Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “CASKS”

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 109 records · Page 6

Radiation Templates of Spent Fuel in Casks

BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, but baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.

36 MATERIALS SCIENCE↗

Analysis of Gas Samples Taken from the High Burnup Demonstration Cask

The High Burn-Up Demonstration Project was recently initiated by the Department of Energy (DOE) to evaluate the effects of fuel drying and long term dry storage on high burn-up spent nuclear fuel. As part of the project, samples of the He backfill gas were collected 5 hours, 5 days, and 12 days after completion of drying. The samples provide information on the state of the fuel at closure, and on the environment within the cask. At Sandia National Laboratories, the samples were analyzed by gamma-ray spectroscopy to quantify fission product gases and by gas mass spectrometry to quantify bulk and trace gases; water content was measured via humidity probe. Gamma-ray spectroscopy results indicated no detectible 85 Kr, indicating no failed fuel rods were present after drying. Mass spectrometry indicated build-up of CO 2 to 930 ppmv over two weeks, attributed to oxidation of organic compounds (possibly vacuum grease or vacuum pump oil) within the cask. H 2 , generated by either radiolysis or metal corrosion, also increased up to —500 ppmv. Water contents in the cask were higher than anticipated, increasing to —17,400 ppmv ±10% after 12 days. Measuring water content proved challenging, and possible improvements to the method for future analyses are proposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Materials and Dimensional Reference Handbook for the Boiling Water Reactor Dry Cask Simulator.

The performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed analytical modeling of the system’s thermal performance. A recent investigation has been completed that produced a data set that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. To produce these data sets under well-controlled boundary conditions, the dry cask simulator (DCS) was built to study the thermal-hydraulic response of fuel under a variety of heat loads, internal vessel pressures, and external configurations [Durbin and Lindgren, 2017]. The experiments were conducted in Albuquerque, New Mexico where the local ambient atmospheric pressure is typically 83 kPa. The purpose of this handbook is to document the pertinent geometric and material property information needed to perform model validation efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask

This report provides a best-estimate evaluation of residual water content (post-dry out) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask, and evaluates the radiolysis of the residual free water, and the physisorbed and chemisorbed waters on the surfaces of the fuel and cask internal contents. The evaluation of radiolytic breakdown of those waters with gamma radiation causing the generation of hydrogen gas (H 2 ) is made using available literature data and models. This evaluation is part of the overall materials performance evaluation of the SNF-in-canister system, and is part of the technical bases for their continued safe dry storage. The TN-32 cask contents included 32 HBU LWR spent fuel assemblies each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless steel structural components. The residual free and surface (physisorbed/chemisorbed) waters are ascribed to water vapor in the free volume and to components’ surfaces, respectively. The total potential radiolytic hydrogen inventory from the water vapor and from waters ascribed to surfaces has been calculated assuming all the water produced molecular H 2 . The residual water that is chemically incorporated into the bulk of a hydrated oxide, i.e., chemisorbed water, and its total potential hydrogen inventory has been calculated. These calculations are at the physical limit of material available and are used for a bounding assessment purpose only.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

24.1.3.1 - Verification of Spent Fuel Inside Dry Storage Casks by Fast/Epi-thermal Neutron Mapping

Development of a spent-fuel verification method based upon positional measurement of fast neutrons emitted from the top surface of the dry storage casks. A proof-of-principle field test on a partially loaded spent-fuel cask will be conducted by measuring the fast neutron flux on the top of the cask as a function of position using a single detector. If a useful verification signature is demonstrated, a multi-detector scanning array will be built and tested in FY2021 to demonstrate a practical rapid verification system in FY2022.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Verification of Spent Fuel Inside Dry Storage Casks Using Neutron Mapping (2022 Midyear Report)

Development of a spent-fuel verification method based upon positional measurement of fast neutrons emitted from the top surface of the dry storage casks. A proof-of-principle field test on a partially loaded spent-fuel cask will be conducted by measuring the fast neutron flux on the top of the cask as a function of position using a single detector. If a useful verification signature is demonstrated, a multi-detector scanning array will be built and tested to demonstrate a practical rapid verification system.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Verification of Spent Fuel Inside Dry Storage Casks by Fast/Epi-thermal Neutron Mapping (FY2022 NA-241 Annual Report)

Development of a spent-fuel verification method based upon positional measurement of fast neutrons emitted from the top surface of the dry storage casks. A proof-of-principle field test on a partially loaded spent-fuel cask will be conducted by measuring the fast neutron flux on the top of the cask as a function of position using a single detector. If a useful verification signature is demonstrated, a multidetector scanning array will be built and tested to demonstrate a practical rapid verification system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sensor Development for Liquid Water Detection in Dry Storage Casks: FY 2023 Status

Modeling efforts were undertaken in fiscal year (FY) 2023 to evaluate the feasibility and capability for sensing of liquid water inside canister-based dry cask storage systems (DCSSs). The focus was on the development of full-scale finite element models (FEMs) of ultrasound propagation through DCSS canister components. For these initial investigations, simulation of the baseplate component was targeted, envisioning water collection at the bottom of a vertically oriented canister. The effects of internal canister components (i.e., fuel basket, fuel assemblies) on ultrasound propagation in the baseplate component were evaluated in these efforts. The environment inside a DCSS confinement is intended to be inert and free of water to prevent potential corrosion of used fuel cladding or other internal hardware. Spent fuel assemblies are dried, after storage in water pools, to make sure water has been removed from assembly cavities. However, there is some uncertainty about the amount of residual water potentially left behind in a DCSS after drying processes, because water can become trapped in cavities or other small crevices in the surfaces formed by the fuel cladding, fuel assemblies, and other internal hardware components. Considering the complex space- and time-dependent temperature profiles in DCSSs, water may be in a liquid or gas phase depending on its location in the cask and how long the cask has been in storage. Evacuating most water and oxidizing agents contained within a canister is recommended by NUREG-1536 (NRC 2010), which covers DCSSs. As summarized by Salazar et al. (2020), existing guidance typically relies on achievable vacuum pumping pressures sustained over a hold period as a signal of dryness and water removal. However, time series data about water removal from full-scale commercial drying procedures are lacking (Hanson and Alsaed 2019). A review of drying specifications from several vendors led to the conclusion that if the specifications are followed correctly, the residual moisture in DCSSs should present an insignificant risk of cladding degradation (Knoll and Gilbert 1987). A more recent analysis concluded that much larger quantities of residual water could remain in DCSSs, but the amount would still not be expected to lead to significant corrosion of fuel cladding or other internal components (Jung et al. 2013). Industry drying procedures are mostly prescriptive in nature, and operational issues arising during the process could result in incomplete drying. A summary of operational issues and potential negative effects of residual water is provided by Salazar et al. (2019). Experimental efforts are ongoing to validate the extent of water removal in a DCSS based on drying procedures used at nuclear power plants through well-designed investigations of drying process efficacy and water retention (Durbin et al. 2021; Pulido et al. 2022a, Pulido et al. 2022b). This has been approached by simulating limited portions of a DCSS internal volume and fuel assemblies. So far, these efforts have focused on the effects of potential water trapping in the dashpot region of control rod guide tubes but are expected to be expanded to include the effects of other internal hardware features and failed fuel rod cladding. A method for detection and measurement of liquid water (Meyer et al. 2022; Meyer et al. 2021), in tandem with the drying information gained through experimental investigations, provides comprehensive bases for understanding the internal conditions of DCSSs and provides technical information to support licensing decisions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations onspent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC)applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurationsare presented. The considered operating parameters, which were independently varied in thedepletion simulations for the assembly, included fuel temperature, bypass water density, specificpower, and operating history. Different operating history scenarios were considered for the assemblydepletion to determine the effect of relative power distribution during the irradiation cycles, as wellas the downtime between cycles. Depletion, decay, and criticality simulations were performed usingcomputer codes and associated nuclear data within the SCALE code system. Results quantifying thedependence of cask reactivity on the assembly depletion parameters are presented herein.

Ilas, Germina↗

CISCC Countermeasures for Spent Fuel Canisters in Concrete Casks

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

Takeda, Hirofumi↗

Personnel Qualification and Certification Program in Pressure Decay Leak Testing on the RT-100 Type (B) Cask - 17229

The requirements for qualifications and certification of leak test personnel on Type (B) shipping casks have been more clearly defined with the recent release of ANSI N14.5-2014, Leak Tests on Radioactive Packages for Shipment. The latest version creates a standardized approach for qualifying and certifying personnel to conduct pre-shipment leakage tests on Type (B) packages. The U.S Nuclear Regulatory Commission (NRC) is currently updating the Regulatory Guide R.G 7.4, Leakage Tests on Packages for shipment of Radioactive Material, to endorse ANSI N14.5-2014. According to ANSI N14.5-2014, leakage testing shall be performed in accordance with written procedures by qualified and certified personnel in accordance with the requirements of ASNT SNT-TC-1A-2006, Recommended Practice for Personnel Qualification and Certification in Nondestructive Testing. This is a significant modification to the old standard, ANSI N14.5-1997, which simply states that pressure leak testing shall be conducted by trained and certified personnel without reference to any specific certification standards or regulations. SNT-TC-1A published by the American Society for Nondestructive Testing (ASNT) has several advantages to employers; namely, it allows for a Limited NDT Level II certification program. Limited certifications are especially valuable for personnel who possess limited experience and wish to become a certified Limited NDT Level II in a specific technique on a specific cask using a specific procedure and equipment. Robatel Technologies recently developed a comprehensive qualification and training program based on the recommendations of SNT-TC-1A for Exelon Corporation to certify their employees to a Limited NDT Level II in Pressure Decay Leak Testing on the RT-100 Cask. The project scope included the development of a training and qualification program (Written practice), writing a pressure decay leak test procedure, the design of a pressure decay leak test manifold, and the delivery of classroom training. The written practice outlines the requirements to ensure that personnel participating in the program have the necessary qualification, experience, and training to become certified in the leak test method. The program will give Exelon the liberty to conduct their own leak testing on the RT- 100 without the need to rely on outside resources for leak testing. When fully implemented, the program is expected to improve productivity and eliminate schedule delays, along with substantial economic benefits. Following the appropriate standards and industry best practices, the project was successfully completed and scheduled to be fully implemented by Exelon's Nuclear Stations by the end of 2016. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

An assessment of total depletion uncertainties for a GBC-32 dry storage cask with PLUS7 fuels using Monte Carlo sampling and two-way ANOVA F-test

It is important to accurately evaluate depletion uncertainties that exist due to the limitations of a computer code on accuracy in a nuclear depletion calculation for criticality safety analysis. In this work, the bias and bias uncertainty in the effective neutron multiplication factor (k{sub eff}) for the generic 32 PWR-assembly burnup credit (GBC-32) cask system were assessed using the Monte Carlo uncertainty sampling method for three different burnups and the isotopic biases and uncertainties obtained with the calculated and measured isotopic concentrations. The PLUS7 used nuclear fuel assemblies discharged from the Hanbit Nuclear Power Plant Unit 3 were assumed to be loaded into the GBC-32 cask. In addition, the statistical F-test in the two-way analysis of variance (ANOVA) was suggested to efficiently estimate the statistically converged total depletion uncertainty in k{sub eff}, with a significant reduction of computing efforts in the Monte Carlo criticality calculations, and it was applied to the GBC-32 cask system after a sensitivity analysis was performed to select the important nuclides having significant contributions to the total depletion uncertainty. With the statistical F-tests in the two-way ANOVA, 100 criticality calculations were needed to get a statistically converged total depletion uncertainty for 10 and 30 GWD/MTU, while 120 criticality calculations were needed for 50 GWD/MTU. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Benchmarking Data for the Proposed Signature of Used Fuel Casks

A set of benchmarking measurements to test facets of the proposed extended storage signature was conducted on May 17, 2016. The measurements were designed to test the overall concept of how the proposed signature can be used to identify a used fuel cask based only on the distribution of neutron sources within the cask. To simulate the distribution, 4 Cf-252 sources were chosen and arranged on a 3x3 grid in 3 different patterns and raw neutron totals counts were taken at 6 locations around the grid. This is a very simplified test of the typical geometry studied previously in simulation with simulated used nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Momentum informed muon scattering tomography for monitoring spent nuclear fuels in dry storage cask

Development of an effective monitoring method for spent nuclear fuel (SNF) in a dry storage cask (DSC) is important to meet the increasing demand for dry storage investigations. The DSC investigation should provide information about the quantity of stored SNF, and quality assurance of materials should be possible without opening the cask. However, traditional nondestructive examination (NDE) methods such as x-rays are difficult to deploy for DSC investigation because a typical DSC is intentionally designed to shield against radiation. To address this challenge, cosmic ray muons (CRMs) are used as an alternative NDE radiation probe because they can easily penetrate an entire DSC system; however, a wide application of muons is often hindered due to the naturally low CRM flux (~10 4 muons/m 2 /min). This paper introduces a newly proposed imaging algorithm, momentum-informed muon scattering tomography (MMST) , and presents how a limitation of the current muon scattering tomography technique has been addressed by measuring muon momentum. To demonstrate its functionality, a commercial DSC with 24 pressurized light water reactor fuel assemblies (FAs) and the MMST system were designed in GEANT4. Three noticeable improvements were observed for MMST system as a DSC investigation tool: (1) a signal stabilization, (2) an enhanced capability to differentiate various materials, and (3) statistically increased precision to identify and locate missing FAs. The results show that MMST improves the investigation accuracy from 79 to 98% when one FA is missing and 51% to 88% when one-half FA is missing. The advancement of the NDE technique using CRM for DSC verification is expected to resolve long-standing problems in increasing demand for DSC inspections and nuclear security.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

(Project 18-15531) Repair and Mitigation of Chloride-Induced Pitting and Chloride-Induced Stress Corrosion Cracking in Used Spent Nuclear Fuel Dry Cask Canister Materials (Final Technical Report)

Several experimental techniques were utilized to understand the corrosion and SCC performance of possible techniques for repairing cracks and pits in SS304 utilized for dry cask storage of spent nuclear fuel. The repairs examined included cold spray, FSW, soldering, and VFAW. The corrosion and SCC assessment techniques utilized for comparing repairs against the base material were 4-pt bending SCC testing in boiling MgCl 2 at stress levels between 0 MPa and 350 MPa, immersion testing in 40% RH brine solution meant to mimic a service-like environment, SSRT SCC testing in 40% RH brine solution. For FSW alone, scanning electrochemical cell microscopy and microelectrochemical syringe cell were utilized to understand the corrosion performance across the weld, in the retreating/advancing sides, and base material. Based on the results of this study, the FSW and cold spray repair techniques proved most viable for SCC repair in SS304 dry cask storage applications and continuing research is highly recommended to understand and quantify in further detail their SCC performance compared to unrepaired SS304. Additionally, solder and VFAW (patch or incline) would not be recommended based on the corrosion and SCC findings in this study.

42 ENGINEERING↗

TREAT Cask Stand Drawings for Unlimited Use

This report contains drawings of the HFEF-15 Cask stand in the TREAT Reactor Building (MFC-720) and the Cask Stand Tower on the south wall of the TREAT High Bay. Public release is sought so these can be sent to DOD-SCO and their contracted companies looking at options for fueling the PELE reactor prior to demonstration in the DOME or elsewhere at INL.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Fuel Assembly and Irradiation Parametric Study for Extended-Enrichment and High-Burnup Light-Water Reactor Spent Nuclear Fuel in Dry Storage Casks and Transportation Packages

There is an increased interest in operating commercial light-water reactors (LWRs) in the United States with improved economics that would result from longer fuel cycle lengths, fewer and shorter refueling outages, and fewer fuel assemblies requiring storage at the back end of the fuel cycle. To support this, fuel discharge burnups, as well as initial 235 U enrichments, must be higher than those used in current commercial LWRs. The typical upper limit considered for assembly average burnup in this report is 75 gigawatt-days (GWd) per metric ton of uranium (MTU), as opposed to the current typical upper bound of approximately 62 GWd/MTU. The upper limit considered for initial 235 U enrichment is 8 weight percent (8 wt %), as opposed to the current regulatory limit of 5 wt %. The enrichment range from 5 to 8 wt % is referred to in this report as extended enrichment . To investigate the effect of high burnup and extended enrichment conditions on dose rates and burnup credit for dry storage casks and transportation packages, a fuel assembly and irradiation parametric study was performed. The conclusions from this study will assist U.S. Nuclear Regulatory Commission staff in reviewing applications for dry storage casks and transportation packages that contain high-burnup and extended enrichment fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Inerting of Casks Cavity Atmospheres - The GNS Approach to Treat Hydrogen Generation in Sealed ILW Packages - 16043

A new approach to deal with the issue of hydrogen generation in sealed packages for radioactive waste with organic content is presented. The so-called 'inerting' of the cask cavity atmosphere can be either executed after loading or after interim storage via a new facility introduced by GNS. This new approach is approved by the German competent authority and yield to the recent prolongation of the Type B(U) certificate of approval of the GNS MOSAIK{sup R} cask. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗