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At least 199 records · Page 11

A Low-Tech, Low-Budget Storage Solution for High Level Radioactive Sources

The need for safe, secure, and economical storage of radioactive material becomes increasingly important as beneficial uses of radioactive material expand (increases inventory), as political instability rises (increases threat), and as final disposal and treatment facilities are delayed (increases inventory and storage duration). Several vendor-produced storage casks are available for this purpose but are often costly — due to the required design, analyses, and licensing costs. Thus the relatively high costs of currently accepted storage solutions may inhibit substantial improvements in safety and security that might otherwise be achieved. This is particularly true in areas of the world where the economic and/or the regulatory infrastructure may not provide the means and/or the justification for such an expense. This paper considers a relatively low-cost, low-technology radioactive material storage solution. The basic concept consists of a simple shielded storage container that can be fabricated locally using a steel pipe and a corrugated steel culvert as forms enclosing a concrete annulus. Benefits of such a system include 1) a low-tech solution that utilizes materials and skills available virtually anywhere in the world, 2) a readily scalable design that easily adapts to specific needs such as the geometry and radioactivity of the source term material), 3) flexible placement allows for free-standing above-ground or in-ground (i.e., below grade or bermed) installation, 4) the ability for future relocation without direct handling of sources, and 5) a long operational lifetime . ‘Le mieux est l’ennemi du bien’ (translated: The best is the enemy of good) applies to the management of radioactive materials – particularly where the economic and/or regulatory justification for additional investment is lacking. Development of a low-cost alternative that considerably enhances safety and security may lead to a greater overall risk reduction than insisting on solutions that remain economically and/or politically ‘out of reach’.

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TRANSIENT THERMAL MODELING OF THE HIGH BURNUP DEMONSTRATION RESEARCH PROJECT CASK USING STAR-CCM+ AND COBRA-SFS

The Department of Energy in collaboration with the Electric Power Research Institute is in the process of conducting the High Burnup Demonstration Project. Where the objective is to characterize the performance of high-burnup fuel in long term storage. As part of this demonstration, a TN-32B dry storage cask was instrumented and loaded with spent nuclear fuel at North Anna Nuclear Generating Station in November 2017. The project cask provides a unique opportunity to gain information on spent fuel and cask performance from an in-service operational system. The cask was instrumented with thermocouples inside of the fuel assemblies, then loaded and dried using normal procedures. After the drying process a thermal soak period where the cask was left indoors was used to obtain steady temperatures for model comparison along with surface temperatures. This paper details thermal modeling validation work that was done to model both the steady state and transient cases. Two modeling tools were used to predict temperatures in the cask. The general purpose CFD and heat transfer code STAR-CCM+ was used with both a detailed pin-by-pin model was used along with a more efficient k-effective simplification. The other code is COBRA-SFS, a purpose-built detailed thermal modeling tool developed and maintained at PNNL. Results from all the modeling tools were compared blind to the test data. Each model utilized general design information and compared reasonably well to the blind steady state data using convection and surface temperature boundary conditions. Post-test adjustments were later made to better reflect the “as built” conditions of the cask. Based on lessons learned from the steady state analysis the vacuum drying process was modeled in both codes. Utilizing the measured loading conditions the transient models were able to compare very well with the measured data. Overall the project showed an ability to model spent fuel storage conditions very well and future work is planned to generalize the methodology used for vacuum drying.

Thermal Analysis, Spent Fuel, Dry Storage↗

Going for complete defueling - A comprehensive solution for damaged spent fuel

The complete defueling of a NPP is a prerequisite for decommissioning and dismantling. In particular the defueling of damaged spent fuel and the following dry storage remains a challenge. It is the aim of the Quiver project to develop a one-fits-it-all solution for all different kinds of damaged spent fuel and to develop in addition comprehensive handling equipment to ensure a thoroughly dispatch of the damaged spent fuel inside the Quiver. As a comprehensive solution, GNS developed a Quiver System which complies with the requirements of the transport and storage cask CASTOR V. This paper details, first, the loading of damaged fuel rods into the Quivers including the necessary equipment, secondly, the dispatch of the Quivers including an illustration of the required equipment and facilities, and thirdly, the loading and dispatch of Quivers in the CASTOR V casks.

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Criticality Safety and Fuel Performance Considerations for Enrichment Above 5 Weight Percent in the Uranium Dioxide Fuel Cycle

The U.S. Nuclear Regulatory Commission (NRC) is preparing for anticipated licensing applications requesting fuel enrichments of UO 2 ceramic fuels enriched in excess of 5 weight percent (w/o) 235 U for commercial light-water reactors (LWRs). PNNL has been tasked with providing technical assistance to the NRC related to the review and approval of requests for enrichment increases above the current 5 w/o 235 U enrichment limit. This report will provide the agency with technical assistance to enhance the staff’s knowledge base and to identify developmental needs required to support the licensing of greater than 5 w/o UO 2 ceramic fuels. This report provides a background on increasing enrichment of UO 2 LWR fuel beyond 5 w/o 235 U. The primary concern of increasing enrichment is the impacts to criticality safety. This report will focus on criticality safety as it relates to enrichment, conversion of uranium hexafluoride to UO 2 , fuel assembly fabrication, transportation of fresh/spent fuel assemblies, reactor operations, and storage of fresh/spent fuel assemblies. The report provides a general summary of each of the steps in the UO 2 fuel cycle starting from transportation of enriched UF 6 through dry cask storage of spent fuel assemblies. This report also discusses other impacts of increased enrichment during reactor operation and the surrounding fuel cycle activities. Current needs regarding both data and analytical codes to support the increase above 5 w/o 235 U are identified and discussed.

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

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 framework 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 spent fuel verification technique.

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Evaluation of Nondestructive Examination Responses from Chloride-Induced Stress Corrosion Cracks: Fabrication of Base Metal Test Specimens

This report documents efforts to create actual chloride-induced stress corrosion cracks (CISCCs) in stainless-steel specimens for evaluating application of nondestructive examination (NDE) technologies for detection and sizing of CISCC in stainless-steel canisters in dry storage systems (DSSs) for spent nuclear fuel. In the United States, several DSSs for commercial spent nuclear fuel are approaching the end of their initial licensed or certified term. Many of these systems were originally licensed or certified for 20 years, after which the licensing or certification may be renewed for periods up to 40 years, according to Title 10 of the Code of Federal Regulations (10 CFR), Part 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste.” Guidance for review of renewal applications is provided in NUREG-1927, Standard Review Plan for Renewal of Spent Fuel Dry Cask Storage System Licenses and Certificates of Compliance. A review of aging effects that could affect important-to-safety structures, systems, and components is required in renewal applications and applicants are required to address aging effects using either a time-limited aging analysis or an aging management program. The U.S. Nuclear Regulatory Commission (NRC) issued NUREG-2214, Managing Aging Processes In Storage (MAPS) Report, which provides guidance on the evaluation of known aging degradation mechanisms to determine if they could affect the ability of dry storage system components to fulfill their safety functions in the 20- to 60-year period of extended operation.

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Toward integrating high-fidelity CFD approaches in the thermal-hydraulic analysis of turbulent dry cask systems

Nuclear power plants have been supplying resilient and reliable electricity for decades, contributing to energy independence of the U.S.. However, nuclear waste management remains one of the most significant challenges in the industry. The safety of dry cask storage systems relies heavily on their thermal-hydraulic performance. Computational Fluid Dynamics (CFD) simulations are often used to demonstrate this performance and ensure that the system design meets safety standards. This study presents reduced numerical models for various types of dry cask systems. These numerical models can produce efficient and fast results based on the employed modeling strategies. Additionally, the study uses a novel approach to high-fidelity simulations to evaluate modeling assumptions in dry cask modeling. Large Eddy Simulations (LES) are used for this purpose, particularly in regions where fluid velocity is relatively high and the turbulence characteristics become important. Furthermore, the results of these high-fidelity simulations will enhance the interpretation of outcomes produced from a lower-fidelity CFD model.

CFD↗

Instrumentation: Nondestructive Examination for Verification of Canister and Cladding Integrity. FY2014 Status Update

This report documents FY14 efforts for two instrumentation subtasks under storage and transportation. These instrumentation tasks relate to developing effective nondestructive evaluation (NDE) methods and techniques to (1) verify the integrity of metal canisters for the storage of used nuclear fuel (UNF) and to (2) verify the integrity of dry storage cask internals.

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Monitoring of helium gas leakage from canister storing spent nuclear fuel: Radiological consequences and management

Chloride-induced stress corrosion cracking (CISCC) of canisters has been identified as a potential degradation mechanism requiring ageing management — especially in a marine air environment. Methods for monitoring helium gas leakage from canisters have been developed on the basis of a phenomenon that the bottom temperature of the canister increases as the top temperature of the canister decreases when helium gas leaks during storage. In this paper, we consider the radiological consequences of CISCC based on two scenarios of helium gas leakage and propose effective management that includes a timeline for monitoring and mitigation. Monitoring helium gas leakage by measuring the difference in temperature between canister surfaces in the dry cask storage of spent fuel (SF) can detect leakage before SF undergoes oxidation as a result of air ingress and before country-specific regulatory limits of off-site doses are exceeded, thereby enabling mitigatory action.

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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↗

Evaluation of Neutron Absorber Panels from Zion Spent Fuel Pool

In spent fuel pools (SFPs), fixed neutron absorber materials are used to increase storage capacity while maintaining criticality safety margins. BORAL{sup R} is a commonly used neutron absorber material for both wet (SFPs) and dry storage (casks). The specific questions surrounding fixed neutron absorber materials include the following: 1. What are the conditions of the neutron absorber panels in SFPs? Is there any gross degradation that could cause potential concern for criticality safety of the pools? 2. Is the coupon monitoring approach adequate for monitoring the conditions of the panels as part of an aging management program? 3. For plants that do not have coupons, do current in situ measurement approaches provide accurate results? Could such in situ approaches be used as an alternative monitoring approach? In order to shed light on these technical questions using actual plant data, EPRI initiated the Zion comparative analysis project at the decommissioning Zion Nuclear Power Plant in late 2014. An overview of the Zion plant, SFP, and project was described in a previous publication. (authors)

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Detailed SCALE Dose Rate Evaluations for a Consolidated Interim Spent Nuclear Fuel Storage Facility

This paper presents a confirmatory dose rate calculation for a proposed consolidated interim spent nuclear fuel (SNF) storage facility (CISF). The proposed CISF was modeled in detail using SCALE 6.2.1 to determine the radiation dose map of the surrounding areas of the facility and determine the location of the controlled area boundary based on 10 CFR 72.104 requirements. The spatial dose rate distribution was determined as the total dose rate contributions by each of the 467 storage casks comprising the CISF. A rectangular boundary outside of which no dose rate exceeds 25 mrem/yr has the dimensions of 1.25 km x 1.25 km. It denotes the smallest rectangle that could serve as the controlled area boundary. The results (dose map) show the unique capability of the SCALE code system for a high-resolution radiation dose simulation of a large SNF storage facility.

Radulescu, Georgeta↗

Modeling of Particulate Behavior in Pinhole Breaches

A model is presented for calculating depressurization time for and particulate release from used nuclear fuel dry storage containers that have developed a pinhole breach. Particular attention is given to particulate deposition and transmission within the breach pathway. The model is modular in nature and is developed in a way that allows for more advanced treatments of internal temperature, internal component geometry, or aerosol flow to be readily incorporated. The model can be treated as a basis for addressing concerns associated with monitoring and verification efforts during long-term dry cask storage

Dry Storage↗

A Misload Analysis Methodology Supporting Criticality Analysis of Spent Nuclear Fuel Canisters Using As-Loaded Configurations

This paper presents an assembly misload analysis methodology developed to support criticality safety analysis of spent nuclear fuel (SNF) dual-purpose canisters (DPCs) using as-loaded configurations. The misload analysis approach is based on Interim Staff Guidance 8 “Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks” (ISG 8 rev. 3), but extended to support as-loaded configurations. The two misload scenarios analyzed included; (1) the correct assemblies are selected from the pool but placed incorrectly into the most reactive configuration inside the canister, and (2) the incorrect, most reactive assembly/assemblies in the pool is/are placed into the most reactive position(s) in the canister. Results from this misload analysis approach can be combined with misload probability to support criticality safety assessment of SNF during storage, transportation, and disposal. The misload analysis has been applied to 67 loaded DPCs at two sites using a disposal scenario. The analyzed sites include both PWR and BWR reactors and two canister variants—MPC-32 and MPC-68.

Liljenfeldt, Henrik↗

Using temperature and flow fields to detect gas leakage from canisters containing spent nuclear fuel: Applications to RAMM-TM

The multi-physics STAR-CCM+ code has been used to simulate the temperature and flow fields in canister gas leakage experiments conducted by using a 1/4.5-scale model storage cask. The simulations were conducted at Argonne National Laboratory’s Laboratory Computing Resource Center, utilizing high-performance computing resources. Development of STAR-CCM+ simulation models for the 1/4.5-scale model cask is described herein, followed by validation of the simulation results against experimental data. Canister depressurization and thermal response during gas leakage are discussed, along with analyses of the leakage path and allowable leakage rate. The insights gained from the STAR-CCM+ simulations and leakage analyses will help guide future experiments and actual industry applications of Argonne’s Remote Area Modular Monitoring system for canister surface temperature measurement (RAMM-TM) to enable effective aging management of spent fuel during extended dry storage, as well as help reduce risks to public safety and health and protect the environment.

Canister gas leakage↗

Simulated Gamma Doses to Coating Materials for Use in High Level Waste Packages

An outer shield barrier for spent nuclear fuel is being developed in order to improve dry cask performance under mechanical, chemical, and radiological degradation and stresses. The materials proposed for the outer shield barrier are titanium nitride (TiN), a composite oxide layer consisting of ZrO{sub 2}, TiO{sub 2}, and Al{sub 2}O{sub 3}, and a final layer of molybdenum disulfide (MoS{sub 2}). The ultimate aim of this research is to develop a multi component shield material either as multiple layers or a multi material composite that will improve the performance of steel used in long-term storage casks. Two approaches are being examined through this research effort: multilayer and composite. The composition of the oxide layer is crucial to optimizing the shield's properties and performance. At equal oxide compositions the mixture of these specific oxides is commonly known as zirconolite. This material has been shown to provide a water barrier and to provide substantial corrosion resistance to the dry cask. In order to optimize the radiological performance of this layer the compositions of each oxide have been varied and analyzed for chemical, mechanical, and radiological effects on performance. Through the course of this research effort materials performance changes after gamma and neutron exposure as well as gamma and neutron attenuation of the potential shield materials have been measured and correlated to coating composition. As it is particularly difficult to design a novel, five component material with the goal of improving corrosion, wear behavior, heat tolerance, radiation tolerance, and general survivability, it is necessary to understand which coating components contribute to each property enhancement, and which are critical to design and eventual coating survivability. To compliment the experimental research effort, a study of gamma radiation doses using the relevant zirconolite constituent oxides and assorted combinations off all three, ZrO{sub 2}-TiO{sub 2}-Al{sub 2}O{sub 3}, has been conducted in order to gain insight into which, if any, ratios of the three constituent oxides may provide optimal gamma radiation shielding. Analyzing and comparing the oxides to each other it is apparent that the outer shield barrier with no zirconia has the highest absorbed dose rate at 4.88 E6 mrad/hr. The simulations run with no alumina and no titania are remarkably similar in absorbed dose rates. Alumina and titania compositions increase the absorbed dose rate increase. This is to be expected as during these simulations the zirconia composition is decreasing.

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Code of Record: DOE Standard Canister (DOESC)

The United States Department of Energy (DOE) Spent Nuclear Fuel (SNF) Packaging Demonstration seeks to develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE SNF for road-ready dry storage (RRDS); and establish the processes that will be used in a future production facility. The Packaging Demonstration will utilize the DOE Standard Canister (DOESC) for packaging select DOE-managed SNF types for interim storage, transportation and disposal as part of RRDS. Placing DOE-managed SNF into RRDS is part of the strategic framework for SNF on the Idaho National Laboratory (INL) site. To comply with DOE, INL and Nuclear Regulatory Commission (NRC) requirements, this Code of Record establishes the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 3 as the Code that will govern DOESC and internal support structure constructioni, with certain clarifications. This Code of Record establishes a rationale for proceeding without construction certification of the DOESC (i.e., “N-stamping) and the extent to which a Registered Professional Engineer is required for the DOE Spent Fuel Packaging Demonstration. Given the (i) standard industry practice to pursue independent licensure of commercial storage casks and transportation packages by the NRC in lieu of ASME certification and (ii) guidance provided by 10 CFR 830, DOESC construction activities need not be certified (i.e., “stamped”) to the ASME BPVC. However, additional quality assurance requirements will apply as outlined in this Code of Record.

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History and Status of DOE’s Standardized Canister

[Poster] The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) currently in its possession, as well as SNF that will come into its possession in the future. DOE will rely upon the standardized canister to confine radionuclides and precluding the moderator in order to ensure safe, long-term management and disposition of SNF in its possession. This strategy helps minimize characterization costs (i.e., funds and exposure) and other technical challenges associated with the traditional approach of relying on fuel-specific properties to demonstrate compliance with the regulator. The use of the standardized canister offers the following benefits: • The standardized canister design is complete. It has been tested under a variety of drop and accident scenarios and was subsequently licensed under 10 CFR 72 as a storage system as part of the Foster Wheeler Environmental Corporation (FWENC) Facility. It is also part of the license submittal for Yucca Mountain. • There is high confidence that the standardized canister design can be stored within commercially-available storage casks and licensed as part of a transportation-system package. • A robust standardized canister simplifies accident analyses and does not rely on characterization of the contained waste. This simplifies safety analysis, provides a greater level of confidence with stakeholders and reduces cost and risk. • A common design for the associated packaging, closure, and loadout facilities results in significant savings. • The DOE standardized canister would simplify design efforts, procurement, and fabrication. • Standardizing handling activities and associated training, procedures, equipment, and tooling leads to enhanced safety and surety of operations. The standardized canister is the preferred package for these materials. It withstood scrutiny by the Yucca Mountain Repository License Application, the Nuclear Regulatory Commission’s (NRC) Advisory Committee on Nuclear Waste, the Nuclear Waste Technical Review Board, an American Society of Mechanical Engineers peer review, and the FWENC storage license application. This evaluation summarizes and documents the status of the design and rationale for the DOE standardized canister, and outlines remaining needs before deployment. In addition, it discusses the factors that have the potential to influence storage, transportation, and disposal of DOE SNF.

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