Search NASA⌕ Search

SEARCH · Search NASA

Results for “Storage Cask”

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

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

At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A New Group Crystallization Approach in Used Nuclear Fuel Recycling: A Look at Fission Product Behaviors

Nuclear energy has the potential to play a large role in meeting future needs of the U.S. sustainable energy strategy, as a clean, carbon-free power source. However, to take advantage of the inherent benefits, many hurdles must be overcome in order for nuclear power to achieve these expectations. One of the largest challenges faced, is the management of used nuclear fuel (UNF). Currently, no strategy has been identified to deal with the growing UNF problem, and so the UNF remains stored on site at nuclear power plants either in cooling pools or in dry cask storage. The amount of stored UNF continues to grow, and with it, so does the concern surrounding its long-term radiotoxicity. This material could be reprocessed to recover the actinides, maximizing the energy potential of the UNF and minimizing the waste generation. Not only do U and Pu, the major energy producing actinides, need to be recovered, but recovering Np and Am would make waste generated from reprocessing significantly safer for long-term storage in a geological repository. There are a number of methods currently in use to reprocess UNF; throughout Europe, the Purex process is widely implemented, while the Japanese have developed the Next process. The Next process removes a bulk of the U by crystallization prior to a Purex-like separation scheme. In the Next process, it is undesirable to have any other species removed from solution in the crystallization step, but rather, the goal is to obtain a pure uranyl nitrate hexahydrate (UNH) product. Alternatively, a group separation of U-Am by a crystallization process would provide an elegant solution. Such a process has become attainable with the renewed interest in accessing Am(VI) in nitric acid solutions. In order to synthesize and stabilize Am(VI) which has only been demonstrated on a limited basis, a highly oxidizing system is required. By oxidizing Np, Pu, and Am to their hexavalent state, they can be co-crystallized with UNH, and thereby removed from solution. Recent results have demonstrated that all the hexavalent actinide (An) species are removed from solution in near proportion, by a simple reduction of temperature in 6-8 M HNO{sub 3}.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Neutron Absorber Plate Characterization Plan for Criticality Experiments Design

After being used in nuclear installations, depleted fuel can still be highly reactive and must be handled securely to prevent any radiological or criticality concerns. In particular, spent fuel from use in nuclear power reactors must be stored and transported in specifically designed containers using neutron absorber materials to prevent criticality. Various neutron absorber material types exist and are manufactured by various entities, as thoroughly described in the Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications written by EPRI. Presently, one of the most modern and most widely used types of neutron absorber material contains particles of boron carbide, or B 4 C, embedded in aluminum matrix: Boralcan, manufactured by Rio Tinto. It is very important for the community to know as much as possible about such neutron absorber materials. Therefore, in the recent years, a US Department of Energy National Nuclear Security Administration–Nuclear Criticality Safety Program funded project initiated design of an experiment that places Boralcan neutron-absorbing plates in an established critical assembly using low-enriched uranium fuel at the Sandia Pulsed Reactor Facility/Critical Experiments (SPRF/CX) apparatus at Sandia National Laboratories. The goal of the experiment is to produce high-quality benchmark data to submit to the International Criticality Safety Benchmark Evaluation Project (ICSBEP), for use in validating calculational tools and nuclear data by criticality safety analysts. The project, named IER-554, is currently in its final design stage, following a successful preliminary design. In the work documented in the design study, ten critical configurations using Boralcan neutron absorber plates were designed, and the experiment was proven to be feasible, with a predicted low k eff uncertainty around 100 pcm. An overview of the modeled cutout of the critical assembly with a Boralcan plate is shown in Figure 1, representing one of the configurations planned for the critical experiments. Before the plates are inserted in the critical assembly, it is necessary to know more about their composition and uniformity. This summary focuses on the plate characterization plans. Each plate will undergo (1) neutron transmission measurements at different locations to determine the 10 B areal density and (2) an in-depth x-ray computed tomography (XCT) examination to obtain the exact Sizes and distribution of the B4C powder particles inside the plates. In parallel, plate modeling studies are performed with a goal to determine the validity of the currently used approximation of modeling the neutron absorber plates as a homogeneous mixture of Aluminum 1100 alloy and B4C— instead of explicitly modeling the B4C particles. By using the experimental 10 B areal density measurements, and the exact size and location of the B4C particles obtained by XCT, a plate model can theoretically be built that reproduces the plate with extremely high fidelity. The results of this modeling study could increase the confidence of the criticality safety community in its modeling methods when using this type of neutron absorber material, and the industry could use these validations to change the boron loading credit limits from the U.S. Nuclear Regulatory Commission standard review plan for dry cask storage of spent nuclear fuel. The modeling calculations are performed with SCALE 6.3.0 using the KENO V.a sequence for criticality calculations with the ENDF/B-VIII.0 continuous-energy cross section library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Uncertainties in Predictions of Material Performance Using Experimental Data That Is Only Distantly Related to the System of Interest

There is a need for predictive material “aging” models in the nuclear energy industry, where applications include life extension of existing reactors, the development of high burnup fuels, and dry cask storage of used nuclear fuel. These problems require extrapolating from the validation domain, where there is available experimental data, to the application domain, where there is little or no experimental data. The need for predictive material aging models will drive the need for associated assessments of the uncertainties in the predictions. Methods to quantify uncertainties in model predictions, using experimental data that is only distantly related to the application domain, are discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multiphysics modeling of a critical dual-purpose canister in a saturated geological repository

Dual purpose dry cask storage canisters for spent nuclear fuel are designed for storage and transportation, but are not licensed for permanent disposal in a geological repository. If dual purpose canisters were to be used to dispose of spent nuclear fuel in a geological repository, they would be expected to eventually breach and be flooded with groundwater, and it is shown that some fraction of these canisters will achieve criticality. In order to evaluate the consequences of canisters going critical in a repository, an initial capability has been developed for estimating the quasi-static power level of a critical canister using loosely coupled multiphysics simulations. The low power level in a critical canister enables coupling through precomputed physics proxies. This calculated power level is then used to compute the change in the critical canister’s isotopic inventory as a function of time. Three as-loaded canisters are evaluated and two were found to have power levels below 4 kW, with a modest effect on the radiological inventory over time. This effort also shows that although some DPCs will have extremely peaked power shapes, the relatively low power and long-time scales result in relatively homogenized thermohydraulic properties in the water within the DPC.

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

Force reduction of friction stir welding and processing of steel

Friction stir welding and processing (FSW&P) is considered as a new fabrication method for canisters and repair of existing dry cask storage system canisters (generally made of austenitic stainless steel). Low heat input compared to fusion joining processes and grain refinement associated with low-temperature FSW&P result in a material with improved mechanical properties and equivalent generalized corrosion performance relative to the base metal. But FSW&P of austenitic stainless steel requires significant tool forge axis force to produce a weld (or a run). Exploratory runs utilizing forge axis force control were conducted to investigate the minimum force required to achieve defect-free runs. The correlation between forge axis force and resultant plunge depth was also investigated. Furthermore, detailed microstructural characterization was conducted to display the microstructural variation within the nugget zone.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Economic Losses From a Fire in a Dense-Packed U.S. Spent Fuel Pool

In 2013, the staff of the U.S. Nuclear Regulatory Commission estimated the reduction of the off-site economic losses from a fire in a drained U.S. spent fuel pool if fuel that had cooled for more than five years were transferred to dry cask storage—an option it called “expedited transfer.” In this article, it is shown that the savings would be much higher than the NRC estimated. Savings increase to about $2 trillion if: losses beyond 50 miles are included; the land-contamination threshold for long-term population relocation is changed to that used for the Chernobyl and Fukushima accidents and recommended by the U.S. Environmental Protection Agency; and, based on the experience of Japan, decontamination of land areas to levels acceptable for population return is assumed to take at least four years. If expedited transfer were implemented, the off-site economic losses would be reduced by about 98%.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Development of local-power-free, remote α -particle detection using optical fibers

We demonstrate the application of fluorescence optical fiber coupled to a telecom grade fiber as a sensor for alpha particles using alpha-specific ZnS(Ag) scintillation materials whose wavelength is down-shifted into a low-loss region of the telecom grade fiber transmission band. Telecom-grade fiber optics offer a solution for sensing alpha radiation in deep repositories and cask storage for radioactive materials due to the stability of SiO 2 under normal environmental conditions and its relative radiation hardness at low radiation doses. Long-term nuclear waste storage facilities require sensors for the detection of leakage of radioactive materials that are maintenance-free, do not require power and can survive with no ‘wear out’ mechanisms for decades. By accomplishing the wavelength transformation, we maximize efficiencies in the detection of α-particles and signal transport and can detect alpha scintillation at distances on the order of >1 km with a sensor that is ~3% efficient and can be easily scaled as a sensor array. This paper describes the construction and testing of the sensor including manufacture of the controlled thickness films, verification of the wavelength shift from 450 to 620 nm and optimization of the sensitivity as a function of thickness. We also model the relative sensitivity of the film as a function of film thickness, and we demonstrate a signal-to-noise ratio of 10 at a range of greater than 1 km.

alpha particles↗
DOE OSTI2013

PACKAGE IMPACT MODELS AS A PRECURSOR TO CLADDING ANALYSIS

The evaluation of spent nuclear fuel storage casks and transportation packages under impact loading is an important safety topic that is reviewed as part of cask and package certification by the United States Nuclear Regulatory Commission. Explicit dynamic finite element models of full systems are increasingly common in industry for determining structural integrity during hypothetical drop accidents. Full cask and package model results are also used as the loading basis for single fuel pin impact models, which evaluate the response of fuel cladding under drop conditions. In this paper, a simplified package system is evaluated to illustrate several important structural dynamic phenomena, including the effect of gaps between components, the difference in local response at various points on a package during impact, and the effect of modeling various simplified representations of the basket and fuel assemblies. This paper focuses on the package impact analysis, and how loading conditions for a subsequent fuel assembly or fuel cladding analysis can be extracted.

Klymyshyn, Nicholas A.↗
DOE OSTI2013

Monitoring critical facilities by using advanced RF devices

The ability to monitor critical environment parameters of nuclear plants at all times, particularly during and after a disruptive accident, is vital for the safety of plant personnel, rescue and recovery crews, and the surrounding communities. Conventional hard-wired assets that depend on supplied power may be decimated as a result of such events, as witnessed in the Japanese Fukushima nuclear power plant in March 2011. Self-powered monitoring devices operating on a wireless platform, on the other hand, may survive such calamity and remain functional. The devices would be pre-positioned at strategic locations, particularly where the dangerous build-up of contamination and radiation may preclude subsequent manned entrance and surveillance. Equipped with sensors for β-γ radiation, neutrons, hydrogen gas, temperature, humidity, pressure, and water level, as well as with criticality alarms and imaging equipment for heat, video, and other capabilities, these devices can provide vital surveillance information for assessing the extent of plant damage, mandating responses (e.g., evacuation before impending hydrogen explosion), and enabling overall safe and efficient recovery in a disaster. A radio frequency identification (RFID)-based system - called ARG-US - may be modified and adapted for this task. Developed by Argonne for DOE, ARG-US (meaning 'watchful guardian') has been used successfully to monitor and track sensitive nuclear materials packages at DOE sites. It utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users when any of the preset sensor thresholds is violated. By adding plant-specific monitoring sensors to the already strong sensor suite and adopting modular hardware, firmware, and software subsystems that are tailored for specific subsystems of a plant, a Remote Area Modular Monitoring (RAMM) system, built on a wireless sensor network (WSN) platform, is being developed by Argonne National Laboratory. ARG-US RAMM, powered by on-board battery, can sustain extended autonomous surveillance operation during and following an incident. The benefits could be invaluable to such critical facilities as nuclear power plants, research and test reactors, fuel cycle manufacturing centers, spent-fuel dry-cask storage facilities, and other nuclear installations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

The Effect of Boron on the Thermal Sensitization of Grade ‘A’ Type 304 Stainless Steel

This technical work plan describes experimental work carried out at Los Alamos National Laboratory to evaluate the potential for thermal sensitization of Type 304 borated stainless steel, the leading candidate neutron absorber material for the permanent disposal of used nuclear fuel. This material may also be applicable for very long term storage, particularly if the canister is planned for direct disposal after storage. The workscope described here is new and seeks to provide information as to whether borated stainless steel would become sensitized with extended exposure to relatively high temperatures present in storage casks and disposal containers. This information is necessary to remove some of the conservatism contained in the corrosion allowances that have been applied to the Yucca Mountain Project waste packages, and are currently planned for the proposed can-in-can approach for direct disposal. Currently the corrosion allowance is a factor of nine above the measured corrosion rates for borated stainless steel in the relatively benign environments expected in a failed waste package, which is expected to be similar to dilute water in equilibrium with host rock.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Used Fuel Disposition Campaign: Embrittlement and DBTT of High-Burnup PWR Fuel Cladding Alloys

Structural analyses of high-burnup (HBU) fuel require cladding mechanical properties and failure limits to assess fuel behavior during long-term dry-cask storage and transportation. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). On the basis of previous test results, susceptibility to radial-hydride precipitation depends on cladding material and microstructure, hydrogen content (CH), and pre-drying hydride distribution, as well as peak hoop stresses and temperatures during drying operations and storage. Susceptibility to embrittlement depends on the extent of radial-hydride precipitation, the number density of circumferential hydrides, and the thickness of the outer-surface hydride rim. The goal of the testing program is to determine drying-storage conditions for which the cladding remains ductile at ambient temperature, in other words, with the DBTT ≤20°C.

Billone, M. C.↗
DOE OSTI2012

Aerial Neutron Detection: Neutron Signatures for Nonproliferation and Emergency Response Applications

From 2007 to the present, the Remote Sensing Laboratory has been conducting a series of studies designed to expand our fundamental understanding of aerial neutron detection with the goal of designing an enhanced sensitivity detection system for long range neutron detection. Over 35 hours of aerial measurements in a helicopter were conducted for a variety of neutron emitters such as neutron point sources, a commercial nuclear power reactor, nuclear reactor spent fuel in dry cask storage, depleted uranium hexafluoride and depleted uranium metal. The goals of the project were to increase the detection sensitivity of our instruments such that a 5.4 × 104 neutron/second source could be detected at 100 feet above ground level at a speed of 70 knots and to enhance the long-range detection sensitivity for larger neutron sources, i.e., detection ranges above 1000 feet. In order to increase the sensitivity of aerial neutron detection instruments, it is important to understand the dynamics of the neutron background as a function of altitude. For aerial neutron detection, studies have shown that the neutron background primarily originates from above the aircraft, being produced in the upper atmosphere by galactic cosmic-ray interactions with air molecules. These interactions produce energetic neutrons and charged particles that cascade to the earth’s surface, producing additional neutrons in secondary collisions. Hence, the neutron background increases as a function of altitude which is an impediment to long-range neutron detection. In order to increase the sensitivity for long range detection, it is necessary to maintain a low neutron background as a function of altitude. Initial investigations show the variation in the neutron background can be decreased with the application of a cosmic-ray shield. The results of the studies along with a representative data set are presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Consequence Analysis for Used Fuel Extended Storage

Early identification and evaluation of security issues related to the extended storage of used nuclear fuel is critical. A breach in a dry fuel storage container has the possibility of external gas from the atmosphere interacting with the used fuel rods at high temperatures, resulting in rapid oxidation and possibly the ignition of a zirconium fire. In support of this idea, the current work aims to develop a computational model of heat transfer and fluid flow in and through a breached dry fuel storage cask to determine if the resulting flow conditions are likely to result in a fire.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Remote Whispering Applying Time Reversal

The purpose of this project was to explore the use of time reversal technologies as a means for communication to a targeted individual or location. The idea is to have the privacy of whispering in one’s ear, but to do this remotely from loudspeakers not located near the target. Applications of this work include communicating with hostages and survivors in rescue operations, communicating imaging and operational conditions in deep drilling operations, monitoring storage of spent nuclear fuel in storage casks without wires, or clandestine activities requiring signaling between specific points. This technology provides a solution in any application where wires and radio communications are not possible or not desired. It also may be configured to self calibrate on a regular basis to adjust for changing conditions. These communications allow two people to converse with one another in real time, converse in an inaudible frequency range or medium (i.e. using ultrasonic frequencies and/or sending vibrations through a structure), or send information for a system to interpret (even allowing remote control of a system using sound). The time reversal process allows one to focus energy to a specific location in space and to send a clean transmission of a selected signal only to that location. In order for the time reversal process to work, a calibration signal must be obtained. This signal may be obtained experimentally using an impulsive sound, a known chirp signal, or other known signals. It may also be determined from a numerical model of a known environment in which the focusing is desired or from passive listening over time to ambient noise.

42 ENGINEERING↗
DOE OSTI2016

Physics Flash August 2016

Physics Flash is the newsletter for the Physics Division at Los Alamos National Laboratory. This newsletter is for August 2016. The following topics are covered: "Accomplishments in the Trident Laser Facility", "David Meyerhofer elected as chair-elect APS Nominating Committee", "HAWC searches for gamma rays from dark matter", "Proton Radiography Facility commissions electromagnetic magnifier", and "Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks."

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2017

The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Assessment of Impact of Monoenergetic Photon Sources on Prioritized Nonproliferation Applications: Simulation Study Report

Near-monoenergetic photon sources (MPSs) have the potential to improve sensitivity at greatly reduced dose in existing applications and enable new capabilities in other applications. MPS advantages include the ability to select energy, energy spread, flux, and pulse structures to deliver only the photons needed for the application, while suppressing extraneous dose and background. Some MPSs also offer narrow divergence photon beams which can target dose and/or mitigate scattering contributions to image contrast degradation. Current broad-band, bremsstrahlung photon sources (e.g., linacs and betatrons) deliver unnecessary dose that in some cases also interferes with the signature to be detected and/or restricts operations, and must be collimated (reducing flux) to generate narrow divergence beams. While MPSs can in principle resolve these issues, they are technically challenging to produce. Candidate MPS technologies for nonproliferation applications are now being developed, each of which have different properties (e.g. broad divergence vs. narrow). Within each technology, source parameters trade off against one another (e.g. flux vs. energy spread), representing a large operation space. To guide development, requirements for each application of interest must be defined and simulations conducted to define MPS parameters that deliver benefit relative to current systems. The present project conducted a broad assessment of potential nonproliferation applications where MPSs may provide new capabilities or significant performance enhancement (reported separately), which led to prioritization of several applications for detailed analysis. The applications prioritized were: cargo screening and interdiction of Special Nuclear Materials (SNM), detection of hidden SNM, treaty/dismantlement verification, and spent fuel dry storage cask content verification. High resolution imaging for stockpile stewardship was considered as a sub-area of the treaty topic, as it is also of interest for future treaty use. This report presents higher-fidelity calculations and modeling results to quantitatively evaluate the prioritized applications, and to derive the key MPS properties that drive application benefit. Simulations focused on the conventional signatures of radiography, photofission, and NRF to enable comparison to present methods and evaluation of benefit.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗