severe deterministic health effects and reduce stochastic health effects: (1) precautionary protective actions and other response actions for those near the facility (i.e., within the zones identified by the International Atomic Energy Agency) taken immediately upon detection of facility conditions indicating possible severe damage to the fuel in the core or in the spent fuel pool; and (2) protective actions and other response actions taken based on environmental monitoring and sampling results following a release. This paper addresses the second item by providing default operational intervention levels [OILs, which are similar to the U.S. derived response levels (DRLs)] for promptly assessing radioactive material deposition, as well as skin, food, milk and drinking water contamination, following a major release of fission products from the core or spent fuel pool of a light water reactor (LWR) or a high power channel reactor (RBMK), based on the International Atomic Energy Agency's guidance.

Comparative analyses of spent nuclear fuel transport modal options: Transport options under existing site constraints

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brentlinger, L.A.; Hofmann, P.L.; Peterson, R.W.

1989-08-01

The movement of nuclear waste can be accomplished by various transport modal options involving different types of vehicles, transport casks, transport routes, and intermediate intermodal transfer facilities. A series of systems studies are required to evaluate modal/intermodal spent fuel transportation options in a consistent fashion. This report provides total life-cycle cost and life-cycle dose estimates for a series of transport modal options under existing site constraints. 14 refs., 7 figs., 28 tabs.

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2010 CFR

2010-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved spent...

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2011 CFR

2011-01-01

... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved spent...

Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian J; Bowman, Stephen M; Gauld, Ian C

2015-01-01

[Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (k eff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, andmore » it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

Neutron Focusing Mirrors for Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

NASA Astrophysics Data System (ADS)

Rai, Durgesh K.; Wu, Huarui; Abir, Muhammad; Giglio, Jeffrey; Khaykovich, Boris

Post irradiation examination (PIE) of samples irradiated in nuclear reactors is a challenging but necessary task for the development on novel nuclear power reactors. Idaho National Laboratory (INL) has neutron radiography capabilities, which are especially useful for the PIE of irradiated nuclear fuel. These capabilities are limited due to the extremely high gamma-ray radiation from the irradiated fuel, which precludes the use of standard digital detectors, in turn limiting the ability to do tomography and driving the cost of the measurements. In addition, the small 250 kW Neutron Radiography Reactor (NRAD) provides a relatively weak neutron flux, which leads to low signal-to-noise ratio. In this work, we develop neutron focusing optics suitable for the installation at NRAD. The optics would separate the sample and the detector, potentially allowing for the use of digital radiography detectors, and would provide significant intensity enhancement as well. The optics consist of several coaxial nested Wolter mirrors and is suited for polychromatic thermal neutron radiation. Laboratory Directed Research and Development program of Idaho National Laboratory.

Comparison between the Strength Levels of Baseline Nuclear-Grade Graphite and Graphite Irradiated in AGC-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carroll, Mark Christopher

2015-07-01

This report details the initial comparison of mechanical strength properties between the cylindrical nuclear-grade graphite specimens irradiated in the second Advanced Graphite Creep (AGC-2) experiment with the established baseline, or unirradiated, mechanical properties compiled in the Baseline Graphite Characterization program. The overall comparative analysis will describe the development of an appropriate test protocol for irradiated specimens, the execution of the mechanical tests on the AGC-2 sample population, and will further discuss the data in terms of developing an accurate irradiated property distribution in the limited amount of irradiated data by leveraging the considerably larger property datasets being captured in themore » Baseline Graphite Characterization program. Integrating information on the inherent variability in nuclear-grade graphite with more complete datasets is one of the goals of the VHTR Graphite Materials program. Between “sister” specimens, or specimens with the same geometry machined from the same sub-block of graphite from which the irradiated AGC specimens were extracted, and the Baseline datasets, a comprehensive body of data will exist that can provide both a direct and indirect indication of the full irradiated property distributions that can be expected of irradiated nuclear-grade graphite while in service in a VHTR system. While the most critical data will remain the actual irradiated property measurements, expansion of this data into accurate distributions based on the inherent variability in graphite properties will be a crucial step in qualifying graphite for nuclear use as a structural material in a VHTR environment.« less

10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Conditions for spent fuel storage cask reapproval. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trond Bjornard; Philip C. Durst

2012-05-01

This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA)more » of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a country’s safeguards agreement

REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

DOE Office of Scientific and Technical Information (OSTI.GOV)

W. Lee Poe, Jr

1998-10-01

In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are themore » same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR

Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

NASA Astrophysics Data System (ADS)

Kaplan, Alexis C.; Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P.; Flaska, Marek; Pozzi, Sara A.

2014-11-01

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

2018-03-01

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

TEST SYSTEM FOR EVALUATING SPENT NUCLEAR FUEL BENDING STIFFNESS AND VIBRATION INTEGRITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom

2013-01-01

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements specified by federal regulations. For normal conditions of transport, vibration loads incident to transport must be considered. This is particularly relevant for high-burnup fuel (>45 GWd/MTU). As the burnup of the fuel increases, a number of changes occur that may affect the performance of the fuel and cladding in storage and during transportation. The mechanical properties of high-burnup de-fueled cladding have been previously studied by subjecting defueled cladding tubes to longitudinal (axial) tensile tests, ring-stretch tests, ring-compression tests, and biaxial tube burst tests. The objective of this study ismore » to investigate the mechanical properties and behavior of both the cladding and the fuel in it under vibration/cyclic loads similar to the sustained vibration loads experienced during normal transport. The vibration loads to SNF rods during transportation can be characterized by dynamic, cyclic, bending loads. The transient vibration signals in a specified transport environment can be analyzed, and frequency, amplitude and phase components can be identified. The methodology being implemented is a novel approach to study the vibration integrity of actual SNF rod segments through testing and evaluating the fatigue performance of SNF rods at defined frequencies. Oak Ridge National Laboratory (ORNL) has developed a bending fatigue system to evaluate the response of the SNF rods to vibration loads. A three-point deflection measurement technique using linear variable differential transformers is used to characterize the bending rod curvature, and electromagnetic force linear motors are used as the driving system for mechanical loading. ORNL plans to use the test system in a hot cell for SNF vibration testing on high burnup, irradiated fuel to evaluate the pellet-clad interaction and bonding on the effective lifetime of fuel-clad structure bending fatigue performance

Gamma and fast neutron radiation monitoring inside spent reactor fuel assemblies

NASA Astrophysics Data System (ADS)

Lakosi, L.; Tam Nguyen, C.

2007-09-01

Gamma and neutron signatures of spent reactor fuel were monitored by small-size silicon diode and track etch detectors, respectively, in a nuclear power plant (NPP). These signatures, reflecting gross gamma intensity and the 242,244Cm content, contain information on the burn-up (BU) and cooling time (CT) of the fuel. The small size of the detectors allows close access to inside parts of the assemblies out of reach of other methods. A commercial Si diode was encapsulated in a cylindrical steel case and was used for gross γ monitoring. CR-39 detectors were used for neutron measurements. Irradiation exposures at the NPP were implemented in the central dosimetric channel of spent fuel assemblies (SFAs) stored in borated water. Gross γ and neutron axial profiles were taken up by scanning with the aid of a long steel guide tube, lowered down to the spent fuel pond by crane and fitted to the headpiece of the fuel assemblies. Gamma measurements were performed using a long cable introduced in this tube, with the Si diode at the end. A long steel wire was also led through the guide tube, to which a chain of 15 sample holder capsules was attached, each containing a track detector. Gamma dose rates of 0.1-10 kGy h -1, while neutron fluxes in a range of (0.25-26) 10 4 cm -2 s -1 were recorded. The results are in good correlation with those of a calculation for spent fuel neutron yield.

A discussion on the methodology for calculating radiological and toxicological consequences for the spent nuclear fuel project at the Hanford Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

RITTMANN, P.D.

1999-07-14

This report contains technical information used to determine accident consequences for the Spent Nuclear Fuel Project safety documents. It does not determine accident consequences or describe specific accident scenarios, but instead provides generic information.

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

DOE PAGES

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.; ...

2018-03-09

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levelsmore » of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 10 3to 4 × 104 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins.

PubMed

Bagwell, Christopher E; Noble, Peter A; Milliken, Charles E; Li, Dien; Kaplan, Daniel I

2018-01-01

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 10 3 to 4 × 10 4 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could spur

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levelsmore » of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 10 3to 4 × 104 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

PubMed Central

Bagwell, Christopher E.; Noble, Peter A.; Milliken, Charles E.; Li, Dien; Kaplan, Daniel I.

2018-01-01

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and environmental DNA was extracted for 454 pyrosequencing of bacterial 16S rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of organic carbon were co-localized with the precipitant relative to bulk dissolved organic concentrations. Bacterial densities were highly variable between sampling locations and with depth within the water filled storage basin; cell numbers ranged from 4 × 103to 4 × 104 cells/mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs/amplicon library (3% cutoff) and the majority of sequences were aligned to the families Burkholderiaceae (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across a selection of natural and engineered aquatic ecosystems were performed and microbial community and taxonomic signatures unique to the spent nuclear fuel (SNF) storage basin environment were revealed. These insights could spur the

78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-18

...-0308] RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear... proposing to amend its spent fuel storage regulations by revising the NAC International, Inc., Modular Advanced Generation Nuclear All-purpose Storage (MAGNASTOR[supreg]) Cask System listing within the ``List...

Spent Fuel Ratio Estimates from Numerical Models in ALE3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Margraf, J. D.; Dunn, T. A.

Potential threat of intentional sabotage of spent nuclear fuel storage facilities is of significant importance to national security. Paramount is the study of focused energy attacks on these materials and the potential release of aerosolized hazardous particulates into the environment. Depleted uranium oxide (DUO 2) is often chosen as a surrogate material for testing due to the unreasonable cost and safety demands for conducting full-scale tests with real spent nuclear fuel. To account for differences in mechanical response resulting in changes to particle distribution it is necessary to scale the DUO 2 results to get a proper measure for spentmore » fuel. This is accomplished with the spent fuel ratio (SFR), the ratio of respirable aerosol mass released due to identical damage conditions between a spent fuel and a surrogate material like depleted uranium oxide (DUO 2). A very limited number of full-scale experiments have been carried out to capture this data, and the oft-questioned validity of the results typically leads to overly-conservative risk estimates. In the present work, the ALE3D hydrocode is used to simulate DUO 2 and spent nuclear fuel pellets impacted by metal jets. The results demonstrate an alternative approach to estimate the respirable release fraction of fragmented nuclear fuel.« less

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, J.A.; Turner, D.W.

1994-12-31

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shieldingmore » Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.« less

High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Jiang, Hao

2015-01-01

The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, 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 from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into localmore » stress-strain profiles for further investigation. 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 SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.« less

Shipment of spent nuclear fuel from U.S. Navy ships and submarines to the Idaho National Engineering Laboratory (INEL). Hearing before the Subcommittee on Nuclear Deterrence, Arms Control and Defense Intelligence of the Committee on Armed Services, United States Senate, One Hundred Third Congress, First Session, July 28, 1993

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1994-12-31

The purpose of the hearing was to review the impact of the U.S. District Court of Idaho ruling prohibiting receipt of spent nuclear fuel by the Department of Energy (DOE). The court`s ruling enjoined the DOE from receiving spent nuclear fuel, including nuclear fuel from naval surface ships and submarines, at the Idaho National Engineering Laboratory until such time as the DOE completes an environmental impact statement on the transportation, shipment, processing, and storage of spent fuel. Statements of government officials are included. The text of the Court ruling is also included.

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... required under this section for shipments of irradiated reactor fuel in quantities less than that subject...

Coccomyxa actinabiotis sp. nov. (Trebouxiophyceae, Chlorophyta), a new green microalga living in the spent fuel cooling pool of a nuclear reactor.

PubMed

Rivasseau, Corinne; Farhi, Emmanuel; Compagnon, Estelle; de Gouvion Saint Cyr, Diane; van Lis, Robert; Falconet, Denis; Kuntz, Marcel; Atteia, Ariane; Couté, Alain

2016-10-01

Life can thrive in extreme environments where inhospitable conditions prevail. Organisms which resist, for example, acidity, pressure, low or high temperature, have been found in harsh environments. Most of them are bacteria and archaea. The bacterium Deinococcus radiodurans is considered to be a champion among all living organisms, surviving extreme ionizing radiation levels. We have discovered a new extremophile eukaryotic organism that possesses a resistance to ionizing radiations similar to that of D. radiodurans. This microorganism, an autotrophic freshwater green microalga, lives in a peculiar environment, namely the cooling pool of a nuclear reactor containing spent nuclear fuels, where it is continuously submitted to nutritive, metallic, and radiative stress. We investigated its morphology and its ultrastructure by light, fluorescence and electron microscopy as well as its biochemical properties. Its resistance to UV and gamma radiation was assessed. When submitted to different dose rates of the order of some tens of mGy · h -1 to several thousands of Gy · h -1 , the microalga revealed to be able to survive intense gamma-rays irradiation, up to 2,000 times the dose lethal to human. The nuclear genome region spanning the genes for small subunit ribosomal RNA-Internal Transcribed Spacer (ITS) 1-5.8S rRNA-ITS2-28S rRNA (beginning) was sequenced (4,065 bp). The phylogenetic position of the microalga was inferred from the 18S rRNA gene. All the revealed characteristics make the alga a new species of the genus Coccomyxa in the class Trebouxiophyceae, which we name Coccomyxa actinabiotis sp. nov. © 2016 Phycological Society of America.

Scientists warn of 'trillion-dollar' spent-fuel risk

NASA Astrophysics Data System (ADS)

Gwynne, Peter

2016-07-01

A study by two Princeton University physicists suggests that a major fire in the spent nuclear fuel stored on the sites of US nuclear reactors could “dwarf the horrific consequences of the Fukushima accident”.

Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

NASA Astrophysics Data System (ADS)

Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

Numerical Estimation of the Spent Fuel Ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric R.; Durbin, Samuel; Wilke, Jason

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on amore » mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank« less

Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit

DOE Office of Scientific and Technical Information (OSTI.GOV)

John H. Strumpell

2004-12-31

Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiationmore » examination (PIE).« less

Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trtilek, Radek; Podlaha, Josef

After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the secondmore » shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)« less

Creating NDA working standards through high-fidelity spent fuel modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E; Gauld, Ian C; Romano, Catherine E

2012-01-01

The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is beingmore » performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from

Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

NASA Astrophysics Data System (ADS)

Butkovich, T. R.

1981-08-01

A generic test of the geologic storage of spent-fuel assemblies from an operating nuclear reactor is being made by the Lawrence Livermore National Laboratory at the US Department of Energy's Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canister, preliminary calculations were made using a pair of existing finite-element codes. Calculational modeling of a spent-fuel repository requires a code with a multiple capability. The effects of both the mining operation and the thermal load on the existing stress fields and the resultant displacements of the rock around the repository must be calculated. The thermal loading for each point in the rock is affected by heat transfer through conduction, radiation, and normal convection, as well as by ventilation of the drifts. Both the ADINA stress code and the compatible ADINAT heat-flow code were used to perform the calculations because they satisfied the requirements of this project. ADINAT was adapted to calculate radiative and convective heat transfer across the drifts and to model the effects of ventilation in the drifts, while the existing isotropic elastic model was used with the ADINA code. The results of the calculation are intended to provide a base with which to compare temperature, stress, and displacement data taken during the planned 5-y duration of the test. In this way, it will be possible to determine how the existing jointing in the rock influences the results as compared with a homogeneous, isotropic rock mass. Later, new models will be introduced into ADINA to account for the effects of jointing.

Small-scale characterisation of irradiated nuclear materials: Part I – Microstructure

DOE PAGES

Edmondson, P. D.; London, A.; Xu, A.; ...

2014-11-26

The behaviour of nanometre-scale precipitates in oxide dispersion strengthened (ODS) ferritic alloys and tungsten-rhenium alloys for nuclear applications has been examined by atom probe tomography (APT). Low Re content tungsten alloys showed no evidence of Re clustering following self-ion irradiation whereas the 25 at.% Re resulted in cluster formation. The size and composition of clusters varied depending on the material form during irradiation (pre-sharpened needle or bulk). Lastly, these results highlight the care that must be taken in interpreting data from ion irradiated pre-sharpened needles due to the presence of free surfaces. Self-ion irradiation of the ODS ferritic alloy resultedmore » in a change in the composition of the clusters, indicating a transition from a near-stoichiometric Y 2Ti 2O 7 composition towards a Ti 2YO 5.« less

Experimental set up for the irradiation of biological samples and nuclear track detectors with UV C

PubMed Central

Portu, Agustina Mariana; Rossini, Andrés Eugenio; Gadan, Mario Alberto; Bernaola, Omar Alberto; Thorp, Silvia Inés; Curotto, Paula; Pozzi, Emiliano César Cayetano; Cabrini, Rómulo Luis; Martin, Gisela Saint

2016-01-01

Aim In this work we present a methodology to produce an “imprint” of cells cultivated on a polycarbonate detector by exposure of the detector to UV C radiation. Background The distribution and concentration of 10B atoms in tissue samples coming from BNCT (Boron Neutron Capture Therapy) protocols can be determined through the quantification and analysis of the tracks forming its autoradiography image on a nuclear track detector. The location of boron atoms in the cell structure could be known more accurately by the simultaneous observation of the nuclear tracks and the sample image on the detector. Materials and Methods A UV C irradiator was constructed. The irradiance was measured along the lamp direction and at different distances. Melanoma cells were cultured on polycarbonate foils, incubated with borophenylalanine, irradiated with thermal neutrons and exposed to UV C radiation. The samples were chemically attacked with a KOH solution. Results A uniform irradiation field was established to expose the detector foils to UV C light. Cells could be seeded on the polycarbonate surface. Both imprints from cells and nuclear tracks were obtained after chemical etching. Conclusions It is possible to yield cellular imprints in polycarbonate. The nuclear tracks were mostly present inside the cells, indicating a preferential boron uptake. PMID:26933396

Container materials in environments of corroded spent nuclear fuel

NASA Astrophysics Data System (ADS)

Huang, F. H.

1996-07-01

Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE PAGES

Jiang, Hao; Wang, Jy-An John

2016-10-14

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Spent nuclear fuel system dynamic stability under normal conditions of transportation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside themore » cask during NCT. In conclusion, dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.« less

Spent Nuclear Fuel Transportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marsha Keister; Kathryn McBride

The Nuclear Waste Policy Act of 1982 (NWPA), as amended, assigned the Department of Energy (DOE) responsibility for developing and managing a Federal system for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for accepting, transporting, and disposing of SNF and HLW at the Yucca Mountain repository in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. OCRWM faces a near-term challenge—to develop and demonstrate a transportation system that will sustain safe and efficient shipments ofmore » SNF and HLW to a repository. To better inform and improve its current planning, OCRWM has extensively reviewed plans and other documents related to past high-visibility shipping campaigns of SNF and other radioactive materials within the United States. This report summarizes the results of this review and, where appropriate, lessons learned.« less

Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

NASA Astrophysics Data System (ADS)

Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

2018-01-01

Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

NASA Astrophysics Data System (ADS)

Johnson, Lawrence; Ferry, Cécile; Poinssot, Christophe; Lovera, Patrick

2005-11-01

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO 2 or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.

Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

NASA Astrophysics Data System (ADS)

Krishna, R.; Jones, A. N.; McDermott, L.; Marsden, B. J.

2015-12-01

Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated 'D'peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of 'G' and 'D' in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.

Watkins: Store spent fuel at federal sites

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tompkins, B.

If the Department of Energy under the Clinton Administration decides to follow the strategy of departing Energy Secretary James Watkins, spent nuclear fuel (SNF) from utilities would be stored at federal sites starting in 1998. In what amounts to an admission that a monitored retrievable storage (MRS) facility will not be ready to accept SNF by 1998, Watkins proposed in a December 17 letter to Senator J. Bennett Johnston that current work on an MRS be terminated, and that efforts be redirected toward standardized spent-fuel container design. According to the DOE's Samantha Williams, however, the Waste Negotiator's Office will continuemore » its activities, and the new strategy is intended to supplement the ongoing MRS work. She also said that DOE Undersecretary Hugo Pomrehn confirmed this in a statement to the Nuclear Waste TEchnical Review Board during its meeting on January 5. The motivation for the new strategic actions is in part related to the passage of the National Energy Policy Act of 1992, which envisions the continued use of nuclear power in the overall US energy plan. Watkins points out, however, that in order for nuclear power to remain a viable alternative, the management and disposal of spent commercial nuclear fuel is essential to avoid premature and unwarranted shutdowns of operating nuclear plants, to permit renewals of existing plant licenses to provide life extensions from 40 to 60 years, and to enable new orders of advanced-design nuclear plants. In addition, the new strategy is required, according to Watkins, because the Office of the Nuclear Waste Negotiator has been unsuccessful in identifying a voluntary host site for an MRS after two years of efforts to that end. David Leroy, the Negotiator, disagrees with this statement, and has noted that it still may be possible to present a candidate host site for the MRS to Congress by June 1993.« less

Enhanced hydrogen production of Enterobacter aerogenes mutated by nuclear irradiation.

PubMed

Cheng, Jun; Liu, Min; Song, Wenlu; Ding, Lingkan; Liu, Jianzhong; Zhang, Li; Cen, Kefa

2017-03-01

Nuclear irradiation was used for the first time to generate efficient mutants of hydrogen-producing bacteria Enterobacter aerogenes, which were screened with larger colour circles of more fermentative acid by-products. E. aerogenes cells were mutated by nuclear irradiation of 60 Co γ-rays. The screened E. aerogenes ZJU1 mutant with larger colour circles enhanced the hydrogenase activity from 89.8 of the wild strain to 157.4mLH 2 /(gDWh). The hereditary stability of the E. aerogenes ZJU1 mutant was certified after over ten generations of cultivation. The hydrogen yield of 301mLH 2 /gglucose with the mutant was higher by 81.8% than that of 166mL/gglucose with the wild strain. The peak hydrogen production rate of 27.2mL/(L·h) with the mutant was higher by 40.9% compared with that of 19.3mL/(L·h) with the wild strain. The mutant produced more acetate and butyrate but less ethanol compared with the wild strain during hydrogen fermentation. Copyright © 2016 Elsevier Ltd. All rights reserved.

Reversible Bending Fatigue Test System for Investigating Vibration Integrity of Spent Nuclear Fuel during Transportation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in themore » area of safety and security of spent nuclear fuel storage and transport operations. The ORNL developed test system can perform reversible-bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot-cell operation, including remote installation and detachment of the SNF test specimen, in-situ test specimen deformation measurement, and implementation of a driving system suitable for use in a hot cell. The system contains a U-frame set-up equipped with uniquely designed grip rigs, to protect SNF rod and to ensure valid test results, and use of 3 specially designed LVDTs to obtain the in-situ curvature measurement. A variety of surrogate test rods have been used to develop and calibrate the test system as well as in performing a series of systematic cyclic fatigue tests. The surrogate rods include stainless steel (SS) cladding, SS cladding with cast epoxy, and SS cladding with alumina pellets inserts simulating fuel pellets. Testing to date has shown that the interface bonding between the SS cladding and the alumina pellets has a significant impact on the bending response of the test rods as well as their fatigue strength. The failure behaviors observed

Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takeuchi, M.; Koizumi, T.; Inoue, M.

2013-07-01

Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution inmore » the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)« less

AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dewes, J.

2014-02-24

The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/Cmore » was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.« less

Spent fuel burnup estimation by Cerenkov glow intensity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuribara, Masayuki

1994-10-01

The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to themore » declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.« less

Influence of neutron irradiation on the microstructure of nuclear graphite: An X-ray diffraction study

NASA Astrophysics Data System (ADS)

Zhou, Z.; Bouwman, W. G.; Schut, H.; van Staveren, T. O.; Heijna, M. C. R.; Pappas, C.

2017-04-01

Neutron irradiation effects on the microstructure of nuclear graphite have been investigated by X-ray diffraction on virgin and low doses (∼ 1.3 and ∼ 2.2 dpa), high temperature (750° C) irradiated samples. The diffraction patterns were interpreted using a model, which takes into account the turbostratic disorder. Besides the lattice constants, the model introduces two distinct coherent lengths in the c-axis and the basal plane, that characterise the volumes from which X-rays are scattered coherently. The methodology used in this work allows to quantify the effect of irradiation damage on the microstructure of nuclear graphite seen by X-ray diffraction. The results show that the changes of the deduced structural parameters are in agreement with previous observations from electron microscopy, but not directly related to macroscopic changes.

MeV per Nucleon Ion Irradiation of Nuclear Materials with High Energy Synchrotron X-ray Characterization

DOE PAGES

Pellin, M. J.; Yacout, Abdellatif M.; Mo, Kun; ...

2016-01-14

The combination of MeV/Nucleon ion irradiation (e.g. 133 MeV Xe) and high energy synchrotron x-ray characterization (e.g. at the Argonne Advanced Photon Source, APS) provides a powerful characterization method to understand radiation effects and to rapidly screen materials for the nuclear reactor environment. Ions in this energy range penetrate ~10 μm into materials. Over this range, the physical interactions vary (electronic stopping, nuclear stopping and added interstitials). Spatially specific x-ray (and TEM and nanoindentation) analysis allow individual quantification of these various effects. Hard x-rays provide the penetration depth needed to analyze even nuclear fuels. Here, this combination of synchrotron x-raymore » and MeV/Nucleon ion irradiation is demonstrated on U-Mo fuels. A preliminary look at HT-9 steels is also presented. We suggest that a hard x-ray facility with in situ MeV/nucleon irradiation capability would substantially accelerate the rate of discovery for extreme materials.« less

Neutron field characterization at the independent spent fuel storage installation of the Trillo nuclear power plant.

PubMed

Campo, Xandra; Méndez, Roberto; Embid, Miguel; Ortego, Alberto; Novo, Manuel; Sanz, Javier

2018-05-01

Neutron fields inside and outside the independent spent fuel storage installation of Trillo Nuclear Power Plant are characterized exhaustively in terms of neutron spectra and ambient dose equivalent, measured by Bonner sphere system and LB6411 monitor. Measurements are consistent with storage casks and building shield characteristics, and also with casks distribution inside the building. Outer values at least five times lower than dose limit for free access area are found. Measurements with LB6411 and spectrometer are consistent with each other. Copyright © 2018 Elsevier Ltd. All rights reserved.

Status of the nuclear measurement stations for the process control of spent fuel reprocessing at AREVA NC/La Hague

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eleon, Cyrille; Passard, Christian; Hupont, Nicolas

2015-07-01

Nuclear measurements are used at AREVA NC/La Hague for the monitoring of spent fuel reprocessing. The process control is based on gamma-ray spectroscopy, passive neutron counting and active neutron interrogation, and gamma transmission measurements. The main objectives are criticality and safety, online process monitoring, and the determination of the residual fissile mass and activities in the metallic waste remained after fuel shearing and dissolution (empty hulls, grids, end pieces), which are put in radioactive waste drums before compaction. The whole monitoring system is composed of eight measurement stations which will be described in this paper. The main measurement stations no.more » 1, 3 and 7 are needed for criticality control. Before fuel element shearing for dissolution, station no. 1 allows determining the burn-up of the irradiated fuel by gamma-ray spectroscopy with HP Ge (high purity germanium) detectors. The burn-up is correlated to the {sup 137}Cs and {sup 134}Cs gamma emission rates. The fuel maximal mass which can be loaded in one bucket of the dissolver is estimated from the lowest burn-up fraction of the fuel element. Station no. 3 is dedicated to the control of the correct fuel dissolution, which is performed with a {sup 137}Cs gamma ray measurement with a HP Ge detector. Station no. 7 allows estimating the residual fissile mass in the drums filled with the metallic residues, especially in the hulls, from passive neutron counting (spontaneous fission and alpha-n reactions) and active interrogation (fission prompt neutrons induced by a pulsed neutron generator) with proportional {sup 3}He detectors. The measurement stations have been validated for the reprocessing of Uranium Oxide (UOX) fuels with a burn-up rate up to 60 GWd/t. This paper presents a brief overview of the current status of the nuclear measurement stations. (authors)« less

Nuclear Energy Policy

DTIC Science & Technology

2007-07-12

Nuclear Waste Storage Act of 2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage ...enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons materials...that had used the leased fuel , along with supplies of fresh nuclear fuel , according to the GNEP concept; see [http://www.gnep.energy.gov].

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Yulan; Hu, Shenyang; Sun, Xin

Complex microstructure changes occur in nuclear fuel and structural materials due to the extreme environments of intense irradiation and high temperature. This paper evaluates the role of the phase field (PF) method in predicting the microstructure evolution of irradiated nuclear materials and the impact on their mechanical, thermal, and magnetic properties. The paper starts with an overview of the important physical mechanisms of defect evolution and the significant gaps in simulating microstructure evolution in irradiated nuclear materials. Then, the PF method is introduced as a powerful and predictive tool and its applications to microstructure and property evolution in irradiated nuclearmore » materials are reviewed. The review shows that 1) FP models can correctly describe important phenomena such as spatial dependent generation, migration, and recombination of defects, radiation-induced dissolution, the Soret effect, strong interfacial energy anisotropy, and elastic interaction; 2) The PF method can qualitatively and quantitatively simulate 2-D and 3-D microstructure evolution, including radiation-induced segregation, second phase nucleation, void migration, void and gas bubble superlattice formation, interstitial loop evolution, hydrate formation, and grain growth, and 3) The FP method correctly predicts the relationships between microstructures and properties. The final section is dedicated to a discussion of the strengths and limitations of the PF method, as applied to irradiation effects in nuclear materials.« less

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ludewigt, Bernhard A; Quiter, Brian J.; Ambers, Scott D.

2011-01-14

The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity ofmore » the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size

Method for storing spent nuclear fuel in repositories

DOEpatents

Schweitzer, Donald G.; Sastre, Cesar; Winsche, Warren

1981-01-01

A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

Method for storing spent nuclear fuel in repositories

DOEpatents

Schweitzer, D.G.; Sastre, C.; Winsche, W.

A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

DOE PAGES

Li, Yulan; Hu, Shenyang; Sun, Xin; ...

2017-04-14

Here, complex microstructure changes occur in nuclear fuel and structural materials due to the extreme environments of intense irradiation and high temperature. This paper evaluates the role of the phase field method in predicting the microstructure evolution of irradiated nuclear materials and the impact on their mechanical, thermal, and magnetic properties. The paper starts with an overview of the important physical mechanisms of defect evolution and the significant gaps in simulating microstructure evolution in irradiated nuclear materials. Then, the phase field method is introduced as a powerful and predictive tool and its applications to microstructure and property evolution in irradiatedmore » nuclear materials are reviewed. The review shows that (1) Phase field models can correctly describe important phenomena such as spatial-dependent generation, migration, and recombination of defects, radiation-induced dissolution, the Soret effect, strong interfacial energy anisotropy, and elastic interaction; (2) The phase field method can qualitatively and quantitatively simulate two-dimensional and three-dimensional microstructure evolution, including radiation-induced segregation, second phase nucleation, void migration, void and gas bubble superlattice formation, interstitial loop evolution, hydrate formation, and grain growth, and (3) The Phase field method correctly predicts the relationships between microstructures and properties. The final section is dedicated to a discussion of the strengths and limitations of the phase field method, as applied to irradiation effects in nuclear materials.« less

A review: applications of the phase field method in predicting microstructure and property evolution of irradiated nuclear materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Yulan; Hu, Shenyang; Sun, Xin

Here, complex microstructure changes occur in nuclear fuel and structural materials due to the extreme environments of intense irradiation and high temperature. This paper evaluates the role of the phase field method in predicting the microstructure evolution of irradiated nuclear materials and the impact on their mechanical, thermal, and magnetic properties. The paper starts with an overview of the important physical mechanisms of defect evolution and the significant gaps in simulating microstructure evolution in irradiated nuclear materials. Then, the phase field method is introduced as a powerful and predictive tool and its applications to microstructure and property evolution in irradiatedmore » nuclear materials are reviewed. The review shows that (1) Phase field models can correctly describe important phenomena such as spatial-dependent generation, migration, and recombination of defects, radiation-induced dissolution, the Soret effect, strong interfacial energy anisotropy, and elastic interaction; (2) The phase field method can qualitatively and quantitatively simulate two-dimensional and three-dimensional microstructure evolution, including radiation-induced segregation, second phase nucleation, void migration, void and gas bubble superlattice formation, interstitial loop evolution, hydrate formation, and grain growth, and (3) The Phase field method correctly predicts the relationships between microstructures and properties. The final section is dedicated to a discussion of the strengths and limitations of the phase field method, as applied to irradiation effects in nuclear materials.« less

SIRT1 interacts with and protects glyceraldehyde-3-phosphate dehydrogenase (GAPDH) from nuclear translocation: Implications for cell survival after irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joo, Hyun-Yoo; Laboratory of Biochemistry, School of Life Sciences and Biotechnology, Korea University, Seoul 136-713; Woo, Seon Rang

2012-08-10

Highlights: Black-Right-Pointing-Pointer SIRT1 serves to retain GAPDH in the cytosol, preventing GAPDH nuclear translocation. Black-Right-Pointing-Pointer When SIRT1 is depleted, GAPDH translocation occurs even in the absence of stress. Black-Right-Pointing-Pointer Upon irradiation, SIRT1 interacts with GAPDH. Black-Right-Pointing-Pointer SIRT1 prevents irradiation-induced nuclear translocation of GAPDH. Black-Right-Pointing-Pointer SIRT1 presence rather than activity is essential for inhibiting GAPDH translocation. -- Abstract: Upon apoptotic stimulation, glyceraldehyde-3-phosphate dehydrogenase (GAPDH), a cytosolic enzyme normally active in glycolysis, translocates into the nucleus and activates an apoptotic cascade therein. In the present work, we show that SIRT1 prevents nuclear translocation of GAPDH via interaction with GAPDH. SIRT1 depletion triggeredmore » nuclear translocation of cytosolic GAPDH even in the absence of apoptotic stress. Such translocation was not, however, observed when SIRT1 enzymatic activity was inhibited, indicating that SIRT1 protein per se, rather than the deacetylase activity of the protein, is required to inhibit GAPDH translocation. Upon irradiation, SIRT1 prevented irradiation-induced nuclear translocation of GAPDH, accompanied by interaction of SIRT1 and GAPDH. Thus, SIRT1 functions to retain GAPDH in the cytosol, protecting the enzyme from nuclear translocation via interaction with these two proteins. This serves as a mechanism whereby SIRT1 regulates cell survival upon induction of apoptotic stress by means that include irradiation.« less

Radiolytic and Thermal Process Relevant to Dry Storage of Spent Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, Steven C.; Haustein, Peter E.; Madey, Theodore E.

1999-06-01

This project involves basic research in chemistry and physics aimed at providing information pertinent to the safe long-term dry storage of spent nuclear fuel (SNF), thousands of tons of which remain in water storage across the DOE complex. The Hanford Site K-Basins alone hold 2300 tons of spent fuel, much of it severely corroded, and similar situations exist at Savannah River and Idaho National Engineering and Environmental Laboratory. DOE plans to remove this fuel and seal it in overpack canisters for ''dry'' interim storage for up to 75 years while awaiting permanent disposition. Chemically bound water will remain in thismore » fuel even after the proposed drying steps, leading to possible long-term corrosion of the containers and/or fuel rods themselves, generation of H2 and O2 gas via radiolysis (which could lead to deflagration or detonation), and reactions of pyrophoric uranium hydrides. No thoroughly tested model is now available to predict fuel behavior during preprocessing, processing, or storage. In a collaborative effort among Rutgers University, Pacific Northwest National Laboratory, and Brookhaven National Laboratory, we are studying the radiolytic reaction, drying processes, and corrosion behavior of actual SNF materials and of pure and mixed-phase samples. We propose to determine what is omitted from current models: radiolysis of water adsorbed on or in hydrates or hydroxides, thermodynamics of interfacial phases, and kinetics of drying. A model will be developed and tested against actual fuel rod behavior to ensure validity and applicability to the problems associated with developing dry storage strategies for DOE-owned SNF.« less

A shielded measurement system for irradiated nuclear fuel measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mosby, W.R.; Aumeier, S.E.; Klann, R.T.

1999-07-01

The US Department of Energy (DOE) is driving a transition toward dry storage of irradiated nuclear fuel (INF), toward characterization of INF for final disposition, and toward resumption of measurement-based material control and accountability (MC and A) efforts for INF. For these reasons, the ability to efficiently acquire radiological measurements of INF in a dry environment is important. The DOE has recently developed a guidance document proposing MC and A requirements for INF. The intent of this document is to encourage the direct measurement of INF on inventory within DOE. The guidance document reinforces and clarifies existing material safeguards requirementsmore » as they pertain to INF. Validation of nuclear material contents of non-self-protecting INF must be accomplished by direct measurement, application of validated burnup codes using qualified initial fissile content, burnup data, and age or by other valid means. The fuel units must remain intact with readable identification numbers. INF may be subject to periodic inventories with visual item accountability checks. Quantitative measurements may provide greater assurance of the integrity of INF inventories at a lower cost and with less personnel exposure than visual item accountability checks. Currently, several different approaches are used to measure the radiological attributes of INF. Although these systems are useful for a wide variety of applications, there is currently no relatively inexpensive measurement system that is readily deployable for INF measurements for materials located in dry storage. The authors present the conceptual design of a shielded measurement system (SMS) that could be used for this purpose. The SMS consists of a shielded enclosure designed to house a collection of measurement systems to allow measurements on spent fuel outside of a hot cell. The phase 1 SMS will contain {sup 3}He detectors and ionization chambers to allow for gross neutron and gamma-ray measurements. The

Technology Implementation Plan: Irradiation Testing and Qualification for Nuclear Thermal Propulsion Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, Thomas J.; Howard, Richard H.; Rader, Jordan D.

This document is a notional technology implementation plan (TIP) for the development, testing, and qualification of a prototypic fuel element to support design and construction of a nuclear thermal propulsion (NTP) engine, specifically its pre-flight ground test. This TIP outlines a generic methodology for the progression from non-nuclear out-of-pile (OOP) testing through nuclear in-pile (IP) testing, at operational temperatures, flows, and specific powers, of an NTP fuel element in an existing test reactor. Subsequent post-irradiation examination (PIE) will occur in existing radiological facilities. Further, the methodology is intended to be nonspecific with respect to fuel types and irradiation or examinationmore » facilities. The goals of OOP and IP testing are to provide confidence in the operational performance of fuel system concepts and provide data to program leadership for system optimization and fuel down-selection. The test methodology, parameters, collected data, and analytical results from OOP, IP, and PIE will be documented for reference by the NTP operator and the appropriate regulatory and oversight authorities. Final full-scale integrated testing would be performed separately by the reactor operator as part of the preflight ground test.« less

75 FR 9452 - Solicitation of Topics for Discussion at a Spent Fuel Storage and Transportation Licensing...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-02

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0076] Solicitation of Topics for Discussion at a Spent...: Solicitation of Topics for Discussion at a Spent Fuel Storage and Transportation Licensing Conference. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is soliciting input on topics for discussion at a...

Neutron radiation characteristics of the IVth generation reactor spent fuel

NASA Astrophysics Data System (ADS)

Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

2018-03-01

Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial

77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-02

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... of pressurized water reactor spent nuclear fuel (SNF) in transportation packages and storage casks... for the licensing basis, (b) provide recommendations regarding advanced isotopic depletion and...

Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

NASA Astrophysics Data System (ADS)

Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

2017-09-01

Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

Estimation of 85Kr dispersion from the spent nuclear fuel reprocessing plant in Rokkasho, Japan, using an atmospheric dispersion model.

PubMed

Abe, K; Iyogi, T; Kawabata, H; Chiang, J H; Suwa, H; Hisamatsu, S

2015-11-01

The spent nuclear fuel reprocessing plant of Japan Nuclear Fuel Limited (JNFL) located in Rokkasho, Japan, discharged small amounts of (85)Kr into the atmosphere during final tests of the plant with actual spent fuel from 31 March 2006 to October 2008. During this period, the gamma-ray dose rates due to discharged (85)Kr were higher than the background rates measured at the Institute for Environmental Sciences and at seven monitoring stations of the Aomori prefectural government and JNFL. The dispersion of (85)Kr was simulated by means of the fifth-generation Penn State/NCAR Mesoscale Model and the CG-MATHEW/ADPIC models (ver. 5.0) with a vertical terrain-following height coordinate. Although the simulated gamma-ray dose rates due to discharged (85)Kr agreed fairly well with measured rates, the agreement between the estimated monthly mean (85)Kr concentrations and the observed concentrations was poor. Improvement of the vertical flow of air may lead to better estimation of (85)Kr dispersion. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Geringer, J. W.; Katoh, Yutai; Howard, Richard H.

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterizationmore » of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.« less

75 FR 27401 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1; Correction

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-17

... Storage Casks: NUHOMS[reg] HD System Revision 1; Correction AGENCY: Nuclear Regulatory Commission. ACTION... HD spent fuel storage cask system. This action is necessary to correctly specify the effective date... on May 6, 2010 (75 FR 24786), that amends the regulations that govern storage of spent nuclear fuel...

An experience of electron beam (EB) irradiated gemstones in Malaysian nuclear agency

NASA Astrophysics Data System (ADS)

Idris, Sarada; Hairaldin, Siti Zulaiha; Tajau, Rida; Karim, Jamilah; Jusoh, Suhaimi; Ghazali, Zulkafli; Ahmad, Shamshad

2014-02-01

In Nuclear Malaysia, a study on gemstone irradiation using beta particle is conducted. The purpose of the study is to evaluate the gemstone colour enhancement by using different kind of precious and non-precious gemstones. By using irradiation technique, selected gemstones are exposed to highly ionizing radiation electron beam to knock off electrons to generate colour centres culminating in the introduction of deeper colours. The colour centres may be stable or unstable depending on the nature of colour centre produced. The colour change of irradiated stones were measured by HunterLab colour measurement. At 50 kGy, Topaz shows changes colour from colourless to golden. Meanwhile pearl shows changes from pale colour to grey. Kunzite and amethyst shows colour changes from colorless to green and pale colour to purple. Gamma survey meter measurement confirmed that irradiation treatment with 3 MeV electron beam machine does not render any activation that activate the gems to become radioactive.

An experience of electron beam (EB) irradiated gemstones in Malaysian nuclear agency

DOE Office of Scientific and Technical Information (OSTI.GOV)

Idris, Sarada, E-mail: sarada@nuclearmalaysia.gov.my; Hairaldin, Siti Zulaiha, E-mail: sarada@nuclearmalaysia.gov.my; Tajau, Rida, E-mail: sarada@nuclearmalaysia.gov.my

In Nuclear Malaysia, a study on gemstone irradiation using beta particle is conducted. The purpose of the study is to evaluate the gemstone colour enhancement by using different kind of precious and non-precious gemstones. By using irradiation technique, selected gemstones are exposed to highly ionizing radiation electron beam to knock off electrons to generate colour centres culminating in the introduction of deeper colours. The colour centres may be stable or unstable depending on the nature of colour centre produced. The colour change of irradiated stones were measured by HunterLab colour measurement. At 50 kGy, Topaz shows changes colour from colourlessmore » to golden. Meanwhile pearl shows changes from pale colour to grey. Kunzite and amethyst shows colour changes from colorless to green and pale colour to purple. Gamma survey meter measurement confirmed that irradiation treatment with 3 MeV electron beam machine does not render any activation that activate the gems to become radioactive.« less

77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

DOE PAGES

Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; ...

2016-06-15

We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δk eff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

Development of irradiation capabilities to address the challenges of the nuclear industry

NASA Astrophysics Data System (ADS)

Leay, L.; Bower, W.; Horne, G.; Wady, P.; Baidak, A.; Pottinger, M.; Nancekievill, M.; Smith, A. D.; Watson, S.; Green, P. R.; Lennox, B.; LaVerne, J. A.; Pimblott, S. M.

2015-01-01

With the announcement of the U.K. new nuclear build and the requirement to decommission old facilities, researchers require bespoke facilities to undertake experiments to inform decision making. This paper describes development of The University of Manchester's Dalton Cumbrian Facility, a custom built research environment which incorporates a 5 MV tandem ion accelerator as well as a self-shielded 60Co irradiator. The ion accelerator allows the investigation into the radiolytic consequences of various charged particles, including protons, alpha particles and a variety of heavier (metal and nonmetal) ions, while the 60Co irradiator allows the effects of gamma radiation to be studied. Some examples of work carried out at the facility are presented to demonstrate how this equipment can improve our mechanistic understanding of various aspects of the deleterious effects of radiation in the nuclear industry. These examples include applications in waste storage and reprocessing as well as geological storage and novel surveying techniques. The outlook for future research is also discussed.

Nuclear Power Plant Security and Vulnerabilities

DTIC Science & Technology

2009-03-18

Commercial Spent Nuclear Fuel Storage , Public Report...systems that prevent hot nuclear fuel from melting even after the chain reaction has stopped, and storage facilities for highly radioactive spent nuclear ... nuclear fuel cycle facilities must defend against to prevent radiological sabotage and theft of strategic special nuclear material. NRC licensees use

78 FR 32077 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-29

... Fuel Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct... final rule that would have revised its spent fuel storage regulations to include Amendment No. 3 to... All-purpose Storage (MAGNASTOR[supreg]) System listing within the ``List of Approved Spent Fuel...

Modeling evaporation from spent nuclear fuel storage pools: A diffusion approach

NASA Astrophysics Data System (ADS)

Hugo, Bruce Robert

Accurate prediction of evaporative losses from light water reactor nuclear power plant (NPP) spent fuel storage pools (SFPs) is important for activities ranging from sizing of water makeup systems during NPP design to predicting the time available to supply emergency makeup water following severe accidents. Existing correlations for predicting evaporation from water surfaces are only optimized for conditions typical of swimming pools. This new approach modeling evaporation as a diffusion process has yielded an evaporation rate model that provided a better fit of published high temperature evaporation data and measurements from two SFPs than other published evaporation correlations. Insights from treating evaporation as a diffusion process include correcting for the effects of air flow and solutes on evaporation rate. An accurate modeling of the effects of air flow on evaporation rate is required to explain the observed temperature data from the Fukushima Daiichi Unit 4 SFP during the 2011 loss of cooling event; the diffusion model of evaporation provides a significantly better fit to this data than existing evaporation models.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-18

... Storage Casks: MAGNASTOR[supreg] System AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced Generation Nuclear All-purpose Storage...

78 FR 40199 - Draft Spent Fuel Storage and Transportation Interim Staff Guidance

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0140] Draft Spent Fuel Storage and Transportation Interim... Spent Fuel Storage and Transportation Interim Staff Guidance No. 24 (SFST-ISG-24), Revision 0, ``The Use of a Demonstration Program as Confirmation of Integrity for Continued Storage of High Burnup Fuel...

Doses from external irradiation to Marshall Islanders from Bikini and Enewetak nuclear weapons tests.

PubMed

Bouville, André; Beck, Harold L; Simon, Steven L

2010-08-01

Annual doses from external irradiation resulting from exposure to fallout from the 65 atmospheric nuclear weapons tests conducted in the Marshall Islands at Bikini and Enewetak between 1946 and 1958 have been estimated for the first time for Marshallese living on all inhabited atolls. All tests that deposited fallout on any of the 23 inhabited atolls or separate reef islands have been considered. The methodology used to estimate the radiation doses at the inhabited atolls is based on test- and location-specific radiation survey data, deposition density estimates of 137Cs, and fallout times-of-arrival provided in a companion paper (Beck et al.), combined with information on the radionuclide composition of the fallout at various times after each test. These estimates of doses from external irradiation have been combined with corresponding estimates of doses from internal irradiation, given in a companion paper (Simon et al.), to assess the cancer risks among the Marshallese population (Land et al.) resulting from exposure to radiation from the nuclear weapons tests.

Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

NASA Astrophysics Data System (ADS)

Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

2016-12-01

Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO2) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine (131/129I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50-60 MW d/kgU, and with increasing burn-up, the IRF of 131/129I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO2 fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from 239Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO2 nuclear fuels may reduce the impact of long-lived 129I on the repository performance assessment calculations.

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregson, Michael Warren; Mo, Tin; Sorenson, Ken Bryce

The authors provide a detailed overview of an on-going, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy-density device. The program participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research program. Sandia National Laboratoriesmore » has the lead role for conducting this research program; test program support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. The authors provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. They focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests, and briefly summarize similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods.« less

Applications in Nuclear Energy Security

NASA Astrophysics Data System (ADS)

Sheffield, Richard

2009-05-01

A key roadblock to development of additional nuclear power capacity is a concern over management of nuclear waste. Nuclear waste is predominantly comprised of used fuel discharged from operating nuclear reactors. The roughly 100 operating US reactors currently produce about 20% of the US electricity and will create about 87,000 tons of such discharged or ``spent'' fuel over the course of their lifetimes. The long-term radioactivity of the spent fuel drives the need for deep geologic storage that remains stable for millions of years. Nearly all issues related to risks to future generations arising from long-term disposal of such spent nuclear fuel is attributable to approximately the 1% made up primarily of minor actinides. If we can reduce or eliminate this 1% of the spent fuel, then within a few hundred years the toxic nature of the spent fuel drops below that of the natural uranium ore that was originally mined for nuclear fuel. The minor actinides can be efficiently eliminated through nuclear transmutation using as a driver fast-neutrons produced by a spallation process initiated with a high-energy proton beam. This presentation will cover the system design considerations and issues of an accelerator driven transmutation system.

Is Yucca Mountain a long-term solution for disposing of US spent nuclear fuel and high-level radioactive waste?

PubMed

Thorne, M C

2012-06-01

On 26 January 2012, the Blue Ribbon Commission on America's Nuclear Future released a report addressing, amongst other matters, options for the managing and disposal of high-level waste and spent fuel. The Blue Ribbon Commission was not chartered as a siting commission. Accordingly, it did not evaluate Yucca Mountain or any other location as a potential site for the storage or disposal of spent nuclear fuel and high-level waste. Nevertheless, if the Commission's recommendations are followed, it is clear that any future proposals to develop a repository at Yucca Mountain would require an extended period of consultation with local communities, tribes and the State of Nevada. Furthermore, there would be a need to develop generally applicable regulations for disposal of spent fuel and high-level radioactive waste, so that the Yucca Mountain site could be properly compared with alternative sites that would be expected to be identified in the initial phase of the site-selection process. Based on what is now known of the conditions existing at Yucca Mountain and the large number of safety, environmental and legal issues that have been raised in relation to the DOE Licence Application, it is suggested that it would be imprudent to include Yucca Mountain in a list of candidate sites for future evaluation in a consent-based process for site selection. Even if there were a desire at the local, tribal and state levels to act as hosts for such a repository, there would be enormous difficulties in attempting to develop an adequate post-closure safety case for such a facility, and in showing why this unsaturated environment should be preferred over other geological contexts that exist in the USA and that are more akin to those being studied and developed in other countries.

Flaw Stability Considering Residual Stress for Aging Management of Spent Nuclear Fuel Multiple-Purpose Canisters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lam, Poh-Sang; Sindelar, Robert L.

A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. Because heat treatment for stress relief is not required for the construction of the MPC, the canister is susceptible to stress corrosion cracking in the weld or heat affected zone regions under long-term storage conditions. Logic for flaw acceptance is developed should crack-like flaws be detected by Inservice Inspection. The procedure recommended by API 579-1/ASME FFS-1, Fitness-for-Service, is used to calculate the instability crack length or depth by failure assessment diagram. It is demonstrated that the welding residual stress has amore » strong influence on the results.« less

Flaw Stability Considering Residual Stress for Aging Management of Spent Nuclear Fuel Multiple-Purpose Canisters

DOE PAGES

Lam, Poh-Sang; Sindelar, Robert L.

2016-04-28

A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. Because heat treatment for stress relief is not required for the construction of the MPC, the canister is susceptible to stress corrosion cracking in the weld or heat affected zone regions under long-term storage conditions. Logic for flaw acceptance is developed should crack-like flaws be detected by Inservice Inspection. The procedure recommended by API 579-1/ASME FFS-1, Fitness-for-Service, is used to calculate the instability crack length or depth by failure assessment diagram. It is demonstrated that the welding residual stress has amore » strong influence on the results.« less

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

Roles of Radiolytic and Externally Generated H2 in the Corrosion of Fractured Spent Nuclear Fuel.

PubMed

Liu, Nazhen; Wu, Linda; Qin, Zack; Shoesmith, David W

2016-11-15

A 2-D model for the corrosion of spent nuclear fuel inside a failed nuclear waste container has been modified to determine the influence of various redox processes occurring within fractures in the fuel. The corrosion process is driven by reaction of the fuel with the dominant α radiolysis product, H 2 O 2 . A number of reactions are shown to moderate or suppress the corrosion rate, including H 2 O 2 decomposition and a number of reactions involving dissolved H 2 produced either by α radiolysis or by the corrosion of the steel container vessel. Both sources of H 2 lead to the suppression of fuel corrosion, with their relative importance being determined by the radiation dose rate, the steel corrosion rate, and the dimensions of the fractures in the fuel. The combination of H 2 from these two sources can effectively prevent corrosion when only micromolar quantities of H 2 are present.

324 Building spent fuel segments pieces and fragments removal summary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

SMITH, C L

2003-01-09

As part of the 324 Building Deactivation Project, all Spent Nuclear Fuel (SNF) and Special Nuclear Material were removed. The removal entailed packaging the material into a GNS-12 cask and shipping it to the Central Waste Complex (CWC).

Development of Techniques for Spent Fuel Assay – Differential Dieaway Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; Goodsell, Alison; Ianakiev, Kiril Dimitrov

This report summarizes the work done under a DNDO R&D funded project on the development of the differential dieaway method to measure plutonium in spent fuel. There are large amounts of plutonium that are contained in spent fuel assemblies, and currently there is no way to make quantitative non-destructive assay. This has led NA24 under the Next Generation Safeguards Initiative (NGSI) to establish a multi-year program to investigate, develop and implement measurement techniques for spent fuel. The techniques which are being experimentally tested by the existing NGSI project do not include any pulsed neutron active techniques. The present work coversmore » the active neutron differential dieaway technique and has advanced the state of knowledge of this technique as well as produced a design for a practical active neutron interrogation instrument for spent fuel. Monte Carlo results from the NGSI effort show that much higher accuracy (1-2%) for the Pu content in spent fuel assemblies can be obtained with active neutron interrogation techniques than passive techniques, and this would allow their use for nuclear material accountancy independently of any information from the operator. The main purpose of this work was to develop an active neutron interrogation technique for spent nuclear fuel.« less

Fast Neutron Emission Tomography of Used Nuclear Fuel Assemblies

NASA Astrophysics Data System (ADS)

Hausladen, Paul; Iyengar, Anagha; Fabris, Lorenzo; Yang, Jinan; Hu, Jianwei; Blackston, Matthew

2017-09-01

Oak Ridge National Laboratory is developing a new capability to perform passive fast neutron emission tomography of spent nuclear fuel assemblies for the purpose of verifying their integrity for international safeguards applications. Most of the world's plutonium is contained in spent nuclear fuel, so it is desirable to detect the diversion of irradiated fuel rods from an assembly prior to its transfer to ``difficult to access'' storage, such as a dry cask or permanent repository, where re-verification is practically impossible. Nuclear fuel assemblies typically consist of an array of fuel rods that, depending on exposure in the reactor and consequent ingrowth of 244Cm, are spontaneous sources of as many as 109 neutrons s-1. Neutron emission tomography uses collimation to isolate neutron activity along ``lines of response'' through the assembly and, by combining many collimated views through the object, mathematically extracts the neutron emission from each fuel rod. This technique, by combining the use of fast neutrons -which can penetrate the entire fuel assembly -and computed tomography, is capable of detecting vacancies or substitutions of individual fuel rods. This paper will report on the physics design and component testing of the imaging system. This material is based upon work supported by the U.S. Department of Energy, Office of Defense Nuclear Nonproliferation Research and Development within the National Nuclear Security Administration, under Contract Number DE-AC05-00OR22725.

DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sindelar, R.; Deible, R.

2011-04-27

The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report bymore » the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall

Nuclear Energy Policy

DTIC Science & Technology

2008-01-28

2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage casks and then convey title to the Secretary of Energy...far more economical options for reducing fossil fuel use .15 (For more on federal incentives and the economics of nuclear power, see CRS Report RL33442...uranium enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons

Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amharrak, H.; Reynard-Carette, C.; Carette, M.

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR),more » under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the

Portable instrument for inspecting irradiated nuclear fuel assemblies

DOEpatents

Nicholson, Nicholas; Dowdy, Edward J.; Holt, David M.; Stump, Jr., Charles J.

1985-01-01

A portable instrument for measuring induced Cerenkov radiation associated with irradiated nuclear fuel assemblies in a water-filled storage pond is disclosed. The instrument includes a photomultiplier tube and an image intensifier which are operable in parallel and simultaneously by means of a field lens assembly and an associated beam splitter. The image intensifier permits an operator to aim and focus the apparatus on a submerged fuel assembly. Once the instrument is aimed and focused, an illumination reading can be obtained with the photomultiplier tube. The instrument includes a lens cap with a carbon-14/phosphor light source for calibrating the apparatus in the field.

Determination of 90Sr / 238U ratio by double isotope dilution inductively coupled plasma mass spectrometer with multiple collection in spent nuclear fuel samples with in situ 90Sr / 90Zr separation in a collision-reaction cell

NASA Astrophysics Data System (ADS)

Isnard, H.; Aubert, M.; Blanchet, P.; Brennetot, R.; Chartier, F.; Geertsen, V.; Manuguerra, F.

2006-02-01

Strontium-90 is one of the most important fission products generated in nuclear industry. In the research field concerning nuclear waste disposal in deep geological environment, it is necessary to quantify accurately and precisely its concentration (or the 90Sr / 238U atomic ratio) in irradiated fuels. To obtain accurate analysis of radioactive 90Sr, mass spectrometry associated with isotope dilution is the most appropriated method. But, in nuclear fuel samples the interference with 90Zr must be previously eliminated. An inductively coupled plasma mass spectrometer with multiple collection, equipped with an hexapole collision cell, has been used to eliminate the 90Sr / 90Zr interference by addition of oxygen in the collision cell as a reactant gas. Zr + ions are converted into ZrO +, whereas Sr + ions are not reactive. A mixed solution, prepared from a solution of enriched 84Sr and a solution of enriched 235U was then used to quantify the 90Sr / 238U ratio in spent fuel sample solutions using the double isotope dilution method. This paper shows the results, the reproducibility and the uncertainties that can be obtained with this method to quantify the 90Sr / 238U atomic ratio in an UOX (uranium oxide) and a MOX (mixed oxide) spent fuel samples using the collision cell of an inductively coupled plasma mass spectrometer with multiple collection to perform the 90Sr / 90Zr separation. A comparison with the results obtained by inductively coupled plasma mass spectrometer with multiple collection after a chemical separation of strontium from zirconium using a Sr spec resin (Eichrom) has been performed. Finally, to validate the analytical procedure developed, measurements of the same samples have been performed by thermal ionization mass spectrometry, used as an independent technique, after chemical separation of Sr.

Optimization of spent fuel pool weir gate driving mechanism

NASA Astrophysics Data System (ADS)

Liu, Chao; Du, Lin; Tao, Xinlei; Wang, Shijie; Shang, Ertao; Yu, Jianjiang

2018-04-01

Spent fuel pool is crucial facility for fuel storage and nuclear safety, and the spent fuel pool weir gate is the key related equipment. In order to achieve a goal of more efficient driving force transfer, loading during the opening/closing process is analyzed and an optimized calculation method for dimensions of driving mechanism is proposed. The result of optimizing example shows that the method can be applied to weir gates' design with similar driving mechanism.

Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spentmore » nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Hypothetical Case and Scenario Description for International Transportation of Spent Nuclear Fuel.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Adam David; Osborn, Douglas; Jones, Katherine A.

To support more rigorous analysis on global security issues at Sandia National Laboratories (SNL), there is a need to develop realistic data sets without using "real" data or identifying "real" vulnerabilities, hazards or geopolitically embarrassing shortcomings. In response, an interdisciplinary team led by subject matter experts in SNL's Center for Global Security and Cooperation (CGSC) developed a hypothetical case description. This hypothetical case description assigns various attributes related to international SNF transportation that are representative, illustrative and indicative of "real" characteristics of "real" countries. There is no intent to identify any particular country and any similarity with specific real-world eventsmore » is purely coincidental. To support the goal of this report to provide a case description (and set of scenarios of concern) for international SNF transportation inclusive of as much "real-world" complexity as possible -- without crossing over into politically sensitive or classified information -- this SAND report provides a subject matter expert-validated (and detailed) description of both technical and political influences on the international transportation of spent nuclear fuel. [PAGE INTENTIONALLY LEFT BLANK]« less

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2012 CFR

2012-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2014 CFR

2014-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

10 CFR 72.214 - List of approved spent fuel storage casks.

Code of Federal Regulations, 2013 CFR

2013-01-01

... specified in their Certificates of Compliance. Certificate Number: 1000 SAR Submitted by: General Nuclear Systems, Inc. SAR Title: Topical Safety Analysis Report for the Castor V/21 Cask Independent Spent Fuel... Model Number: CASTOR V/21 Certificate Number: 1002 SAR Submitted by: Nuclear Assurance Corporation SAR...

Ultra-High Rate Measurements of Spent Fuel Gamma-Ray Emissions

NASA Astrophysics Data System (ADS)

Rodriguez, Douglas; Vandevender, Brent; Wood, Lynn; Glasgow, Brian; Taubman, Matthew; Wright, Michael; Dion, Michael; Pitts, Karl; Runkle, Robert; Campbell, Luke; Fast, James

2014-03-01

Presently there are over 200,000 irradiated spent nuclear fuel (SNF) assemblies in the world, each containing a concerning amount of weapons-usable material. Both facility operators and safeguards inspectors want to improve composition determination. Current measurements are expensive and difficult so new methods are developed through models. Passive measurements are limited since a few specific decay products and the associated down-scatter overwhelm the gamma rays of interest. Active interrogation methods produce gamma rays beyond 3 MeV, minimizing the impact of the passive emissions that drop off sharply above this energy. New devices like the Ultra-High Rate Germanium (UHRGe) detector are being developed to advance these novel measurement methods. Designed for reasonable resolution at 106 s-1 output rates (compared to ~ 1 - 10 e 3 s-1 standards), SNF samples were directly measured using UHRGe and compared to models. Model verification further enables using Los Alamos National Laboratory SNF assembly models, developed under the Next Generation Safeguards Initiative, to determine emission and signal expectations. Measurement results and future application requirements for UHRGe will be discussed.

Present experience of NRI REZ with preparation of spent nuclear fuel shipment to Russian Federation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Svitak, F.; Broz, V.; Hrehor, M.

2008-07-15

The Nuclear Research Institute Rez plc (NRI) jointed the Russian Research Reactor Fuel Return (RRRFR) programme under the US-Russian Global Threat Reduction Initiative (GTRI) initiative and started the preparation of the spent nuclear fuel (SNF) shipment from the LVR-15 research reactor back to the Russian Federation (RF). The transport of 16 SKODA VPVR/M casks with EK-10, IRT-2M 80 %, and IRT-2M 36% fuel types is planned for the autumn of 2007. The paper describes the experience gained so far during the preparatory works for the SNF shipment (facility equipment modification, cask licenses) and the actual preparation of the SNF formore » transport, in particular its checking, repacking in a hot cell, loading into the VPVR/M casks, drying, manipulation, completion of the transport documentation, etc., including its transport to the SNF storage facility at the NRI before it is shipped to the RF. The paper also briefly describes a regulatory framework for these activities with a focus on legislative and methodological aspects of the return of vitrified waste back to the Czech Republic. (author)« less

International remote monitoring project Argentina Nuclear Power Station Spent Fuel Transfer Remote Monitoring System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, S.; Lucero, R.; Glidewell, D.

1997-08-01

The Autoridad Regulataria Nuclear (ARN) and the United States Department of Energy (DOE) are cooperating on the development of a Remote Monitoring System for nuclear nonproliferation efforts. A Remote Monitoring System for spent fuel transfer will be installed at the Argentina Nuclear Power Station in Embalse, Argentina. The system has been designed by Sandia National Laboratories (SNL), with Los Alamos National Laboratory (LANL) and Oak Ridge National Laboratory (ORNL) providing gamma and neutron sensors. This project will test and evaluate the fundamental design and implementation of the Remote Monitoring System in its application to regional and international safeguards efficiency. Thismore » paper provides a description of the monitoring system and its functions. The Remote Monitoring System consists of gamma and neutron radiation sensors, RF systems, and video systems integrated into a coherent functioning whole. All sensor data communicate over an Echelon LonWorks Network to a single data logger. The Neumann DCM 14 video module is integrated into the Remote Monitoring System. All sensor and image data are stored on a Data Acquisition System (DAS) and archived and reviewed on a Data and Image Review Station (DIRS). Conventional phone lines are used as the telecommunications link to transmit on-site collected data and images to remote locations. The data and images are authenticated before transmission. Data review stations will be installed at ARN in Buenos Aires, Argentina, ABACC in Rio De Janeiro, IAEA Headquarters in Vienna, and Sandia National Laboratories in Albuquerque, New Mexico. 2 refs., 2 figs.« less

Spent fuel behavior under abnormal thermal transients during dry storage

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stahl, D.; Landow, M.P.; Burian, R.J.

1986-01-01

This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment wasmore » heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.« less

Heavy component of spent nuclear fuel: Efficiency of model-substance ionization by electron-induced discharge

DOE Office of Scientific and Technical Information (OSTI.GOV)

Antonov, N. N., E-mail: antonovnickola@gmail.com; Gavrikov, A. V.; Samokhin, A. A.

The method of plasma separation of spent nuclear fuel can be tested with a model substance which has to be transformed from the condensed to plasma state. For this purpose, electron-induced discharge in lead vapor injected into the interelectrode gap is simulated using the kinetic approach. The ionization efficiency, the electrostatic-potential distribution, and those of the ion and electron densities in the discharge gap are derived as functions of the discharge-current density and concentration of the vapor of the model substance. Given a discharge-current density of 3.5 A/cm{sup 2} and a lead-vapor concentration of 2 × 10{sup 12} cm{sup –3},more » the simulated ionization efficiency proves to be nearly 60%. The discharge in lead vapor is also investigated experimentally.« less

Measuring spent fuel assembly multiplication in borated water with a passive neutron albedo reactivity instrument

NASA Astrophysics Data System (ADS)

Tobin, Stephen J.; Peura, Pauli; Bélanger-Champagne, Camille; Moring, Mikael; Dendooven, Peter; Honkamaa, Tapani

2018-07-01

The performance of a passive neutron albedo reactivity (PNAR) instrument to measure neutron multiplication of spent nuclear fuel in borated water is investigated as part of an integrated non-destructive assay safeguards system. To measure the PNAR Ratio, which is proportional to the neutron multiplication, the total neutron count rate is measured in high- and low-multiplying environments by the PNAR instrument. The integrated system also contains a load cell and a passive gamma emission tomograph, and as such meets all the recommendations of the IAEA's recent ASTOR Experts Group report. A virtual spent fuel library for VVER-440 fuel was used in conjunction with MCNP simulations of the PNAR instrument to estimate the measurement uncertainties from (1) variation in the water boron content, (2) assembly positioning in the detector and (3) counting statistics. The estimated aggregate measurement uncertainty on the PNAR Ratio measurement is 0.008, to put this uncertainty in context, the difference in the PNAR Ratio between a fully irradiated assembly and this same assembly when fissile isotopes only absorb neutrons, but do not emit neutrons, is 0.106, a 13-sigma effect. The 1-sigma variation of 0.008 in the PNAR Ratio is estimated to correspond to a 3.2 GWd/tU change in assembly burnup.

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE PAGES

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.; ...

2015-10-09

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

Californium interrogation prompt neutron (CIPN) instrument for non-destructive assay of spent nuclear fuel – design concept and experimental demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henzlova, Daniela; Menlove, Howard Olsen; Rael, Carlos D.

Our paper presents results of the first experimental demonstration of the Californium Interrogation Prompt Neutron (CIPN) instrument developed within a multi-year effort launched by the Next Generation Safeguards Initiative Spent Fuel Project of the United States Department of Energy. The goals of this project focused on developing viable non-destructive assay techniques with capabilities to improve an independent verification of spent fuel assembly characteristics. For this purpose, the CIPN instrument combines active and passive neutron interrogation, along with passive gamma-ray measurements, to provide three independent observables. We describe the initial feasibility demonstration of the CIPN instrument, which involved measurements of fourmore » pressurized-water-reactor spent fuel assemblies with different levels of burnup and two initial enrichments. The measurements were performed at the Post-Irradiation Examination Facility at the Korea Atomic Energy Institute in the Republic of Korea. The key aim of the demonstration was to evaluate CIPN instrument performance under realistic deployment conditions, with the focus on a detailed assessment of systematic uncertainties that are best evaluated experimentally. The measurements revealed good positioning reproducibility, as well as a high degree of insensitivity of the CIPN instrument's response to irregularities in a radial burnup profile. Systematic uncertainty of individual CIPN instrument signals due to assembly rotation was found to be <4.5%, even for assemblies with fairly extreme gradients in the radial burnup profile. Lastly, these features suggest that the CIPN instrument is capable of providing a good representation of assembly average characteristics, independent of assembly orientation in the instrument.« less

77 FR 28406 - Spent Fuel Transportation Risk Assessment

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-14

... Regulations (10 CFR) part 71, ``Packaging and Transportation of Radioactive Waste,'' dated January 26, 2004) for the packaging and transport of spent nuclear fuel (and other large quantities of radioactive... NUREG- 0170, ``Final Environmental Statement on the Transportation of Radioactive Material by Air and...

Piezoelectric wafer active sensors under gamma radiation exposure toward applications for structural health monitoring of nuclear dry cask storage systems

NASA Astrophysics Data System (ADS)

Faisal Haider, Mohammad; Mei, Hanfei; Lin, Bin; Yu, Lingyu; Giurgiutiu, Victor; Lam, Poh-Sang; Verst, Christopher

2018-03-01

Structural health monitoring (SHM) is in urgent need and must be integrated into the nuclear-spent fuel storage systems to guarantee the safe operation. The dry cask storage system (DCSS) is such storage facility, which is licensed for temporary storage for nuclear-spent fuel at the independent spent fuel storage installations (ISFSIs) for certain predetermined period of time. Gamma radiation is one of the major radiation sources near DCSS. Therefore, a detailed experimental investigation was completed on the gamma radiation endurance of piezoelectric wafer active sensors (PWAS) transducers for SHM applications to the DCSS system. The irradiation test was done in a Co-60 gamma irradiator. Lead Zirconate Titanate (PZT) and Gallium Orthophosphate (GaPO4) PWAS transducers were exposed to 40.7 kGy gamma radiation. Total radiation dose was achieved in two different radiation dose rates: (a) slower radiation rate at 0.1 kGy/hr for 20 hours (b) accelerated radiation rate at 1.233 kGy/hr for 32 hours. The total cumulative radiation dose of 40.7 kGy is equivalent to 45 years of operation in DCSS system. Electro-mechanical impedance and admittance (EMIA) signatures and electrical capacitance were measured to evaluate the PWAS performance after each gamma radiation exposure. The change in resonance frequency of PZT-PWAS transducer for both in-plane and thickness mode was observed. The GaPO4-PWAS EMIA spectra do not show a significant shift in resonance frequency after gamma irradiation exposure. Radiation endurance of new high-temperature HPZ-HiT PWAS transducer was also evaluated. The HPZ-HiT transducers were exposed to gamma radiation at 1.233 kGy/hr for 160 hours with 80 hours interval. Therefore, the total accumulated gamma radiation dose is 184 kGy. No significant change in impedance spectra was observed due to gamma radiation exposure.

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE PAGES

Ham, Y.; Kerr, P.; Sitaraman, S.; ...

2016-05-05

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.S.; Kerr, P.; Sitaraman, S.

The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, Y.; Kerr, P.; Sitaraman, S.

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dilger, Fred; Halstead, Robert J.; Ballard, James D.

2013-07-01

Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for themore » selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail routes for

Ion-Neutron Irradiated BOR60 Sample Preparation and Characterization: Nuclear Science User Facility 2017 Milestone Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linton, Kory D.; Parish, Chad M.; Smith, Quinlan B.

2017-09-01

This document outlines the results obtained by Oak Ridge National Laboratory (ORNL) in collaboration with the University of Michigan-led Consolidated Innovative Nuclear Research project, “Feasibility of combined ion-neutron irradiation for accessing high dose levels.” In this reporting period, neutron irradiated were prepared and shipped to the University of Michigan for subsequent ion irradiation. The specimens were returned to ORNL’s Low Activation Materials Development and Analysis facility, prepared via focused ion beam for examination using scanning/transmission electron microscopy (S/TEM), and then examined using S/TEM to measure the as-irradiated microstructure. This report briefly summarizes the S/TEM results obtained at ORNL’s Low Activationmore » Materials Development and Analysis facility.« less

Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

NASA Astrophysics Data System (ADS)

Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

2011-06-01

Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

Compensation as Means for Local Acceptance The Case of the Final Disposal of Spent Nuclear Fuel in Eurajoki, Finland

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kojo, M.

The paper sheds light on the local negotiations on compensation as a part of the site selection for the spent nuclear fuel repository in Finland. The negotiation took place between the representatives of the Municipality of Eurajoki, the nuclear power company Teollisuuden Voima Ltd (TVO) and the nuclear waste management company Posiva Ltd in the late 1990's. The compensation negotiation process and the development of the requirements are elucidated in detail on the basis of the analysis of the minutes of the meetings of the Vuojoki working party. The paper helps to understand the smooth site selection process in Finland.more » The context of the local decision-making is viewed from the policy, institutional and economic aspect. It is concluded in the paper that when trying to understand the progress of the Finnish site selection process more emphasis should be put on the role of TVO, the economic dependency of the Municipality of Eurajoki on TVO and the partnership between TVO and the leading local politicians. (authors)« less

Spent fuel and high-level radioactive waste transportation report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educatedmore » layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Spent fuel and high-level radioactive waste transportation report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educatedmore » layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Spent Fuel and High-Level Radioactive Waste Transportation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or nomore » background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.« less

Spent sealed radium sources conditioning in Latin America

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mourao, R.P.

1999-06-01

The management of spent sealed sources is considered by the International Atomic Energy Agency (IAEA) one of the greatest challenges faced by nuclear authorities today, especially in developing countries. One of the Agency`s initiatives to tackle this problem is the Spent Radium Sources Conditioning Project, a worldwide project relying on the regional co-operation between countries. A team from the Brazilian nuclear research institute Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) was chosen as the expert team to carry out the operations in Latin America; since December 1996 radium sources have been safely conditioned in Uruguay, Nicaragua, Guatemala, Ecuador and Paraguay.more » A Quality Assurance Program was established, encompassing the qualification of the capsule welding process, written operational procedures referring to all major steps of the operation, calibration of monitors and information retrievability. A 200L carbon steel drum-based packaging concept was used to condition the sources, its cavity being designed to receive the lead shield device containing stainless steel capsules with the radium sources. As a result of these operations, a total amount of 2,897 mg of needles, tubes, medical applicators, standard sources for calibration, lightning rods, secondary wastes and contaminated objects were stored in proper conditions and are now under control of the nuclear authorities of the visited countries.« less

ATR Spent Fuel Options Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connolly, Michael James; Bean, Thomas E.; Brower, Jeffrey O.

The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage,more » and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center’s (INTEC’s) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and

The study of ionization by electron impact of a substance simulating spent nuclear fuel components

NASA Astrophysics Data System (ADS)

Antonov, N. N.; Bochkarev, E. I.; Gavrikov, A. V.; Samokhin, A. A.; Smirnov, V. P.

2015-11-01

Plasma sources of model substances are necessary to solve problems associated with development of the spent nuclear fuel (SNF) plasma separation method. Lead was chosen to simulate kinetic and dynamic properties of the heavy SNF components. In this paper we present the results of a study of a lead vapor discharge with a lead concentration of 1012-1013 cm-3. Ionization was carried out by an electron beam (with energy of up to 500 eV per electron) inside a centimeter gap between planar electrodes. The discharge was numerically modeled using the hydrodynamic and single-particle approximation. Current-voltage characteristics and single ionization efficiency were obtained as functions of the vapors concentration and thermoelectric current. An ion current of hundreds of microamperes at the ionization efficiency near tenths of a percent was experimentally obtained. These results are in good agreement with our model.

Thermophysical property and pore structure evolution in stressed and non-stressed neutron irradiated IG-110 nuclear graphite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snead, Lance; Contescu, Christian I.; Byun, Thak Sang

2016-08-01

The nuclear graphite, IG-110, was irradiated with and without a compressive load of 5 MPa at ~400 *C up to 9.3E25 n/m2 (E > 0.1 MeV). Following irradiation physical properties were studied to compare the effect of graphite irradiation on microstructure developed under compression and in stress-free conditions. Properties included: dimensional change, thermal conductivity, dynamic modulus, and CTE. The effect of stress on open internal porosity was determined through nitrogen adsorption. The IG-110 graphite experienced irradiation-induced creep that is differentiated from irradiation-induced swelling. Irradiation under stress resulted in somewhat greater thermal conductivity and coefficient of thermal expansion. While a significantmore » increase in dynamic modulus occurs, no differentiation between materials irradiated with and without compressive stress was observed. Nitrogen adsorption analysis suggests a difference in pore evolution in the 0.3e40 nm range for graphite irradiated with and without stress, but this evolution is seen to be a small contributor to the overall dimensional change.« less

Thermophysical property and pore structure evolution in stressed and non-stressed neutron irradiated IG-110 nuclear graphite

DOE PAGES

Snead, Lance L.; Contescu, C. I.; Byun, T. S.; ...

2016-04-23

The nuclear graphite, IG-110, was irradiated with and without a compressive load of 5 MPa at ~400 C up to 9.3x10 25 n/m 2 (E>0.1 MeV.) Following irradiation physical properties were studied to compare the effect of graphite irradiation on microstructure developed under compression and in stress-free condition. Properties included: dimensional change, thermal conductivity, dynamic modulus, and CTE. The effect of stress on open internal porosity was determined through nitrogen adsorption. The IG-110 graphite experienced irradiation-induced creep that is differentiated from irradiation-induced swelling. Irradiation under stress resulted in somewhat greater thermal conductivity and coefficient of thermal expansion. While a significantmore » increase in dynamic modulus occurs, no differentiation between materials irradiated with and without compressive stress was observed. Nitrogen adsorption analysis suggests a difference in pore evolution in the 0.3-40 nm range for graphite irradiated with and without stress, but this evolution is seen to be a small contributor to the overall dimensional change.« less

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE PAGES

Jiang, Hao; Wang, Jy-An John; Wang, Hong

2016-09-26

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Hao; Wang, Jy-An John; Wang, Hong

Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets tomore » the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.« less

75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-15

... of establishing one or more technologies that the [Nuclear Regulatory] Commission may, by rule... technology approved by the Commission under Section 218(a) for use at the site of any civilian nuclear power... NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 [NRC-2010-0140] RIN 3150-AI86 List of Approved Spent...

76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-13

... Fuel Storage Casks: NUHOMS[supreg] HD System Revision 1 AGENCY: Nuclear Regulatory Commission. ACTION... amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN) NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to...

Role of temperature in the radiation stability of yttria stabilized zirconia under swift heavy ion irradiation: A study from the perspective of nuclear reactor applications

NASA Astrophysics Data System (ADS)

Kalita, Parswajit; Ghosh, Santanu; Sattonnay, Gaël; Singh, Udai B.; Grover, Vinita; Shukla, Rakesh; Amirthapandian, S.; Meena, Ramcharan; Tyagi, A. K.; Avasthi, Devesh K.

2017-07-01

The search for materials that can withstand the harsh radiation environments of the nuclear industry has become an urgent challenge in the face of ever-increasing demands for nuclear energy. To this end, polycrystalline yttria stabilized zirconia (YSZ) pellets were irradiated with 80 MeV Ag6+ ions to investigate their radiation tolerance against fission fragments. To better simulate a nuclear reactor environment, the irradiations were carried out at the typical nuclear reactor temperature (850 °C). For comparison, irradiations were also performed at room temperature. Grazing incidence X-ray diffraction and Raman spectroscopy measurements reveal degradation in crystallinity for the room temperature irradiated samples. No bulk structural amorphization was however observed, whereas defect clusters were formed as indicated by transmission electron microscopy and supported by thermal spike simulation results. A significant reduction of the irradiation induced defects/damage, i.e., improvement in the radiation tolerance, was seen under irradiation at 850 °C. This is attributed to the fact that the rapid thermal quenching of the localized hot molten zones (arising from spike in the lattice temperature upon irradiation) is confined to 850 °C (i.e., attributed to the resistance inflicted on the rapid thermal quenching of the localized hot molten zones by the high temperature of the environment) thereby resulting in the reduction of the defects/damage produced. Our results present strong evidence for the applicability of YSZ as an inert matrix fuel in nuclear reactors, where competitive effects of radiation damage and dynamic thermal healing mechanisms may lead to a strong reduction in the damage production and thus sustain its physical integrity.

Direct disposal of spent fuel: developing solutions tailored to Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kawamura, Hideki; McKinley, Ian G

2013-07-01

With the past Government policy of 100% reprocessing in Japan now open to discussion, options for direct disposal of spent fuel (SF) are now being considered in Japan. The need to move rapidly ahead in developing spent fuel management concepts is closely related to the ongoing debate on the future of nuclear power in Japan and the desire to understand the true costs of the entire life cycle of different options. Different scenarios for future nuclear power - and associated decisions on extent of reprocessing - will give rise to quite different inventories of SF with different disposal challenges. Althoughmore » much work has been carried out spent fuel disposal within other national programmes, the potential for mining the international knowledge base is limited by the boundary conditions for disposal in Japan. Indeed, with a volunteer approach to siting, no major salt deposits and few undisturbed sediments, high tectonic activity, relatively corrosive groundwater and no deserts, it is evident that a tailored solution is needed. Nevertheless, valuable lessons can be learned from projects carried out worldwide, if focus is placed on basic principles rather than implementation details. (authors)« less

Nuclear and Physical Properties of Dielectrics under Neutron Irradiation in Fast (BN-600) and Fusion (DEMO-S) Reactors

NASA Astrophysics Data System (ADS)

Blokhin, D. A.; Chernov, V. M.; Blokhin, A. I.

2017-12-01

Nuclear and physical properties (activation and transmutation of elements) of BN and Al2O3 dielectric materials subjected to neutron irradiation for up to 5 years in Russian fast (BN-600) and fusion (DEMO-S) reactors were calculated using the ACDAM-2.0 software complex for different post-irradiation cooling times (up to 10 years). Analytical relations were derived for the calculated quantities. The results may be used in the analysis of properties of irradiated dielectric materials and may help establish the rules for safe handling of these materials.

EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matloch, L.; Vaccaro, S.; Couland, M.

The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction ofmore » encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)« less

Recent developments - US spent fuel disposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

One of a US utility's major risk factors in continuing to operate a nuclear plant is managing discharged spent fuel. The US Department of Energy (DOE) signed contracts with utilities guaranteeing government acceptance of spent fuel by 1988. However, on December 17, 1992, DOE Secretary Watkins wrote to Sen. J. Bennett Johnston (D-LA), Chairman of the Senate Energy Committee, indicating a reassessment of DOE's programs, the results of which will be presented to Congress in January 1993. He indicated the Department may not be able to meet the 1988 date, because of difficulty in finding a site for the Monitoredmore » Retrievable Storage facility. Watkins indicated that DOE has investigated an interim solution and decided to expedite a program to certify a multi-purpose standardized cask system for spent fuel receipt, storage, transport, and disposal. To meet the expectations of US utilities, DOE is considering a plan to use federal sites for interim storage of the casks. Secretary Watkins recommended the waste program be taken off-budget and put in a revolving fund established to ensure that money already collected from utilities will be available to meet the schedule for completion of the repository.« less

AGC-2 Irradiation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohrbaugh, David Thomas; Windes, William; Swank, W. David

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a completemore » properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1996-02-01

On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operationmore » of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.« less

Incorporation mechanisms of actinide elements into the structures of U 6+ phases formed during the oxidation of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Burns, Peter C.; Ewing, Rodney C.; Miller, Mark L.

1997-05-01

Uranyl oxide hydrate and uranyl silicate phases will form due to the corrosion and alteration of spent nuclear fuel under oxidizing conditions in silica-bearing solution. The actinide elements in the spent fuel may be incorporated into the structures of these secondary U6+ phases during the long-term corrosion of the UO 2 in spent fuel. The incorporation of actinide elements into the crystal structures of the alteration products may decrease actinide mobility. The crystal chemistry of the various oxidation states of the actinide elements of environmental concern is examined to identify possible incorporation mechanisms. The substitutions Pu 6+U 6+ and (Pu 5+, Np 5+)U 6+ should readily occur in many U 6+ structures, although structural modification may be required to satisfy local bond-valence requirements. Crystal-chemical characteristics of the U 6+ phases indicate that An 4+ (An: actinide)U 6+ substitution is likely to occur in the sheets of uranyl polyhedra that occur in the structures of the minerals schoepite, [(UO 2) 8O 2(OH) 12](H 2O) 12, ianthinite, [U 24+ (UO 2) 4O 6(OH) 4(H 2O) 4](H 2O) 5, becquerelite, Ca[(UO 2) 3O 2(OH) 3] 2(H 2O) 8, compreignacite, K 2[(UO 2) 3O 2(OH) 3] 2(H 2O) 8, α-uranophane, Ca[(UO 2)(SiO 3OH)] 2(H 2O) 5, and boltwoodite, K(H 2O)[(UO 2)(SiO 4)], all of which are likely to form due to the oxidation and alteration of the UO 2 in spent fuel. The incorporation of An 3+ into the sheets of the structures of α-uranophane and boltwoodite, as well as interlayer sites of various uranyl phases, may occur.

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorgemans, J.; Mulhollem, L.; Glavin, J.

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first levelmore » of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

Spatially resolved nuclear spin relaxation, electron spin relaxation and light absorption in swift heavy ion irradiated LiF crystals.

PubMed

Stork, H; Dinse, K-P; Ditter, M; Fujara, F; Masierak, W; Neumann, R; Schuster, B; Schwartz, K; Trautmann, C

2010-05-12

Spatially resolved (19)F and (7)Li spin-lattice relaxation rates are measured for LiF single crystals after irradiation with two kinds of swift heavy ions ((12)C of 133 MeV and (208)Pb of 1.78 GeV incident energy). Like in earlier studies on (130)Xe and (238)U irradiated LiF crystals, we found a strong enhancement of the nuclear spin-lattice relaxation rate within the ion penetration depth and a slight--but still significant--enhancement beyond. By evaluating the nuclear relaxation rate enhancement within the ion range after irradiation with different projectiles, a universal relationship between the spin-lattice relaxation rate and the dose is deduced. The results of accompanying X-band electron paramagnetic resonance relaxation measurements and optical absorption spectroscopy are included in a physical interpretation of this relationship. Also the reason for the enhanced relaxation rate beyond the ion range is further discussed.

Congress Examines Nuclear Waste Disposal Recommendations

NASA Astrophysics Data System (ADS)

Showstack, Randy

2012-02-01

During an 8 February U.S. congressional hearing to examine how to move forward on dealing with spent nuclear fuel and to review other recommendations of the recently released final report of the White House-appointed Blue Ribbon Commission on America's Nuclear Future (BRC), Yucca Mountain was the 65,000-ton gorilla in the room. BRC's charge was to conduct a comprehensive review of policies to manage the back end of the nuclear fuel cycle and recommend a new strategy for dealing with the 65,000 tons of spent nuclear fuel currently stored at 75 sites around the country and the 2000 tons of new spent fuel being produced each year. However, BRC specifically did not evaluate Yucca Mountain. A 26 January letter from BRC to U.S. secretary of energy Steven Chu states, "You directed that the Commission was not to serve as a siting body. Accordingly, we have not evaluated Yucca Mountain or any other location as a potential site for the storage of spent nuclear fuel or disposal of high-level waste nor have we taken a position on the administration's request to withdraw the Yucca Mountain license application."

A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

DOE Office of Scientific and Technical Information (OSTI.GOV)

Parker, Frank L.

2012-07-01

Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storagemore » sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions

75 FR 24786 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-06

... of establishing one or more technologies that the [Nuclear Regulatory] Commission may, by rule... technology approved by the Commission under Section 218(a) for use at the site of any civilian nuclear power... NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AI75 [NRC-2009-0538] List of Approved Spent...

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2012-06-26

145 Abdul Mannan, “Preventing Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial ...Pakistan’s Civil Nuclear Program.” Some analysts argue that spent nuclear fuel is more vulnerable when being transported . 146 Martellini, 2008. 147...produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment technology, which it mastered by the mid-1980s

Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

PubMed

Rezaeian, M; Kamali, J

2017-01-01

Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B 4 C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

Stability of zinc stearate under alpha irradiation in the manufacturing process of SFR nuclear fuels

NASA Astrophysics Data System (ADS)

Gracia, J.; Vermeulen, J.; Baux, D.; Sauvage, T.; Venault, L.; Audubert, F.; Colin, X.

2018-03-01

The manufacture of new fuels for sodium-cooled fast reactors (SFRs) will involve powders derived from recycling existing fuels in order to keep on producing electricity while saving natural resources and reducing the amount of waste produced by spent MOX fuels. Using recycled plutonium in this way will significantly increase the amount of 238Pu, a high energy alpha emitter, in the powders. The process of shaping powders by pressing requires the use of a solid lubricant, zinc stearate, to produce pellets with no defects compliant with the standards. The purpose of this study is to determine the impact of alpha radiolysis on this additive and its lubrication properties. Experiments were conducted on samples in contact with PuO2, as well as under external helium ion beam irradiation, in order to define the kinetics of radiolytic gas generation. The yield results relating to the formation of these gases (G0) show that the alpha radiation of plutonium can be simulated using external helium ion beam irradiation. The isotopic composition of plutonium has little impact on the yield. However, an increased yield was globally observed with increasing the mean linear energy transfer (LET). A radiolytic degradation process is proposed.

Intact and Degraded Criticality Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

L.M. Montierth

2000-09-15

The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the U.S. Department of Energy's (DOE) Shippingport Light Water Breeder Reactor (SP LWBR) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP), which is to be placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (K{sub eff}) for intact- and degraded-mode internal configurations of the codisposal WP containing Shippingport LWBR seed-type assemblies. The results of this calculation will be used to evaluate criticality issuesmore » and support the analysis that is planed to be performed to demonstrate the viability of the codisposal concept for the MGR. This calculation is associated with the waste package design and was performed in accordance with the DOE SNF Analysis Plan for FY 2000 (See Ref. 22). The document has been prepared in accordance with the Administrative Procedure AP-3.12Q, Calculations (Ref. 23).« less

The ``Nuclear Renaissance'' and the Spread of Nuclear Weapons

NASA Astrophysics Data System (ADS)

Lyman, Edwin S.

2007-05-01

As interest grows around the world in nuclear power as an energy source that could help control greenhouse gas emissions, some have proclaimed the arrival of a ``nuclear renaissance.'' But can the increased risks of more nuclear power be managed? The political crisis surrounding Iran's pursuit of uranium enrichment has exposed weaknesses in the nuclear nonproliferation regime. Also, al Qaeda's declared interest in weapons of mass destruction raises the concern that terrorists could acquire nuclear weapons by stealing materials from poorly secured facilities. Growth of nuclear energy would require the construction of many additional uranium enrichment plants. And the generation of more spent nuclear fuel without a credible waste disposal strategy would increase political support for reprocessing, which separates large quantities of weapon-usable plutonium from spent fuel. There is little evidence that the various institutional arrangements and technical schemes proposed to mitigate the security risks of a major nuclear expansion would be effective. This talk will focus on the measures necessary to allow large-scale global growth of nuclear power without resulting in an unacceptably high risk of nuclear proliferation and nuclear terrorism, and will discuss the feasibility of such measures. To cite this abstract, use the following reference: http://meetings.aps.org/link/BAPS.2007.OSS07.E1.2

Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-12-01

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecastedmore » is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.« less

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Hu, Jianwei; De Baere, P.

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the frameworkmore » of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative

An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L

2009-01-01

There are a variety of reasons for quantifying plutonium (Pu) in spent fuel. Below, five motivations are listed: (1) To verify the Pu content of spent fuel without depending on unverified information from the facility, as requested by the IAEA ('independent verification'). New spent fuel measurement techniques have the potential to allow the IAEA to recover continuity of knowledge and to better detect diversion. (2) To assure regulators that all of the nuclear material of interest leaving a nuclear facility actually arrives at another nuclear facility ('shipper/receiver'). Given the large stockpile of nuclear fuel at reactor sites around the world,more » it is clear that in the coming decades, spent fuel will need to be moved to either reprocessing facilities or storage sites. Safeguarding this transportation is of significant interest. (3) To quantify the Pu in spent fuel that is not considered 'self-protecting.' Fuel is considered self-protecting by some regulatory bodies when the dose that the fuel emits is above a given level. If the fuel is not self-protecting, then the Pu content of the fuel needs to be determined and the Pu mass recorded in the facility's accounting system. This subject area is of particular interest to facilities that have research-reactor spent fuel or old light-water reactor (LWR) fuel. It is also of interest to regulators considering changing the level at which fuel is considered self-protecting. (4) To determine the input accountability value at an electrochemical processing facility. It is not expected that an electrochemical reprocessing facility will have an input accountability tank, as is typical in an aqueous reprocessing facility. As such, one possible means of determining the input accountability value is to measure the Pu content in the spent fuel that arrives at the facility. (5) To fully understand the composition of the fuel in order to efficiently and safely pack spent fuel into a long-term repository. The NDA of spent

78 FR 73566 - Standard Format and Content for a License Application for an Independent Spent Fuel Storage...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-06

...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment draft regulatory guide (DG), DG-3042, ``Standard Format and Content for a License Application for an Independent Spent Fuel Storage Installation or a Monitored Retrievable Storage Facility.'' This draft regulatory guide is proposed revision 2 of Regulatory Guide 3.50, which provides a format that the NRC considers acceptable for submitting the information for license applications to store spent nuclear fuel, high-level radioactive waste, and/or reactor-related Greater than Class C waste.

Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

DOEpatents

Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

2015-12-29

Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

Method For Processing Spent (Trn,Zr)N Fuel

DOEpatents

Miller, William E.; Richmann, Michael K.

2004-07-27

A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

Determination of the plutonium content in a spent fuel assembly by passive and active interrogation using a differential die-away instrument

NASA Astrophysics Data System (ADS)

Henzl, V.; Croft, S.; Richard, J.; Swinhoe, M. T.; Tobin, S. J.

2013-06-01

In this paper, we present a novel approach to estimating the total plutonium content in a spent fuel assembly (SFA) that is based on combining information from a passive measurement of the total neutron count rate (PN) of the assayed SFA and a measure of its multiplication. While PN can be measured essentially with any non-destructive assay (NDA) technique capable of neutron detection, the measure of multiplication is, in our approach, determined by means of active interrogation using an instrument based on the Differential Die-Away technique (DDA). The DDA is a NDA technique developed within the U.S. Department of Energy's Next Generation Safeguards Initiative (NGSI) project focused on the utilization of NDA techniques to determine the elemental plutonium content in commercial nuclear SFA's [1]. This approach was adopted since DDA also allows determination of other SFA characteristics, such as burnup, initial enrichment, and cooling time, and also allows for detection of certain types of diversion of nuclear material. The quantification of total plutonium is obtained using an analytical correlation function in terms of the observed PN and active multiplication. Although somewhat similar approaches relating Pu content with PN have been adopted in the past, we demonstrate by extensive simulation of the fuel irradiation and NDA process that our analytical method is independent of explicit knowledge of the initial enrichment, burnup, and an absolute value of the SFA's reactivity (i.e. multiplication factor). We show that when tested with MCNPX™ simulations comprising the 64 SFA NGSI Spent Fuel Library-1 we were able to determine elemental plutonium content, using just a few calibration parameters, with an average variation in the prediction of around 1-2% across the wide dynamic range of irradiation history parameters used, namely initial enrichment (IE=2-5%), burnup (BU=15-60 GWd/tU) and cooling time (CT=1-80 y). In this paper we describe the basic approach and the

Spent fuel treatment at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Benedict, R.W.; Levinskas, D.

1994-12-31

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Cycle Facility at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will employ a pyrochemical process that also has applications for treating most of the fuel types within the Department of Energy complex. The treatment equipment is in its last stage of readiness, and operations will begin in the Fall of 1994.

An influential factor for external radiation dose estimation for residents after the Fukushima Daiichi Nuclear Power Plant accident-time spent outdoors for residents in Iitate Village.

PubMed

Ishikawa, Tetsuo; Yasumura, Seiji; Ohtsuru, Akira; Sakai, Akira; Akahane, Keiichi; Yonai, Shunsuke; Sakata, Ritsu; Ozasa, Kotaro; Hayashi, Masayuki; Ohira, Tetsuya; Kamiya, Kenji; Abe, Masafumi

2016-06-01

Many studies have been conducted on radiation doses to residents after the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident. Time spent outdoors is an influential factor for external dose estimation. Since little information was available on actual time spent outdoors for residents, different values of average time spent outdoors per day have been used in dose estimation studies on the FDNPP accident. The most conservative value of 24 h was sometimes used, while 2.4 h was adopted for indoor workers in the UNSCEAR 2013 report. Fukushima Medical University has been estimating individual external doses received by residents as a part of the Fukushima Health Management Survey by collecting information on the records of moves and activities (the Basic Survey) after the accident from each resident. In the present study, these records were analyzed to estimate an average time spent outdoors per day. As an example, in Iitate Village, its arithmetic mean was 2.08 h (95% CI: 1.64-2.51) for a total of 170 persons selected from respondents to the Basic Survey. This is a much smaller value than commonly assumed. When 2.08 h is used for the external dose estimation, the dose is about 25% (23-26% when using the above 95% CI) less compared with the dose estimated for the commonly used value of 8 h.

Assessing the effectiveness of safeguards at a medium-sized spent-fuel reprocessing facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Higinbotham, W.; Fishbone, L.G.; Suda, S.

1983-01-01

In order to evaluate carefully and systematically the effectiveness of safeguards at nuclear-fuel-cycle facilities, the International Atomic Energy Agency has adopted a safeguards effectiveness assessment methodology. The methodology has been applied to a well-characterized, medium-sized, spent-fuel reprocessing plant to understand how explicit safeguards inspection procedures would serve to expose conceivable nuclear materials diversion schemes, should such diversion occur.

Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beatty, Randy L; Harrison, Thomas J

IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical ofmore » commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.« less

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlsen, Brett; Chiguer, Mustapha; Grahn, Per

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years untilmore » reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is

Radiolytic and thermal process relevant to dry storage of spent nuclear fuels. 1998 annual progress report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, S.C.; Cowin, J.P.; Orlando, T.M.

1998-06-01

'This project involves basic research in chemistry and physics aimed at providing information pertinent to the safe long-term dry storage of spent nuclear fuel (SNF), thousands of tons of which remain in water storage across the DOE complex. The Hanford Site K-Basins alone hold 2,300 tons of spent fuel, much of it severely corroded, and similar situations exist at Savannah River and Idaho National Engineering and Environmental Laboratory. The DOE plans to remove this fuel and seal it in overpack canisters for dry interim storage for up to 75 years while awaiting permanent disposition. Chemically-bound water will remain in thismore » fuel even following proposed drying steps, leading to possible long-term corrosion of the containers and/or fuel rods themselves, generation of H{sub 2} and O{sub 2} gas via radiolysis (which could lead to deflagration or detonation), and reactions of pyrophoric uranium hydrides. No thoroughly tested model is currently available to predict fuel behavior during pre-processing, processing, or storage. In a collaboration between Rutgers University, Pacific Northwest National Laboratory, and Brookhaven National Laboratory, the authors are studying the radiolytic reaction, drying processes, and corrosion behavior of actual SNF materials, and of pure and mixed-phase samples. The authors propose to determine what is omitted from current models: radiolysis of water adsorbed on or in hydrates or hydroxides, thermodynamics of interfacial phases, and kinetics of drying. A model will be developed and tested against actual fuel rod behavior to insure validity and applicability to the problems associated with developing dry storage strategies for DOE-owned SNF. This report summarizes work after eight months of a three-year project.'« less

Hydride reorientation and its impact on ambient temperature mechanical properties of high burn-up irradiated and unirradiated recrystallized Zircaloy-2 nuclear fuel cladding with an inner liner

NASA Astrophysics Data System (ADS)

Auzoux, Q.; Bouffioux, P.; Machiels, A.; Yagnik, S.; Bourdiliau, B.; Mallet, C.; Mozzani, N.; Colas, K.

2017-10-01

Precipitation of radial hydrides in zirconium-based alloy cladding concomitant with the cooling of spent nuclear fuel during dry storage can potentially compromise cladding integrity during its subsequent handling and transportation. This paper investigates hydride reorientation and its impact on ductility in unirradiated and irradiated recrystallized Zircaloy-2 cladding with an inner liner (cladding for boiling water reactors) subjected to hydride reorientation treatments. Cooling from 400 °C, hydride reorientation occurs in recrystallized Zircaloy-2 with liner at a lower effective stress in irradiated samples (below 40 MPa) than in unirradiated specimens (between 40 and 80 MPa). Despite significant hydride reorientation, unirradiated recrystallized Zircaloy-2 with liner cladding containing ∼200 wppm hydrogen shows a high diametral strain at fracture (>15%) during burst tests at ambient temperature. This ductile behavior is due to (1) the lower yield stress of the recrystallized cladding materials in comparison to hydride fracture strength (corrected by the compression stress arising from the precipitation) and (2) the hydride or hydrogen-depleted zone as a result of segregation of hydrogen into the liner layer. In irradiated Zircaloy-2 with liner cladding containing ∼340 wppm hydrogen, the conservation of some ductility during ring tensile tests at ambient temperature after reorientation treatment at 400 °C with cooling rates of ∼60 °C/h is also attributed to the existence of a hydride-depleted zone. Treatments at lower cooling rates (∼6 °C/h and 0.6 °C/h) promote greater levels of hydrogen segregation into the liner and allow for increased irradiation defect annealing, both of which result in a significant increase in ductility. Based on this investigation, given the very low cooling rates typical of dry storage systems, it can be concluded that the thermal transients associated with dry storage should not degrade, and more likely should actually

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-10-28

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008; and, Chris...related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-07-17

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...planned nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008...contracting between U.S. firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power

Development and calibration of the shielded measurement system for fissile contents measurements on irradiated nuclear fuel in dry storage.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mosby, W. R.; Jensen, B. A.

2002-05-31

In recent years there has been a trend towards storage of Irradiated Nuclear Fuel (INF) in dry conditions rather than in underwater environments. At the same time, the Department of Energy (DOE) has begun encouraging custodians of INF to perform measurements on INF for which no recent fissile contents measurement data exists. INF, in the form of spent fuel from Experimental Breeder Reactor 2 (EBR-II), has been stored in close-fitting, dry underground storage locations at the Radioactive Scrap and Waste Facility (RSWF) at Argonne National Laboratory-West (ANL-W) for many years. In Fiscal Year 2000, funding was obtained from the DOEmore » Office of Safeguards and Security Technology Development Program to develop and prepare for deployment a Shielded Measurement System (SMS) to perform fissile content measurements on INF stored in the RSWF. The SMS is equipped to lift an INF item out of its storage location, perform scanning neutron coincidence and high-resolution gamma-ray measurements, and restore the item to its storage location. The neutron and gamma-ray measurement results are compared to predictions based on isotope depletion and Monte Carlo neutral-particle transport models to provide confirmation of the accuracy of the models and hence of the fissile material contents of the item as calculated by the same models. This paper describes the SMS and discusses the results of the first calibration and validation measurements performed with the SMS.« less

National Policy Implications of Storing Nuclear Waste in the Pacific Region,

DTIC Science & Technology

1981-01-01

US Congress, Senate, Committee on Energy and Natural Resources, Pacific Spent Nuclear Fuel Storage , Hearing...selected. 17 One type of shipping cask which has been used to transport spent fuel assemblies to the Nevada Test Site is a leakproof steel cask that can...discussion the following conclusions on the nuclear waste storage issue appear valid. The Reagan decision to reprocess spent fuel has not changed US

Low-dose gamma-ray irradiation induces translocation of Nrf2 into nuclear in mouse macrophage RAW264.7 cells.

PubMed

Tsukimoto, Mitsutoshi; Tamaishi, Nana; Homma, Takujiro; Kojima, Shuji

2010-01-01

The transcription factor nuclear erythroid-derived 2-related factor 2 (Nrf2) regulates expression of genes encoding antioxidant proteins involved in cellular redox homeostasis, while gamma-ray irradiation is known to induce reactive oxygen species in vivo. Although activation of Nrf2 by various stresses has been studied, it has not yet been determined whether ionizing irradiation induces activation of Nrf2. Therefore, we investigated activation of Nrf2 in response to gamma-irradiation in mouse macrophage RAW264.7 cells. Irradiation of cells with gamma-rays induced an increase of Nrf2 expression. Even 0.1 Gy of gamma-irradiation induced a translocation of Nrf2 from cytoplasm to the nucleus, indicating the activation of Nrf2 by low-dose irradiation. Expression of heme oxygenase-1, which is regulated by Nrf2, was also increased at 24 h after irradiation with more than 0.1 Gy of gamma-rays. Furthermore, the activation of Nrf2 was suppressed by U0126, which is an inhibitor of the extracellular signal regulated protein kinase 1/2 (ERK1/2) pathway, suggesting involvement of ERK1/2-dependent pathway in the irradiation-induced activation of Nrf2. Our results indicate that low-dose gamma-irradiation induces activation of Nrf2 through ERK1/2-dependent pathways.

Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scott, Mark Robert

2005-05-01

An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, thesemore » attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.« less

10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

Code of Federal Regulations, 2010 CFR

2010-01-01

..., reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.128 Criteria for spent fuel, high-level radioactive waste, reactor...

10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

Code of Federal Regulations, 2011 CFR

2011-01-01

..., reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.128 Criteria for spent fuel, high-level radioactive waste, reactor...

ATF Neutron Irradiation Program Technical Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Geringer, J. W.; Katoh, Yutai

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization ofmore » irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.« less

Spent fuel measurements. passive neutron albedo reactivity (PNAR) and photon signatures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eigenbrodt, Julia; Menlove, Howard Olsen

2016-03-29

The International Atomic Energy Agency’s (IAEA) safeguards technical objective is the timely detection of a diversion of a significant quantity of nuclear material from peaceful activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. An important IAEA task towards meeting this objective is the ability to accurately and reliably measure spent nuclear fuel (SNF) to verify reactor operating parameters and verify that the fuel has not been removed from reactors or SNF storage facilities. This dissertation analyzes a method to improvemore » the state-of-the-art of nuclear material safeguards measurements using two combined measurement techniques: passive neutron albedo reactivity (PNAR) and passive spectral photon measurements.« less

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (alpha, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (alpha,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubes and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (alpha,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trahan, Alexis Chanel

New nondestructive assay techniques are sought to better characterize spent nuclear fuel. One of the NDA instruments selected for possible deployment is differential die-away self-interrogation (DDSI). The proposed DDSI approach for spent fuel assembly assay utilizes primarily the spontaneous fission and (α, n) neutrons in the assemblies as an internal interrogating radiation source. The neutrons released in spontaneous fission or (α,n) reactions are thermalized in the surrounding water and induce fission in fissile isotopes, thereby creating a measurable signal from isotopes of interest that would be otherwise difficult to measure. The DDSI instrument employs neutron coincidence counting with 3He tubesmore » and list-mode-based data acquisition to allow for production of Rossi-alpha distributions (RADs) in post-processing. The list-mode approach to data collection and subsequent construction of RADs has expanded the analytical possibilities, as will be demonstrated throughout this thesis. One of the primary advantages is that the measured signal in the form of a RAD can be analyzed in its entirety including determination of die-away times in different time domains. This capability led to the development of the early die-away method, a novel leakage multiplication determination method which is tested throughout the thesis on different sources in simulation space and fresh fuel experiments. The early die-away method is a robust, accurate, improved method of determining multiplication without the need for knowledge of the (α,n) source term. The DDSI technique and instrument are presented along with the many novel capabilities enabled by and discovered through RAD analysis. Among the new capabilities presented are the early die-away method, total plutonium content determination, and highly sensitive missing pin detection. Simulation of hundreds of different spent and fresh fuel assemblies were used to develop the analysis algorithms and the techniques were tested

75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-16

... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-20

... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

Food irradiation studies at the Institute of Nuclear Energy Research, Taiwan, Rep. of China

NASA Astrophysics Data System (ADS)

Fu, Ying-Kai; Tsai, Chao-Ming; Wu, Wen-Shi; Chang, Ming-Shia; Chang, Yung-Nien; Shu, Shih-Lin

The use of radiation to inhibit sprouting of potatoes, onions, gingers and garlic was studied at the Institute of Nuclear Energy Research. The sprout inhibition doses of potatoes, onions, gingers and garlic were found to be 10, 5, 2.5, 7.5 Krads, respectively. Changes in the content of moisture, ash, reducing sugars, total sugars, lipids, proteins, fiber etc. were monitored in various agricultural foodstuffs both with and without γ-irradiation at various doses. Fungicides did not prevent potatoes from decaying at 10 or 25°C with or without gamma radiation. Onions treated with any of fungicides has significantly more healthy tissues than controls at 10°C but not 25°C after 30 and 60 days storage, regardless of the presence or absence of gamma radiation. Insect pests have been causing great damage to stored rice in Taiwan. The four most harmful insects are: Sitophilus Zeamais Mostschulsky. Rhyzopertha dominica. Tribolitum custaneum Herbst and Sitotroga cerealella Oliver. Adults, eggs or larvae of these insect pests were irradiated by 60Co gamma rays. The results show that 40 Krads of gamma-irradiation could completely control these four species of pests in stored rice.

Irradiation pretreatment for coal desulfurization

NASA Technical Reports Server (NTRS)

Hsu, G. C.

1979-01-01

Process using highly-penetrating nuclear radiation (Beta and Gamma radiation) from nuclear power plant radioactive waste to irradiate coal prior to conventional desulfurization procedures increases total extraction of sulfur.

78 FR 20625 - Spent Nuclear Fuel Management at the Savannah River Site

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-05

... processing is a chemical separations process that involves dissolving spent fuel in nitric acid and... Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact... chemical properties, and radionuclide inventory. The fuel groups and the seven technologies that could be...

Quantity and management of spent fuel from prototype and research reactors in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

Comminuting irradiated ferritic steel

DOEpatents

Bauer, Roger E.; Straalsund, Jerry L.; Chin, Bryan A.

1985-01-01

Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

Nuclear Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Silver, E G

This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

FY 2016 Status Report: CIRFT Testing on Spent Nuclear Fuels and Hydride Reorientation Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong; Yan, Yong

This report provides a detailed description of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) testing conducted on spent nuclear fuel (SNF) rods in FY 2016, including hydride reorientation test results. Contact-based measurement, or three-LVDT-based curvature measurement, of SNF rods has proven to be quite reliable in CIRFT testing. However, how the linear variable differential transformer (LVDT) head contacts the SNF rod may have a significant effect on the curvature measurement, depending on the magnitude and direction of rod curvature. To correct such contact/curvature issues, sensor spacing, defined as the amount of separation between the three LVDT probes, is a criticalmore » measurement that can be used to calculate rod curvature once the deflections are obtained. Recently developed CIRFT data analyses procedures were integrated into FY 2016 CIRFT testing results for the curvature measurements. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNFs into a single zone. A detailed examination revealed that, at same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), Limerick Nuclear Power Station (LMK), mixed uranium-plutonium oxide (MOX). If looking at the strain, then LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The knee point of endurance limit in the curve of moment and curvature or equivalent quantities is more clearly defined for LMK and HBR fuels. The treatment affects the fatigue life of specimens. Both a drop of 12 in. and radial hydride treatment (RHT) have a negative impact on fatigue life. The effect of thermal annealing on MOX fuel rods was relatively small at higher amplitude but became significant at low amplitude of moment. Thermal annealing tended to extend the fatigue

Spent refractory reuse

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bennett, J.P.; Kwong, K.S.; Clark, J.A.

1996-12-31

The Albany Research Center is conducting work on spent refractory recycling/alternate use, including a review of refractory usage and current recycling/disposal practices. Research has focused on the hazardous nature of some spent refractory materials, with emphasis on lead pickup. Information on the issues associated with the reuse of spent refractories will be presented, including those associated with hazardous materials.

Concentration of 129I in aquatic biota collected from a lake adjacent to the spent nuclear fuel reprocessing plant in Rokkasho, Japan.

PubMed

Ueda, Shinji; Kakiuchi, Hideki; Hasegawa, Hidenao; Kawamura, Hidehisa; Hisamatsu, Shun'ichi

2015-11-01

The spent nuclear fuel reprocessing plant in Rokkasho, Japan, has been undergoing final testing since March 2006. During April 2006-October 2008, that spent fuel was cut and chemically processed, the plant discharged (129)I into the atmosphere and coastal waters. To study (129)I behaviour in brackish Lake Obuchi, which is adjacent to the plant, (129)I concentrations in aquatic biota were measured by accelerator mass spectrometry. Owing to (129)I discharge from the plant, the (129)I concentration in the biota started to rise from the background concentration in 2006 and was high during 2007-08. The (129)I concentration has been rapidly decreasing after the fuel cutting and chemically processing were finished. The (129)I concentration factors in the biota were higher than those reported by IAEA for marine organisms and similar to those reported for freshwater biota. The estimated annual committed effective dose due to ingestion of foods with the maximum (129)I concentration in the biota samples was 2.8 nSv y(-1). © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Spent nuclear fuel project high-level information management plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Main, G.C.

1996-09-13

This document presents the results of the Spent Nuclear Fuel Project (SNFP) Information Management Planning Project (IMPP), a short-term project that identified information management (IM) issues and opportunities within the SNFP and outlined a high-level plan to address them. This high-level plan for the SNMFP IM focuses on specific examples from within the SNFP. The plan`s recommendations can be characterized in several ways. Some recommendations address specific challenges that the SNFP faces. Others form the basis for making smooth transitions in several important IM areas. Still others identify areas where further study and planning are indicated. The team`s knowledge ofmore » developments in the IM industry and at the Hanford Site were crucial in deciding where to recommend that the SNFP act and where they should wait for Site plans to be made. Because of the fast pace of the SNFP and demands on SNFP staff, input and interaction were primarily between the IMPP team and members of the SNFP Information Management Steering Committee (IMSC). Key input to the IMPP came from a workshop where IMSC members and their delegates developed a set of draft IM principles. These principles, described in Section 2, became the foundation for the recommendations found in the transition plan outlined in Section 5. Availability of SNFP staff was limited, so project documents were used as a basis for much of the work. The team, realizing that the status of the project and the environment are continually changing, tried to keep abreast of major developments since those documents were generated. To the extent possible, the information contained in this document is current as of the end of fiscal year (FY) 1995. Programs and organizations on the Hanford Site as a whole are trying to maximize their return on IM investments. They are coordinating IM activities and trying to leverage existing capabilities. However, the SNFP cannot just rely on Sitewide activities to meet its IM

78 FR 58570 - Environmental Assessment; Entergy Nuclear Operations, Inc., Big Rock Point

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-24

... Assessment; Entergy Nuclear Operations, Inc., Big Rock Point AGENCY: Nuclear Regulatory Commission. ACTION... applicant or the licensee), for the Big Rock Point (BRP) Independent Spent Fuel Storage Installation (ISFSI... Rock Point (BRP) Independent Spent Fuel Storage Installation (ISFSI). II. Environmental Assessment (EA...

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... reactor following irradiation, the constituent elements of which have not been separated by reprocessing. ...

Delayed gamma-ray spectroscopy with lanthanum bromide detector for non-destructive assay of nuclear material

DOE Office of Scientific and Technical Information (OSTI.GOV)

Favalli, Andrea; Iliev, Metodi; Ianakiev, Kiril

High-energy delayed γ-ray spectroscopy is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Requirements for the γ-ray detection system, up to ~6 MeV, can be summarized as follows: high efficiency at high γ-ray energies, high energy resolution, good linearity between γ-ray energy and output signal amplitude, ability to operate at very high count rates, and ease of use in industrial environments such as nuclear facilities. High Purity Germanium Detectors (HPGe) are the state of the artmore » and provide excellent energy resolution but are limited in their count rate capability. Lanthanum Bromide (LaBr 3) scintillation detectors offer significantly higher count rate capabilities at lower energy resolution. Thus, LaBr 3 detectors may be an effective alternative for nuclear spent-fuel applications, where count-rate capability is a requirement. This paper documents the measured performance of a 2” (length) × 2” (diameter) of LaBr3 scintillation detector system, coupled to a negatively biased PMT and a tapered active high voltage divider, with count-rates up to ~3 Mcps. An experimental methodology was developed that uses the average current from the PMT’s anode and a dual source method to characterize the detector system at specific very high count rate values. Delayed γ-ray spectra were acquired with the LaBr 3 detector system at the Idaho Accelerator Center, Idaho State University, where samples of ~3g of 235U were irradiated with moderated neutrons from a photo-neutron source. Results of the spectroscopy characterization and analysis of the delayed γ-ray spectra acquired indicate the possible use of LaBr3 scintillation detectors when high count rate capability may outweigh the lower energy resolution.« less

Delayed gamma-ray spectroscopy with lanthanum bromide detector for non-destructive assay of nuclear material

DOE PAGES

Favalli, Andrea; Iliev, Metodi; Ianakiev, Kiril; ...

2017-10-09

High-energy delayed γ-ray spectroscopy is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Requirements for the γ-ray detection system, up to ~6 MeV, can be summarized as follows: high efficiency at high γ-ray energies, high energy resolution, good linearity between γ-ray energy and output signal amplitude, ability to operate at very high count rates, and ease of use in industrial environments such as nuclear facilities. High Purity Germanium Detectors (HPGe) are the state of the artmore » and provide excellent energy resolution but are limited in their count rate capability. Lanthanum Bromide (LaBr 3) scintillation detectors offer significantly higher count rate capabilities at lower energy resolution. Thus, LaBr 3 detectors may be an effective alternative for nuclear spent-fuel applications, where count-rate capability is a requirement. This paper documents the measured performance of a 2” (length) × 2” (diameter) of LaBr3 scintillation detector system, coupled to a negatively biased PMT and a tapered active high voltage divider, with count-rates up to ~3 Mcps. An experimental methodology was developed that uses the average current from the PMT’s anode and a dual source method to characterize the detector system at specific very high count rate values. Delayed γ-ray spectra were acquired with the LaBr 3 detector system at the Idaho Accelerator Center, Idaho State University, where samples of ~3g of 235U were irradiated with moderated neutrons from a photo-neutron source. Results of the spectroscopy characterization and analysis of the delayed γ-ray spectra acquired indicate the possible use of LaBr3 scintillation detectors when high count rate capability may outweigh the lower energy resolution.« less

Delayed gamma-ray spectroscopy with lanthanum bromide detector for non-destructive assay of nuclear material

NASA Astrophysics Data System (ADS)

Favalli, Andrea; Iliev, Metodi; Ianakiev, Kiril; Hunt, Alan W.; Ludewigt, Bernhard

2018-01-01

High-energy delayed γ-ray spectroscopy is a potential technique for directly assaying spent fuel assemblies and achieving the safeguards goal of quantifying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Requirements for the γ-ray detection system, up to ∼6 MeV, can be summarized as follows: high efficiency at high γ-ray energies, high energy resolution, good linearity between γ-ray energy and output signal amplitude, ability to operate at very high count rates, and ease of use in industrial environments such as nuclear facilities. High Purity Germanium Detectors (HPGe) are the state of the art and provide excellent energy resolution but are limited in their count rate capability. Lanthanum Bromide (LaBr3) scintillation detectors offer significantly higher count rate capabilities at lower energy resolution. Thus, LaBr3 detectors may be an effective alternative for nuclear spent-fuel applications, where count-rate capability is a requirement. This paper documents the measured performance of a 2" (length) × 2" (diameter) of LaBr3 scintillation detector system, coupled to a negatively biased PMT and a tapered active high voltage divider, with count-rates up to ∼3 Mcps. An experimental methodology was developed that uses the average current from the PMT's anode and a dual source method to characterize the detector system at specific very high count rate values. Delayed γ-ray spectra were acquired with the LaBr3 detector system at the Idaho Accelerator Center, Idaho State University, where samples of ∼3g of 235U were irradiated with moderated neutrons from a photo-neutron source. Results of the spectroscopy characterization and analysis of the delayed γ-ray spectra acquired indicate the possible use of LaBr3 scintillation detectors when high count rate capability may outweigh the lower energy resolution.

Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lippek, H.E.; Schuller, C.R.

1979-03-01

A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light mostmore » of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.« less

The study of the plasma jets of lead and silver simulating spent nuclear fuel components

NASA Astrophysics Data System (ADS)

Antonov, N. N.; Gavrikov, A. V.; Smirnov, V. P.; Liziakin, G. D.; Usmanov, R. A.; Vorona, N. A.; Timirkhanov, R. A.

2018-01-01

One of the tasks that must be solved to develop a spent nuclear fuel (SNF) plasma separation method is a creation of plasma source of substances simulating SNF components. Plasma of the diffuse arc discharge in a magnetic field with an incandescent cathode was considered in this paper, as such source. The discharge was initiated in a model substances vapor (lead and silver). Evaporation was carried out by crucible induction heating. Current- voltage characteristics of the discharge were obtained. Spectral analysis of the plasma jets radiation and double probe characteristics measurements in the area behind the anode were carried out. The minimum potential difference between the anode and cathode reached a value of about 7 V at current of about 1 A. When the potential difference in the discharge gap was close to 30 V (4.5 A) and 10 V (5.2 A) electron temperature in the plasma jet was 5-7 eV and 1-3 eV, respectively. Plasma density in jets took the value from 1011 cm-3 to 1012 cm-3. The obtained results indicate the possibility of using this type of discharge for the SNF plasma separation method approbation.

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.

2013-07-01

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. Themore » initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry

Separation of actinides from irradiated An-Zr based fuel by electrorefining on solid aluminium cathodes in molten LiCl-KCl

NASA Astrophysics Data System (ADS)

Souček, P.; Murakami, T.; Claux, B.; Meier, R.; Malmbeck, R.; Tsukada, T.; Glatz, J.-P.

2015-04-01

An electrorefining process for metallic spent nuclear fuel treatment is being investigated in ITU. Solid aluminium cathodes are used for homogeneous recovery of all actinides within the process carried out in molten LiCl-KCl eutectic salt at a temperature of 500 °C. As the selectivity, efficiency and performance of solid Al has been already shown using un-irradiated An-Zr alloy based test fuels, the present work was focused on laboratory-scale demonstration of the process using irradiated METAPHIX-1 fuel composed of U67-Pu19-Zr10-MA2-RE2 (wt.%, MA = Np, Am, Cm, RE = Nd, Ce, Gd, Y). Different electrorefining techniques, conditions and cathode geometries were used during the experiment yielding evaluation of separation factors, kinetic parameters of actinide-aluminium alloy formation, process efficiency and macro-structure characterisation of the deposits. The results confirmed an excellent separation and very high efficiency of the electrorefining process using solid Al cathodes.

Modelling the radiolytic corrosion of α-doped UO2 and spent nuclear fuel

NASA Astrophysics Data System (ADS)

Liu, Nazhen; Qin, Zack; Noël, James J.; Shoesmith, David W.

2017-10-01

A model previously developed to predict the corrosion rate of spent fuel (UO2) inside a failed waste container has been adapted to simulate the rates measured on a wide range of α-doped UO2 and spent fuel specimens. This simulation confirms the validity of the model and demonstrates that the steady-state corrosion rate is controlled by the radiolytic production of H2O2 (which has been shown to be the primary oxidant driving fuel corrosion), irrespective of the reactivity of the UO2 matrix. The model was then used to determine the consequences of corrosion inside a failed container resealed by steel corrosion products. The possible accumulation of O2, produced by H2O2 decomposition, was found to accelerate the corrosion rate in a closed system. However, the simultaneous accumulation of radiolytic H2, which is activated as a reductant on the noble metal (ε) particles in the spent fuel, rapidly overcame this acceleration leading to the eventual suppression of the corrosion rate to insignificant values. Calculations also showed that, while the radiation dose rate, the H2O2 decomposition ratio, and the surface coverage of ε particles all influenced the short term corrosion rate, the influence of the radiolytically produced H2 was the overwhelming influence in reducing the rate to negligible level (i.e., <10-20 mol m-2 s-1).

Image processing analysis of nuclear track parameters for CR-39 detector irradiated by thermal neutron

NASA Astrophysics Data System (ADS)

Al-Jobouri, Hussain A.; Rajab, Mustafa Y.

2016-03-01

CR-39 detector which covered with boric acid (H3Bo3) pellet was irradiated by thermal neutrons from (241Am - 9Be) source with activity 12Ci and neutron flux 105 n. cm-2. s-1. The irradiation times -TD for detector were 4h, 8h, 16h and 24h. Chemical etching solution for detector was sodium hydroxide NaOH, 6.25N with 45 min etching time and 60 C˚ temperature. Images of CR-39 detector after chemical etching were taken from digital camera which connected from optical microscope. MATLAB software version 7.0 was used to image processing. The outputs of image processing of MATLAB software were analyzed and found the following relationships: (a) The irradiation time -TD has behavior linear relationships with following nuclear track parameters: i) total track number - NT ii) maximum track number - MRD (relative to track diameter - DT) at response region range 2.5 µm to 4 µm iii) maximum track number - MD (without depending on track diameter - DT). (b) The irradiation time -TD has behavior logarithmic relationship with maximum track number - MA (without depending on track area - AT). The image processing technique principally track diameter - DT can be take into account to classification of α-particle emitters, In addition to the contribution of these technique in preparation of nano- filters and nano-membrane in nanotechnology fields.

75 FR 2163 - Constellation Energy; Notice of Docketing of Special Nuclear Material License SNM-2505 Amendment...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-14

... NUCLEAR REGULATORY COMMISSION [Docket No. 72-8; NRC-2010-0011] Constellation Energy; Notice of... Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory Commission. ACTION: Notice of license..., Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards, U.S...

75 FR 49813 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1, Confirmation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-16

... Storage Casks: MAGNASTOR System, Revision 1, Confirmation of Effective Date AGENCY: Nuclear Regulatory... spent fuel storage regulations at 10 CFR 72.214 to revise the MAGNASTOR System listing to include...

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, Charles W.

1998-01-01

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Systems and methods for retaining and removing irradiation targets in a nuclear reactor

DOEpatents

Runkle, Gary A.; Matsumoto, Jack T.; Dayal, Yogeshwar; Heinold, Mark R.

2015-12-08

A retainer is placed on a conduit to control movement of objects within the conduit in access-restricted areas. Retainers can prevent or allow movement in the conduit in a discriminatory fashion. A fork with variable-spacing between prongs can be a retainer and be extended or collapsed with respect to the conduit to change the size of the conduit. Different objects of different sizes may thus react to the fork differently, some passing and some being blocked. Retainers can be installed in inaccessible areas and allow selective movement in remote portions of conduit where users cannot directly interface, including below nuclear reactors. Position detectors can monitor the movement of objects through the conduit remotely as well, permitting engagement of a desired level of restriction and object movement. Retainers are useable in a variety of nuclear power plants and with irradiation target delivery, harvesting, driving, and other remote handling or robotic systems.

Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mickalonis, J. I.

2015-08-31

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less

Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mickalonis, J. I.

2015-08-01

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less

Modelling of radiation field around spent fuel container.

PubMed

Kryuchkov, E F; Opalovsky, V A; Tikhomirov, G V

2005-01-01

Operation of nuclear reactors leads to the production of spent nuclear fuel (SNF). There are two basic strategies of SNF management: ultimate disposal of SNF in geological formations and recycle or repeated utilisation of reprocessed SNF. In both options, there is an urgent necessity to study radiation properties of SNF. Information about SNF radiation properties is required at all stages of SNF management. In order to reach more effective utilisation of nuclear materials, new fuel cycles are under development based on uranium-plutonium, uranium-thorium and some other types of nuclear fuel. These promising types of nuclear fuel are characterised by quite different radiation properties at all the stages of nuclear fuel cycle (NFC) listed above. So, comparative analysis is required for radiation properties of different nuclear fuel types at different NFC stages. The results presented here were obtained from the numerical analysis of the radiation field around transport containers of different SNF types and in SNF storage. The calculations are carried out with the application of the computer code packages SCALE-4.3 and MCNP-4C. Comparison of the dose parameters obtained for different models of the transport container with experimental data allowed us to make certain conclusions about the errors of numerical results caused by the approximate geometrical description of the transport container.

Applications of polymer coatings for the fabrication of copper-based containers for the ultimate disposal of Canada's spent nuclear fuel

NASA Astrophysics Data System (ADS)

Mortley, Aba

Oxygen-free, phosphorous doped copper containers have been proposed for the storage of the used nuclear fuel bundles as a part of Canada's multi-barrier, adaptive phased management procedure for long term storage of spent nuclear fuel bundles. The spent nuclear fuel disposal system proposed for Canada has been engineered based on the multi-barrier approach intended to minimize the risk that the radioactive materials enter the biosphere. Copper is known to be susceptible to corrosion and it is thought that the simultaneous exposure to aggressive ionizing radiation field and residual heat produced by the spent nuclear fuel and the surrounding groundwater would all challenge the container's integrity. The goal of the present work is to reduce the impact of corrosion in the early stages of emplacement with the addition of a protective coating. Specifically, castor oil based polyurethanes were assessed as coatings and their ability to act as an additional physical barrier in the multi-barrier system mentioned previously. The novelty of this work stems from the use of a naturally derived non-petroleum based material in the form of castor oil as the polyol component. Two types of castor oil polyurethanes were investigated, one based on an aliphatic hexamethylene diisocyanate (HMDI), and the other based on an aromatic 2,4-toluene diisocyanate (TDI). Radiation and saturation tests were conducted using varying conditions. Mixed field ionizing radiation was provided by a SLOWPOKE-2 pool-type nuclear research reactor, up to accumulated doses of 6 MGy at dose rates of 37 kGy h-1 and 55.5 kGy h-1. Weight gain immersion studies, at temperatures of 25° C, 50° C, 70° C, were used to determine the mass uptake of several different solutions. The solutions utilized in the present work included hydrochloric acids of varying pHs, distilled water, and buffered solutions, which simulated chloride and sulphide rich calcium-sodium bicarbonate waters. After being exposed to radiation and

Technology, safety and costs of decommissioning reference independent spent fuel storage installations. [Contains glossary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ludwick, J D; Moore, E B

1984-01-01

Safety and cost information is developed for the conceptual decommissioning of five different types of reference independent spent fuel storage installations (ISFSIs), each of which is being given consideration for interim storage of spent nuclear fuel in the United States. These include one water basin-type ISFSI (wet) and four dry ISFSIs (drywell, silo, vault, and cask). The reference ISFSIs include all component parts necessary for the receipt, handling and storage of spent fuel in a safe and efficient manner. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, and potential radiation doses tomore » the public. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment followed by long-term surveillance).« less

HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

DUNCAN, D.R.

The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation inmore » other environments.« less

International Maritime Organizational Code For Safe Carriage Of Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes in Flasks On Board Ships -- IMO Resolution A.748(18)

DOT National Transportation Integrated Search

1994-05-26

This Circular calls the attention of Coast Guard field units, marine surveyors, shippers and carriers of nuclear materials to the International Maritime Organization (IMO) Code for the Safe Carriage of Irradiated Nuclear Fuel, Plutonium and High-Leve...

Depleted uranium as a backfill for nuclear fuel waste package

DOEpatents

Forsberg, C.W.

1998-11-03

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Enhancing the efficacy and persistency of Spodoptera litura (Fab.) nuclear polyhedrosis virus using UV irradiation protectants.

PubMed

Arivudainambi, S; Selvanarayanan, V; Vikash, A

2000-11-01

To enhance the field persistency of S. litura nuclear polyhedrosis virus (SLNPV), three chemicals viz. cupric ammonium nitrate, tinopal and cupric sulphate were tried as protectants (0.01 mg/ml) against natural sunlight (UV) irradiation. On exposure for 8 hr and subsequent bioassaying (diet surface treatment), it was found that cupric sulphate protected the polyhedrosis inclusion bodies (PIBs), recording 95.56% mortality which was statistically at par with unexposed PIBs recording 97.78% mortality.

Converting Maturing Nuclear Sites to Integrated Power Production Islands

DOE PAGES

Solbrig, Charles W.

2011-01-01

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage ofmore » already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.« less

The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forinash, Betsy; Schultheisz, Daniel; Peake, Tom

2013-07-01

Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generallymore » applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)« less

RadNuc: A graphical user interface to deliver dose rate patterns encountered in nuclear medicine with a 137Cs irradiator

PubMed Central

Pasternack, Jordan B.; Howell, Roger W.

2012-01-01

The temporal variations in absorbed dose rates to organs and tissues in the body are very large in diagnostic and therapeutic nuclear medicine. The response of biological endpoints of relevance to radiation safety and therapeutic efficacy are generally modulated by dose rate. Therefore, it is important to understand how the complex dose rate patterns encountered in nuclear medicine impact relevant biological responses. Accordingly, a graphical user interface (GUI) was created to control a cesium-137 irradiator to deliver such dose rate patterns. Methods Visual Basic 6.0 was used to create a user-friendly GUI to control the dose rate by varying the thickness of a mercury attenuator. The GUI facilitates the delivery of a number of dose rate patterns including constant, exponential increase or decrease, and multi-component exponential. Extensive visual feedback is provided by the GUI during both the planning and delivery stages. Results The GUI controlled irradiator can achieve a maximum dose rate of 40 cGy/hr and a minimum dose rate of 0.01 cGy/hr. Addition of machined lead blocks can be used to further reduce the minimum dose rate to 0.0001 cGy/hr. Measured dose rate patterns differed from programmed dose rate patterns in total dose by 3.2% to 8.4%. Conclusion The GUI controlled irradiator is able to accurately create dose rate patterns encountered in nuclear medicine and other related fields. This makes it an invaluable tool for studying the effects of chronic constant and variable low dose rates on biological tissues in the contexts of both radiation protection and clinical administration of internal radionuclides. PMID:23265668

RadNuc: a graphical user interface to deliver dose rate patterns encountered in nuclear medicine with a 137Cs irradiator.

PubMed

Pasternack, Jordan B; Howell, Roger W

2013-02-01

The temporal variations in absorbed dose rates to organs and tissues in the body are very large in diagnostic and therapeutic nuclear medicine. The response of biological endpoints of relevance to radiation safety and therapeutic efficacy is generally modulated by dose rate. Therefore, it is important to understand how the complex dose rate patterns encountered in nuclear medicine impact relevant biological responses. Accordingly, a graphical user interface (GUI) was created to control a cesium-137 irradiator to deliver such dose rate patterns. Visual Basic 6.0 was used to create a user-friendly GUI to control the dose rate by varying the thickness of a mercury attenuator. The GUI facilitates the delivery of a number of dose rate patterns including constant, exponential increase or decrease, and multi-component exponential. Extensive visual feedback is provided by the GUI during both the planning and delivery stages. The GUI controlled irradiator can achieve a maximum dose rate of 40 cGy/h and a minimum dose rate of 0.01 cGy/h. Addition of machined lead blocks can be used to further reduce the minimum dose rate to 0.0001 cGy/h. Measured dose rate patterns differed from programmed dose rate patterns in total dose by 3.2% to 8.4%. The GUI controlled irradiator is able to accurately create dose rate patterns encountered in nuclear medicine and other related fields. This makes it an invaluable tool for studying the effects of chronic constant and variable low dose rates on biological tissues in the contexts of both radiation protection and clinical administration of internal radionuclides. Copyright © 2013 Elsevier Inc. All rights reserved.

Comparison of selected foreign plans and practices for spent fuel and high-level waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.

1990-04-01

This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal ofmore » spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.« less

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 2. GSFLS visit findings and evaluations. Interim report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1978-01-31

This report is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. This report describes a global framework that evaluates spent fuel disposition requirements, influencing factors and strategies. A broad sampling of foreign governmental officials, electric utility spokesmen and nuclear power industry officials responsible for GSFLS policies, plans and programs were surveyed as to their views with respect to national and international GSFLS related considerations. The results of these GSFLS visit findings are presented herein. These findings were then evaluated in terms of technical, institutional and legal/regulatory implications. The GSFLS evaluations, in conjunctionmore » with perceived US spent fuel objectives, formed the basis for selecting a set of GSFLS strategies which are reported herein.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE PAGES

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

2017-09-01

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

78 FR 78165 - List of Approved Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-26

... Spent Fuel Storage Casks: HI-STORM 100 Cask System; Amendment No. 9 AGENCY: Nuclear Regulatory... storage regulations by revising the Holtec International HI-STORM 100 Cask System listing within the...

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

Post-dauer life span of Caenorhabditis elegans dauer larvae can be modified by X-irradiation.

PubMed

Onodera, Akira; Yanase, Sumino; Ishii, Takamasa; Yasuda, Kayo; Miyazawa, Masaki; Hartman, Philip S; Ishii, Naoaki

2010-01-01

The time spent as a dauer larva does not affect adult life span in Caenorhabditis elegans, as if aging is suspended in this quiescent developmental stage. We now report that modest doses X-irradiation of dauer larvae increased their post-dauer longevity. Post-irradiation incubation of young dauer larvae did not modify this beneficial effect of radiation. Conversely, holding dauer larvae prior to irradiation rendered them refractory to this X-radiation-induced response. We present a model to explain these results. These experiments demonstrate that dauer larvae provide an excellent opportunity to study mechanisms by which X irradiation can extend life span.

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE PAGES

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; ...

2015-12-01

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-12-23

reactors deployed” in the UAE. Some Members of Congress had welcomed the UAE government’s stated commitments not to pursue proliferation-sensitive...for the planned nuclear reactor or on handling spent reactor fuel. (...continued) May...firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two

Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

DOEpatents

Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

1985-01-01

A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

LWRS ATR Irradiation Testing Readiness Status

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less

10 CFR 72.78 - Nuclear material transaction reports.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Nuclear material transaction reports. 72.78 Section 72.78 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Records...

10 CFR 72.78 - Nuclear material transaction reports.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Nuclear material transaction reports. 72.78 Section 72.78 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Records...

10 CFR 72.78 - Nuclear material transaction reports.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Nuclear material transaction reports. 72.78 Section 72.78 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Records...

10 CFR 72.78 - Nuclear material transaction reports.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Nuclear material transaction reports. 72.78 Section 72.78 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Records...

Irradiation subassembly

DOEpatents

Seim, O.S.; Filewicz, E.C.; Hutter, E.

1973-10-23

An irradiation subassembly for use in a nuclear reactor is described which includes a bundle of slender elongated irradiation -capsules or fuel elements enclosed by a coolant tube and having yieldable retaining liner between the irradiation capsules and the coolant tube. For a hexagonal bundle surrounded by a hexagonal tube the yieldable retaining liner may consist either of six segments corresponding to the six sides of the tube or three angular segments each corresponding in two adjacent sides of the tube. The sides of adjacent segments abut and are so cut that metal-tometal contact is retained when the volume enclosed by the retaining liner is varied and Springs are provided for urging the segments toward the center of the tube to hold the capsules in a closely packed configuration. (Official Gazette)

78 FR 77606 - Security Requirements for Facilities Storing Spent Nuclear Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-24

... regulatory basis; availability of responses to public comments. SUMMARY: The U.S. Nuclear Regulatory... to Carol Gallagher; telephone: 301-287- 3422; email: [email protected] . For technical... Nuclear Security and Incident Response, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001...

``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jerde, E.A.; Glasgow, D.C.

1998-09-01

At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate countingmore » leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.« less

Emulation of reactor irradiation damage using ion beams

DOE PAGES

Was, G. S.; Jiao, Z.; Getto, E.; ...

2014-06-14

The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke

2015-06-01

High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried outmore » at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235U, 239Pu, and 241Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241Pu, a significant fissile constituent in spent fuel, was measured and compared to 239Pu. The 241Pu/ 239Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities

A preliminary evaluation of the ability of from-reactor casks to geometrically accommodate commercial LWR spent nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andress, D.; Joy, D.S.; McLeod, N.B.

The Department of Energy has sponsored a number of cask design efforts to define several transportation casks to accommodate the various assemblies expected to be accepted by the Federal Waste Management System. At this time, three preliminary cask designs have been selected for the final design--the GA-4 and GA-9 truck casks and the BR-100 rail cask. In total, this assessment indicates that the current Initiative I cask designs can be expected to dimensionally accommodate 100% of the PWR fuel assemblies (other than the extra-long South Texas Fuel) with control elements removed, and >90% of the assemblies having the control elementsmore » as an integral part of the fuel assembly. For BWR assemblies, >99% of the assemblies can be accommodated with fuel channels removed. This paper summarizes preliminary results of one part of that evaluation related to the ability of the From-Reactor Initiative I casks to accommodate the physical and radiological characteristics of the Spent Nuclear Fuel projected to be accepted into the Federal Waste Management System. 3 refs., 5 tabs.« less

77 FR 24585 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-25

... Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear Regulatory Commission. ACTION: Direct final rule... revising the Holtec International HI-STORM 100 System listing within the ``List of Approved Spent Fuel...) 72.214, by revising the Holtec International HI-STORM 100 System listing within the ``List of...

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

Nuclear transmutation in steels

NASA Astrophysics Data System (ADS)

Belozerova, A. R.; Shimanskii, G. A.; Belozerov, S. V.

2009-05-01

The investigations of the effects of nuclear transmutation in steels that are widely used in nuclear power and research reactors and in steels that are planned for the application in thermonuclear fusion plants, which are employed under the conditions of a prolonged action of neutron irradiation with different spectra, made it possible to study the effects of changes in the isotopic and chemical composition on the tendency of changes in the structural stability of these steels. For the computations of nuclear transmutation in steels, we used a program complex we have previously developed on the basis of algorithms for constructing branched block-type diagrams of nuclide transformations and for locally and globally optimizing these diagrams with the purpose of minimizing systematic errors in the calculation of nuclear transmutation. The dependences obtained were applied onto a Schaeffler diagram for steels used for structural elements of reactors. For the irradiation in fission reactors, we observed only a weak influence of the effects of nuclear transmutation in steels on their structural stability. On the contrary, in the case of irradiation with fusion neutrons, a strong influence of the effects of nuclear transmutation in steels on their structural stability has been noted.

Accelerator-Driven Subcritical System for Disposing of the U.S. Spent Nuclear Fuel Inventory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, Yousry; Cao, Yan; Kraus, Adam R.

The current United States inventory of the spent nuclear fuel (SNF) is ~80,000 metric tons of heavy metal (MTHM), including ~131 tons of minor actinides (MAs) and ~669 tons of plutonium. This study describes a conceptual design of an accelerator-driven subcritical (ADS) system for disposing of this SNF inventory by utilizing the 131 tons of MAs inventory and a fraction of the plutonium inventory for energy production, and transmuting some long-lived fission products. An ADS system with a homogeneous subcritical fission blanket was first examined. A spallation neutron source is used to drive the blanket and it is produced frommore » the interaction of a 1-GeV proton beam with a lead-bismuth eutectic (LBE) target. The blanket has a liquid mobile fuel using LBE as the fuel carrier. The fuel materials are dissolved, mixed, or suspended in the liquid fuel carrier. Monte Carlo analyses were performed to determine the overall parameters of the concept. Steady-state Monte Carlo simulations were performed for three similar fission blankets. Except for, the loaded amount of actinide materials in the LBE is either 5, 7, or 10% of the total volume of the blanket, respectively. The neutron multiplication factors of the three blankets are ~0.98 and the initial MAs blanket inventories are ~10 tons. In addition, Monte Carlo burnup simulations using the MCB5 code were performed to analyze the performance of the three conceptual ADS systems. During operation, fresh fuel was fed into the fission blanket to adjust its reactivity and to control the system power. The burnup analysis shows that the three ADS concepts consume about 1.2 tons of actinides per full power year and produce 3 GW thermal power, with a proton beam power of 25 MW. For the blankets with 5, 7, or 10% actinide fuel particles loaded in the LBE, assuming that the ADS systems can be operated for 35 full-power years, the total MA materials consumed in the three ADS systems are about 30.6, 35.3, and 37.2 tons

Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

NASA Astrophysics Data System (ADS)

Serrano-Purroy, D.; Casas, I.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; Clarens, F.; Giménez, J.; de Pablo, J.; Martínez-Esparza, A.

2013-03-01

The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used.For uranium and actinides, the results showed that U, Np, Am and Cm gave very similar normalized dissolution rates, while Pu showed slower dissolution rates for both samples. In addition, dissolution rates were consistently two to four times lower for OUT sample compared to CORE sample.Considering the fission products release the main results are that Y, Tc, La and Nd dissolved very similar to uranium; while Cs, Sr, Mo and Rb have up to 10 times higher dissolution rates. Rh, Ru and Zr seemed to have lower dissolution rates than uranium. The lowest dissolution rates were found for OUT sample.Three different contributions were detected on uranium release, modelled and attributed to oxidation layer, fines and matrix release.

Irradiation creep of candidate materials for advanced nuclear plants

NASA Astrophysics Data System (ADS)

Chen, J.; Jung, P.; Hoffelner, W.

2013-10-01

In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2-8 × 10-6 dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

Image processing analysis of nuclear track parameters for CR-39 detector irradiated by thermal neutron

DOE Office of Scientific and Technical Information (OSTI.GOV)

Al-Jobouri, Hussain A., E-mail: hahmed54@gmail.com; Rajab, Mustafa Y., E-mail: mostafaheete@gmail.com

CR-39 detector which covered with boric acid (H{sub 3}Bo{sub 3}) pellet was irradiated by thermal neutrons from ({sup 241}Am - {sup 9}Be) source with activity 12Ci and neutron flux 10{sup 5} n. cm{sup −2}. s{sup −1}. The irradiation times -T{sub D} for detector were 4h, 8h, 16h and 24h. Chemical etching solution for detector was sodium hydroxide NaOH, 6.25N with 45 min etching time and 60 C° temperature. Images of CR-39 detector after chemical etching were taken from digital camera which connected from optical microscope. MATLAB software version 7.0 was used to image processing. The outputs of image processing of MATLABmore » software were analyzed and found the following relationships: (a) The irradiation time -T{sub D} has behavior linear relationships with following nuclear track parameters: i) total track number - N{sub T} ii) maximum track number - MRD (relative to track diameter - D{sub T}) at response region range 2.5 µm to 4 µm iii) maximum track number - M{sub D} (without depending on track diameter - D{sub T}). (b) The irradiation time -T{sub D} has behavior logarithmic relationship with maximum track number - M{sub A} (without depending on track area - A{sub T}). The image processing technique principally track diameter - D{sub T} can be take into account to classification of α-particle emitters, In addition to the contribution of these technique in preparation of nano- filters and nano-membrane in nanotechnology fields.« less

Structural responses of metallic glasses under neutron irradiation.

PubMed

Yang, L; Li, H Y; Wang, P W; Wu, S Y; Guo, G Q; Liao, B; Guo, Q L; Fan, X Q; Huang, P; Lou, H B; Guo, F M; Zeng, Q S; Sun, T; Ren, Y; Chen, L Y

2017-12-01

Seeking nuclear materials that possess a high resistance to particle irradiation damage is a long-standing issue. Permanent defects, induced by irradiation, are primary structural changes, the accumulation of which will lead to structural damage and performance degradation in crystalline materials served in nuclear plants. In this work, structural responses of neutron irradiation in metallic glasses (MGs) have been investigated by making a series of experimental measurements, coupled with simulations in ZrCu amorphous alloys. It is found that, compared with crystalline alloys, MGs have some specific structural responses to neutron irradiation. Although neutron irradiation can induce transient vacancy-like defects in MGs, they are fully annihilated after structural relaxation by rearrangement of free volumes. In addition, the rearrangement of free volumes depends strongly on constituent elements. In particular, the change in free volumes occurs around the Zr atoms, rather than the Cu centers. This implies that there is a feasible strategy for identifying glassy materials with high structural stability against neutron irradiation by tailoring the microstructures, the systems, or the compositions in alloys. This work will shed light on the development of materials with high irradiation resistance.

75 FR 81031 - Consideration of Environmental Impacts of Temporary Storage of Spent Fuel After Cessation of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-23

... Part VI Nuclear Regulatory Commission 10 CFR Part 51 Consideration of Environmental Impacts of... Consideration of Environmental Impacts of Temporary Storage of Spent Fuel After Cessation of Reactor Operation... Commission (NRC or Commission) is revising its generic determination on the environmental impacts of storage...

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

NASA Astrophysics Data System (ADS)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2018-04-01

Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These

Bioremediation of trace cobalt from simulated spent decontamination solutions of nuclear power reactors using E. coli expressing NiCoT genes.

PubMed

Raghu, G; Balaji, V; Venkateswaran, G; Rodrigue, A; Maruthi Mohan, P

2008-12-01

Removal of radioactive cobalt at trace levels (approximately nM) in the presence of large excess (10(6)-fold) of corrosion product ions of complexed Fe, Cr, and Ni in spent chemical decontamination formulations (simulated effluent) of nuclear reactors is currently done by using synthetic organic ion exchangers. A large volume of solid waste is generated due to the nonspecific nature of ion sorption. Our earlier work using various fungi and bacteria, with the aim of nuclear waste volume reduction, realized up to 30% of Co removal with specific capacities calculated up to 1 microg/g in 6-24 h. In the present study using engineered Escherichia coli expressing NiCoT genes from Rhodopseudomonas palustris CGA009 (RP) and Novosphingobium aromaticivorans F-199 (NA), we report a significant increase in the specific capacity for Co removal (12 microg/g) in 1-h exposure to simulated effluent. About 85% of Co removal was achieved in a two-cycle treatment with the cloned bacteria. Expression of NiCoT genes in the E. coli knockout mutant of NiCoT efflux gene (rcnA) was more efficient as compared to expression in wild-type E. coli MC4100, JM109 and BL21 (DE3) hosts. The viability of the E. coli strains in the formulation as well as at different doses of gamma rays exposure and the effect of gamma dose on their cobalt removal capacity are determined. The potential application scheme of the above process of bioremediation of cobalt from nuclear power reactor chemical decontamination effluents is discussed.

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

10 CFR 72.124 - Criteria for nuclear criticality safety.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

A coupled effect of nuclear and electronic energy loss on ion irradiation damage in lithium niobate

DOE PAGES

Liu, Peng; Zhang, Yanwen; Xue, Haizhou; ...

2016-01-09

Understanding irradiation effects induced by elastic energy loss to atomic nuclei and inelastic energy loss to electrons in a crystal, as well as the coupled effect between them, is a scientific challenge. Damage evolution in LiNbO 3 irradiated by 0.9 and 21 MeV Si ions at 300 K has been studied utilizing Rutherford backscattering spectrometry in channeling mode. During the low-energy ion irradiation process, damage accumulation produced due to elastic collisions is described utilizing a disorder accumulation model. Moreover, low electronic energy loss is shown to induce observable damage that increases with ion fluence. For the same electronic energy loss,more » the velocity of the incident ion could affect the energy and spatial distribution of excited electrons, and therefore effectively modify the diameter of the ion track. Furthermore, nonlinear additive phenomenon of irradiation damage induced by high electronic energy loss in pre-damaged LiNbO 3 has been observed. The result indicates that pre-existing damage induced from nuclear energy loss interacts synergistically with inelastic electronic energy loss to promote the formation of amorphous tracks and lead to rapid phase transformation, much more efficient than what is observed in pristine crystal solely induced by electronic energy loss. As a result, this synergistic effect is attributed to the fundamental mechanism that the defects produced by the elastic collisions result in a decrease in thermal conductivity, increase in the electron-phonon coupling, and further lead to higher intensity in thermal spike from intense electronic energy deposition along high-energy ion trajectory.« less

AGC-2 Graphite Pre-irradiation Data Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterizedmore » prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.« less

Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

DOE PAGES

Shimada, Masashi; Cao, G.; Otsuka, T.; ...

2014-12-01

Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor, Oak Ridge National Laboratory at reactor coolant temperatures of 50-70°C to low displacement damage of 0.025 and 0.3 dpa under the framework of the US-Japan TITAN program (2007-2013). After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 °C twice at the ion fluence of 5×10²⁵ m⁻² to reach a total ion fluence of 1×10²⁶ m⁻² in order to investigate the near surface deuterium retention and saturation via nuclear reaction analysis. Finalmore » thermal desorption spectroscopy was performed to elucidate irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near surface (<5 µm depth) deuterium concentration increased from 0.5 at % D/W in 0.025 dpa samples to 0.8 at. % D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near surface retention via nuclear reaction analysis indicated the deuterium was migrated and trapped in bulk (at least 50 µm depth for 0.025 dpa and 35 µm depth for 0.025 dpa) at 500 °C case even in the relatively low ion fluence of 10²⁶ m⁻².« less

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 2A. GSFLS visit findings (appendix). Interim report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1978-01-31

This appendix is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. This appendix provides the legal/regulatory reference material, supportive of Volume 2 - GSFLS Visit Finding and Evaluations; and certain background material on British Nuclear Fuel Limited (BNFL).

Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reynard-Carette, C.; Lyoussi, A.

Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices thatmore » contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, J.P.; Miller, W.E.

1987-11-05

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

PWR and BWR spent fuel assembly gamma spectra measurements

NASA Astrophysics Data System (ADS)

Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

2016-10-01

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

PWR and BWR spent fuel assembly gamma spectra measurements

DOE PAGES

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

2016-07-17

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

PWR and BWR spent fuel assembly gamma spectra measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

Effects of materials and design on the criticality and shielding assessment of canister concepts for the disposal of spent nuclear fuel.

PubMed

Gutiérrez, Miguel Morales; Caruso, Stefano; Diomidis, Nikitas

2018-05-19

According to the Swiss disposal concept, the safety of a deep geological repository for spent nuclear fuel (SNF) is based on a multi-barrier system. The disposal canister is an important component of the engineered barrier system, aiming to provide containment of the SNF for thousands of years. This study evaluates the criticality safety and shielding of candidate disposal canister concepts, focusing on the fulfilment of the sub-criticality criterion and on limiting radiolysis processes at the outer surface of the canister which can enhance corrosion mechanisms. The effective neutron multiplication factor (k-eff) and the surface dose rates are calculated for three different canister designs and material combinations for boiling water reactor (BWR) canisters, containing 12 spent fuel assemblies (SFA), and pressurized water reactor (PWR) canisters, with 4 SFAs. For each configuration, individual criticality and shielding calculations were carried out. The results show that k-eff falls below the defined upper safety limit (USL) of 0.95 for all BWR configurations, while staying above USL for the PWR ones. Therefore, the application of a burnup credit methodology for the PWR case is required, being currently under development. Relevant is also the influence of canister material and internal geometry on criticality, enabling the identification of safer fuel arrangements. For a final burnup of 55MWd/kgHM and 30y cooling time, the combined photon-neutron surface dose rate is well below the threshold of 1 Gy/h defined to limit radiation-induced corrosion of the canister in all cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the "nuclear safe orbit" (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.

Radiation chemistry for modern nuclear energy development

NASA Astrophysics Data System (ADS)

Chmielewski, Andrzej G.; Szołucha, Monika M.

2016-07-01

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

Disposal of spent fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blomeke, J O; Ferguson, D E; Croff, A G

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed betweenmore » the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom.« less