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At least 163 records · Page 9

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DOE OSTI2015

Results of recent tests of GNS transport and storage casks for nuclear waste - 15285

State of the art dimensioning of casks for nuclear waste is changing rapidly in recent years. Therefore even established packages that serve the market for many years have to face new type testing programs that implements changes to its designs. The paper shows some of the recent testing programs for existing packages and provides an outlook to future casks solutions and its testing programs. (authors)

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DOE OSTI2015

Dry Storage Cask Shake Table Experiments - 15733

Recent seismic events have underlined the risks nuclear dry storage casks (DSCs) are exposed. Dominion Virginia Power's North Ann during Virginia 5.8 earthquake, August 2011: - 25 of 27 DSCs moved on their concrete pad; - DSCs shifted from 1.0 to 4 1/2 inches; - Metal components were not damaged; - No radiation was released. The main goals of this study is to evaluate the long-term seismic performance of DSCs using experimental tests. Through use of Scaled DSC Specimens on the UNR six Degree-of-Freedom (DOF) Shaking Table. Scaled model specimen shall satisfy the following: - Model aspect ratio shall cover the range in the available commercial DSCs; - Scaled 1/2.5 specimens shall be presentative for the actual prototype (specially under vertical excitation); - Designed specimen shall sustain handling, sliding and impact during testing. Selected evaluation basis earthquakes: - Near Field Earthquake (M6.0 at 2 km); - Far Field Earthquake (M8.0 at 20 km). The level of spectral acceleration using NUREG 6728 seismic hazard curves. Developed for the following return periods: - 1,000-year return period; - 30,000-year return period. Design: - Design handling system for overpack and MPC; - Check stresses on the outer cylindrical shell of the overpack in case of using anchorage system for the DSCs; - Check resulted stresses due to impact of the scaled specimen on the supporting concrete pad; - Check safety of shell thicknesses; - Design of connections between different parts; - Design of the concrete pads to resist impact during the test. Future Plans: - Evaluate the long-term seismic performance of the existing dry storage casks in U.S. and develop guidelines for re-licensing operating periods of DSCs. - Provide a performance based design guidelines for DSCs to maintain its safety against earthquakes.

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DOE OSTI2016

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)

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DOE OSTI2016

The Effectiveness of Gd Contents of Fuel Basket Wall in UNF Shipping and Storage Cask - 16598

A more effective basket structure, which employs the single wall basket structure and more effective neutron absorbing materials compared to the conventional 3 wall structure with boron neutron absorber, is under development. The criticality of the cask was evaluated as a function of the Gd and B contents in the plate and the Gd-containing SS plate thickness while maintaining the basic geometry the same as the reference cask which uses boron compound as neutron absorber. The KENO-VI code in SCALE with the data library ENDF/B-VII 238-group was used. In all the range of interest, the k{sub eff} values are below 0.4 which is much lower than licensing limit of 0.95. As Gd content increases, the k{sub eff} exponentially decreases. The Gd added SS shows higher performance than B added case in low atomic percent region. The B case curve showed less steep slope than the Gd case. The results of the wall thickness effect calculation indicated that the k{sub eff} decreases with thickness increase until ∼6 mm and it increases as thickness increases over ∼7 mm. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

The New TN{sup R} MW Cask System Focused on Facilities End-of-Life Waste Management - 16627

When preparing for the decommissioning of a nuclear facility, during its 'end of life' management and while performing the actual dismantling operations, one has to consider a large diversity of nuclear waste in term of types, volumes and activities. Customers are frequently faced with the obligation to undertake multiple and costly waste management operations including handling, reconditioning or re-transferring from one package to another, for example when moving from on-site storage to transportation. To address this issue, a new - highly flexible - cask system named TN{sup R} MW is being developed. This cask has a total weight of 10 T and is compliant with the 2012 IAEA regulations. It is developed on a flexible concept basis, adaptable to the various nuclear needs, including: from IP2 to B(U) / B(U)F; onsite/ international transportation; long term interim storage. Licensing and manufacturing of number of items of this TN{sup R} MW family is underway. (authors)

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DOE OSTI2019

Alternative Impact Limiter Designs for Used/Spent Nuclear Fuel Transportation Casks - 19496

With the need to eventually transport used/spent nuclear fuel (UNF/SNF) from existing storage sites to a consolidated interim storage facility, disposal site, and/or recycling facility, a preliminary evaluation was performed by the University of North Carolina at Charlotte (UNCC) in collaboration with Orano to examine the feasibility of an alternate impact limiter design for Used/Spent Nuclear Fuel Transportation Casks. Many of the previous designs involved complex designs and in some cases, utilized scarce materials, which resulted in expensive impact limiter systems to build and procure. The objective of this evaluation was to examine the availability of alternative materials which could be potentially implemented in simpler designs and result in significant cost savings over some of the current impact limiters designs. The project was divided into three phases: (1) performance of a comprehensive survey of the energy absorbing properties of composite materials that have become available since the original impact limiter designs were licensed; (2) designing and modeling of an impact limiter capable of meeting the hypothetical accident conditions (HAC) identified in 10CFR71.73; and (3) testing of the new materials identified for use in the newly designed impact limiter and feeding this test date to model of the new impact limiter. To design and model the new impact limiter, a specific transportation cask system was selected which was the Orano TN MP187. This paper provides details of the three phases of the project and the estimated cost savings from the new impact limiters. (authors)

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DOE OSTI2016

Shielding analysis of TN-32 spent fuel dry cask with SCALE

The objective of the University of Florida Non Destructive Evaluation (NDE) team is to use Emission Source Tomography to determine and verify the spent fuel location and condition in TN-32 dry casks. In order to investigate this technique, a radiation analysis of TN-32 cask will be performed with MAVRIC control module of SCALE 6.1 and MCNP6 code. The radiation source term from active fuel regions and fuel assembly structure components is calculated, and the MAVRIC geometry model is demonstrated in this paper. Finally, the procedure and future work is listed in this paper. And the shielding analysis result will be presented in ANS winter conference. (authors)

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DOE OSTI2015

Impact of modeling choices on inventory and in-cask criticality calculations for FORSMARK 3 BWR spent fuel - 14441

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Characteristics of Muon Computed Tomography of Used Fuel Casks Using Algebraic Reconstruction

Cosmic raymuons passing through matter lose energy from inelastic collisions with electrons and are deflected from nuclei due to multiple Coulomb scattering (MCS). The recent developments in position sensitive muon detectors that can measure incoming and outgoing trajectories of individual muons indicate that MCS could be an excellent candidate for spent nuclear fuel imaging. The main purpose of this paper is to evaluate tomographic scanning of spent nuclear fuel stored within vertical and horizontal dry storage casks. A quantitative analysis of the characteristics of images obtained with filtered back projection (FBP) and algebraic reconstruction techniques (ART) are presented herein, as such a comparison has not been carried out in the past. FBP is a fast tool to determine object boundaries. ART can include muon path models and prior knowledge that can improve resolution and reduce measurement time. The results demonstrate that missing fuel assemblies can be identified with more than 5 projections and that use of muon momentum significantly increases image resolution. It is expected that MCS can be used to successfully reconstruct the dry cask contents and allow identification of all described scenarios in hours. It is also expected that when total variation minimization and a non-local mean filter are applied, ART may yield much better image quality than FBP.

Liu, Zhengzhi↗

DOE OSTI2015

Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP K eff Calculations for PWR Burnup Credit Casks

The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (k eff ) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Final Technical Report: Imaging a Dry Storage Cask with Cosmic Ray Muons

The goal of this project is to build a scaled prototype system for monitoring used nuclear fuel (UNF) dry storage casks (DSCs) through cosmic ray muon imaging. Such a system will have the capability of verifying the content inside a DSC without opening it. Because of the growth of the nuclear power industry in the U.S. and the policy decision to ban reprocessing of commercial UNF, the used fuel inventory at commercial reactor sites has been increasing. Currently, UNF needs to be moved to independent spent fuel storage installations (ISFSIs), as its inventory approaches the limit on capacity of on-site wet storage. Thereafter, the fuel will be placed in shipping containers to be transferred to a final disposal site. The ISFSIs were initially licensed as temporary facilities for ~20-yr periods. Given the cancellation of the Yucca mountain project and no clear path forward, extended dry-cask storage (~100 yr.) at ISFSIs is very likely. From the point of view of nuclear material protection, accountability and control technologies (MPACT) campaign, it is important to ensure that special nuclear material (SNM) in UNF is not stolen or diverted from civilian facilities for other use during the extended storage.

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DOE OSTI2020

Status Report on Development of a Cask to Enable Pulsed Neutron Characterization of Irradiated Fuel

We present the design of a shielding cask that allows pulsed neutron characterization of irradiated fuel rodlets prior to their destructive examination in hot cells. The goal is to provide complementary and informative measurements that will inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm3 to 1 cm3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging. The capability will identify regions of atypical behavior and provide cost effective bulk characterization of entire samples prior to destructive evaluation in hot cells. This effort is a collaboration between LANL, INL, ORNL, and UC Berkeley. A key facet of the initiative is the ability to cost effectively ship irradiated fuels from INL to the pulsed spallation neutron source at LANSCE. The irradiated fuel would be loaded in the custom designed cask (designated RaMHaM) at INL hot cells which then could be shipped in a BRR Type B shipment container between INL and LANL. No hot cell is required after the loading of the sample at INL in the pathway laid out in this report, greatly simplifying handling at LANL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Microreactor Core Transportation Cask Model Description for Criticality Safety Validation Basis Assessment (Rev. 2)

Criticality safety analyses are completed on transportation casks used for microreactor whole core shipment to provide examples of models and analyses to industry, regulators, and nuclear community at large to be used in verification and validation analyses of similar applications. The microreactors considered are based on a Gas-Cooled Microreactor (GCMR) and a Heat-Pipe Microreactor (HPMR), both utilize HALEU fuel in the form of TRISO particles and various other design options considered in industry microreactor designs. Variant design options of GCMR and HPMR were also investigated to provide a wider application range for each technology. Criticality safety analyses for the GCMR and HPMR packages were performed using the CSAS6 sequence of SCALE 6.3.2 with the ENDF/B-VII.1-based continuous energy neutron libraries. Different scenarios were investigated, including normal operation and water flooded conditions to represent nominal and hypothetical accident scenarios. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE 6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with High Enriched Uranium (HEU), Intermediate and mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems. Many ICSBEP experiments were found marginally similar, with similarity index (ck) values greater than 0.8, to the GCMR and HPMR cask models, especially in flooded conditions that may be more constraining due to reduced reactivity margins. Some experiments using TRISO fuel particles and graphite moderator, including the recent THETA and Deimos experiments from LANL, together with some IRPhEP experiments, were also used for the similarity analyses. These TRISO-fueled and graphite moderated experiments behaved similarly to the GCMR models, with maximum ck value greater than 0.9, but they are less similar to the HPMR models. This observation indicates that additional critical experiments might be needed to further validate the criticality safety models, especially for heat-pipe based microreactor transport packages and for non-flooded transportation configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Dose Rate Analysis Capability for Actual Spent Fuel Transportation Cask Contents

The approved contents for a U.S. Nuclear Regulatory Commission (NRC) licensed spent nuclear fuel casks are typically based on bounding used nuclear fuel (UNF) characteristics. However, the contents of the UNF canisters currently in storage at independent spent fuel storage installations are considerably heterogeneous in terms of fuel assembly burnup, initial enrichment, decay time, cladding integrity, etc. Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is an integrated data and analysis system that facilitates automated cask-specific safety analyses based on actual characteristics of the as-loaded UNF. The UNF-ST&DARDS analysis capabilities have been recently expanded to include dose rate analysis of as-loaded transportation packages. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. This paper describes the UNF-ST&DARDS dose rate analysis methodology based on actual UNF canister contents and presents sample dose rate calculation results.

Radulescu, Georgeta↗

DOE OSTI2020

Thermal Analysis Evaluations Using the Dry Cask Simulator

Different thermal analysis models were developed to simulate the dry cask simulator (DCS). The DCS is an experiment designed to simulate dry storage of a single boiling water reactor fuel assembly under a variety of heat loads and internal pressures. The DCS was set up and tested in both a vertical and horizontal configuration to determine cladding temperatures in vertical and horizontal dry cask storage systems. The models included a detailed STAR-CCM+ model with the fuel assembly geometry explicitly modeled, a porous STAR-CCM+ model with the fuel assembly geometry modeled as porous media region with calculated effective properties, and a COBRA-SFS model. COBRA-SFS is a thermal-hydraulic code developed for steady-state and transient analysis of multi assembly spent-fuel storage and transportation systems. STAR-CCM+ is a commercial computation fluid dynamics (CFD) code. Both a detailed and porous STARCCM+ model were developed to look at the effective thermal conductivity (keff) approach to modeling a fuel assembly. A keff fuel model is typically model in CFD thermal analyses due to its significantly lower computational costs. The models were run for a combination of low and high canister pressures (100 kPa and 800 kPa) and low and high internal heat loads (0.5 kW and 5 kW). Results from all three models were compared against experimental data taken from the DCS for the peak cladding temperature (PCT) and inlet air mass flow.

Suffield, Sarah R.↗

DOE OSTI2017

Extended Storage Considerations for Dry Cask Storage Systems Using Welded Stainless Steel Canisters - 17436

Dry storage of used nuclear fuel is likely to continue for several decades due to the lack of a permanent repository. The Electric Power Research Institute (EPRI) conducted a Failure Modes and Effects Analysis (FMEA) in order to identify degradation mechanisms that are relevant to Dry Cask Storage Systems (DCSSs) which store used fuel within a welded stainless steel canister housed inside a vented concrete cask. Chloride-Induced Stress Corrosion Cracking (CISCC) was identified as the potential degradation mechanism that would be most likely to challenge the confinement function of these canisters. This paper describes research efforts aimed at identifying canisters with CISCC prior to loss of confinement and understanding the consequences should such a loss occur. Key areas of research include development of aging management guidance with recommendations for the scope, frequency, and coverage for canister inspections, demonstration of non-destructive evaluation capabilities, and evaluation of the potential consequences in terms of dose to site personnel and the general public should cracking develop and grow through a canister wall. (authors)

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DOE OSTI2022

Self-powered Through-wall communication for dry cask storage monitoring

Many nuclear facilities, such as spent fuel storage dry casks and nuclear reactor pressure vessels, are entirely sealed by metal layers to prevent harmful radiation. For safety and security operations, the temperature, pressure, radiation, and humidity inside the vessel needs to be closely monitored. However, no practical technology is currently available to realize the through-wall data communication and monitoring for these vessels due to the inside harsh environment of high temperature and nuclear radiation. In this paper, an innovative self-powered wireless through-wall data communication system for the nuclear environment is presented, which demonstrates a successful solution to such challenges. The presented system is composed of four modules, i.e., energy harvester with power management circuits, ultrasound wireless communication using high-temperature piezoelectric transducers, electronic circuits for sensing and data transmission, and radiation shielding for electronics. Here, constitutive functions of each module were firstly designed and followed by the system integration. Experiments were conducted subsequently to validate the designed functions and evaluate the performance of the integrated system. Results showed that the average power of over 40 mW was harvested from the thermal flow inside the nuclear spent fuel canisters which could provide enough energy to operate the sensing and data communication systems. The gamma radiation test results showed that the thermoelectric energy harvester and ultrasound transceivers can withstand radiation dosing over 100 Mrad. Furthermore, temperature shock tests demonstrated that the entire system including the shielded electronics can survive and maintain their functionalities at temperatures as high as 195°C. Under the in-lab mocked-up high temperature conditions and radiation shielding, the proposed system is foreseen to survive and operate stably for fifty years inside a nuclear spent fuel canister, and send the frequency modulated data out of the canister for 3 s in every 10 min.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Development of a novel ultrasonic temperature probe for long-term monitoring of dry cask storage systems

With the recent cancellation of the Yucca Mountain repository and the limited availability of wet storage utilities for spent nuclear fuel (SNF), more attention has been directed toward dry cask storage systems (DCSSs) for long-term storage of SNF. Consequently, more stringent guidelines have been issued for the aging management of dry storage facilities that necessitate monitoring of the conditions of DCSSs. Continuous health monitoring of DCSSs based on temperature variations is one viable method for assessing the integrity of the system. In the present work, a novel ultrasonic temperature probe (UTP) is being tested for long-term online temperature monitoring of DCSSs. Its performance was evaluated and compared with type N thermocouple (NTC) and resistance temperature detector (RTD) using a small-scale dry storage canister mockup. Our preliminary results demonstrate that the UTP system developed at Argonne is able to achieve better than 0.8 Degree-Sign C accuracy, tested at temperatures of up to 400 Degree-Sign C. The temperature resolution is limited only by the sampling rate of the current system. The flexibility of the probe allows conforming to complex geometries thus making the sensor particularly suited to measurement scenarios where access is limited.

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