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DOE OSTI2016

Transportation Cask for Bare High Burnup Used Nuclear Fuel - 16362

A conceptual design for a reusable transportation cask (the 6625B-HB) capable of transporting boiling water reactor (BWR) and pressurized water reactor (PWR) used nuclear fuel (UNF) assemblies, including high burnup UNF, was produced for the Department of Energy (DOE) by AREVA. This cask could be configured to ship these assemblies either as bare fuel or fuel loaded into single-assembly damaged fuel cans (DFCs). The 6625B-HB cask has been designed with reasonable assurance it can be certified by the Nuclear Regulatory Commission (NRC) under 10 CFR Part 71 based on AREVA TN's experience with the MP197HB (a transportation cask designed to ship high burnup UNF in a dry storage canister), fabricated within existing facilities, used by most utilities, and transported by rail. The level of detail developed for the 6625B-HB is intended to support analyses and planning activities that DOE is performing in laying the groundwork for an integrated waste management system, which includes preparing for future large-scale transport of UNF to consolidated interim storage or disposal facilities. The details of the conceptual design include cask characteristics (e.g., capacity, dimensions, and masses), characteristics of the UNF that could be shipped in this cask system concept, estimates of costs to fabricate this cask system, an assessment of operational activity durations and associated cumulative doses for loading and unloading this cask system, and identification of potential limitations or anticipated licensing considerations of relevance for this cask system. This design concept for a reusable transportation cask would enhance the high-burnup fuel assembly carrying capacity as compared to other NRC-certified transportation casks and, unlike welded canister systems, could be used for shipment of bare UNF assemblies taken directly from reactor spent fuel pools (SFPs). (authors)

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DOE OSTI2018

Feasibility study for detection of fuel assemblies state inside sealed dry storage casks using external gamma flux measurements - 25445

The license for the storage of spent nuclear fuel assemblies in dry casks is limited to 40 years. In Germany, a site for a final repository of spent fuel is not yet available. Therefore, this license has to be renewed for prolonged storage period. Currently, there is no experience with dry cask storage worldwide that goes beyond a period of 40 years. Beside regulatory and security aspects, there are questions concerning the long-term integrity of the spent fuel assemblies as this is of relevance for final transportation and reloading to final waste repository casks. Once the cask is filled and sealed, the knowledge about the state of the fuel assemblies is limited. This study investigates the feasibility of using external gamma flux measurements to detect the state of the spent fuel assemblies inside the cask. Monte-Carlo simulations were performed to evaluate the gamma flux distribution outside the sidewall of a cask with all intact fuel assemblies and a cask with one damaged fuel assembly, mimicking a fuel assembly with expanded rods and fuel relocation. The evaluation was performed for different location of the damaged fuel assembly and different storage times. The results of the investigation showed that the intensity and the shape of the gamma flux outside the sidewall of the cask are sufficient to identify the damaged fuel assembly, in case it is located near the inner wall of the cask. The detection of the inner damaged fuel assemblies, however, is less feasible. Due to the self-shielding of the fuel assemblies, the gamma photons which come from the inner fuel assemblies stopped within the fuel basket and assemblies materials. The gamma photons which do survive the fuel assemblies self-shielding reach the inner cask wall with not enough energy to escape the shielding of the thick cast iron wall. In general, it can be concluded that the more the damaged fuel assembly is close to the cask inner walls and that the damage is in a large scale (i.e., significant fuel relocation), the more it can be detected by external gamma flux measurements. (authors)

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DOE OSTI2017

Thermal-Hydraulic Results for the Boiling Water Reactor Dry Cask Simulator.

The thermal performance of commercial nuclear spent fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both aboveground and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of this investigation was to produce validation-quality data that can be used to test the validity of the modeling presently used to determine cladding temperatures in modern vertical dry casks. These cladding temperatures are critical to evaluate cladding integrity throughout the storage cycle. 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. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplified interpretation of results. Two different arrangements of ducting were used to mimic conditions for aboveground and belowground storage configurations for vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured throughout the test assembly. The induced air mass flow rate was measured for both the aboveground and belowground configurations. In addition, the impact of cross-wind conditions on the belowground configuration was quantified. Over 40 unique data sets were collected and analyzed for these efforts. Fourteen data sets for the aboveground configuration were recorded for powers and internal pressures ranging from 0.5 to 5.0 kW and 0.3 to 800 kPa absolute, respectively. Similarly, fourteen data sets were logged for the belowground configuration starting at ambient conditions and concluding with thermal-hydraulic steady state. Over thirteen tests were conducted using a custom-built wind machine. The results documented in this report highlight a small, but representative, subset of the available data from this test series. This addition to the dry cask experimental database signifies a substantial addition of first-of-a-kind, high-fidelity transient and steady-state thermal-hydraulic data sets suitable for CFD model validation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and also by increasing the internal convection through greater canister helium pressure. These same canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. The purpose of the current investigation was to produce data sets that can be used to test the validity of the assumptions associated with the calculations used to determine steady-state cladding temperatures in modern dry casks that utilize elevated helium pressure in the sealed canister in an aboveground configuration. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly was deployed inside of a representative storage basket and cylindrical pressure vessel that represents a vertical canister system. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. The arrangement of ducting was used to mimic conditions for an aboveground storage configuration in a vertical, dry cask systems with canisters. Transverse and axial temperature profiles were measured for a wide range of decay power and helium cask pressures. Of particular interest was the evaluation of the effect of increased helium pressure on peak cladding temperatures (PCTs) for identical thermal loads. All steady state peak temperatures and induced flow rates increased with increasing assembly power. Peak cladding temperatures decreased with increasing internal helium pressure for a given assembly power, indicating increased internal convection. In addition, the location of the PCT moved from near the top of the assembly to ~1/3 the height of the assembly for the highest (8 bar absolute) to the lowest (0 bar absolute) pressure studied, respectively. This shift in PCT location is consistent with the varying contribution of convective heat transfer proportional with of internal helium pressure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

3-60B Type-B shipping cask enhanced operational features to facilitate leakage rate testing - 15585

EnergySolutions is fabricating a new 3-60B Type B Shipping Cask that will be ready for use in spring 2015, designed for shipping Type B quantities of activated and/or non-fuel-bearing reactor or accelerator components. EnergySolutions, in consult with Leak Testing Specialists, Inc., has implemented an improved design of the 3-60B cask closures to better facilitate periodic and maintenance leakage rate testing and to allow end users to more easily perform pre-shipment leakage rate testing to confirm proper assembly of the cask closures. A typical Type B cask closure lid design incorporates a double O-ring configuration, where the inner O-ring provides containment and the outer O-ring is used for leakage rate testing of the containment O-ring seal. Leakage rate testing of a double O-ring configuration is typically accomplished by filling the cask cavity with helium, pulling a vacuum on the annular space between the two O-rings and monitoring for helium leakage through the containment O-ring. One issue is the potential for helium saturation of the O-ring elastomeric material, which can result in leakage rate test failures. The cask cavity volume is typically large, and takes a significant amount of time to fill with helium. As a result, the saturation time for certain elastomeric compounds may be exceeded, leading to potential false-negative leakage rate testing results. To avoid this, that the cask cavity typically is loaded with sealed container(s) to reduce the free volume and the fill time. The improved 3-60B cask closure lid incorporates a triple O-ring seal configuration that facilitates maintenance leak rate testing of the closure lid containment O-ring seal and eliminates the issues discussed above. The triple O-ring configuration uses the middle O-ring as the containment seal, with the inner and outer test O-rings creating relatively small annular volumes for testing purposes. These small annular volumes are quick to fill, minimizing the issue of Helium saturation of the O-ring material. This seal configuration also makes it possible to detect and replace a damaged O-ring seal during loading operations, or dress a damaged sealing surface, and perform the required maintenance leakage rate testing of the replacement O-ring at the plant site without the need to unload the cask and send it to a maintenance facility. Also, a second test port has been added to each O-ring annulus to allow leakage rate testing to be accomplished using a formal system calibration that complies with the requirements and recommendations of ASME Section V, Article 10 [4], ANSI N14.5-97 [2], and ANSI N14.5-2014 [3]. This provides a simple, reliable, and accurate means to perform leakage rate testing with a test confidence level that cannot be achieved without the second test port. The improved closure and test port design simplifies and shortens the time required to perform leakage testing, resulting in reduced operating costs, greater schedule security, and dose savings for the cask users. (authors)

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DOE OSTI2016

Investigation of Imaging Spent Nuclear Fuel Dry Casks using Cosmic Ray Muons

Recently, the use of cosmic ray muons for cargo scanning applications has been demonstrated and muons have been shown to have the potential to allow for nondestructive assessment of nuclear material accountancy. Relativistic muons have the ability to penetrate dense materials and monitoring the subsequent scattering of muons can provide information about the structure and composition of materials stored in cargo containers. The suitability of cosmic muons for a number of monitoring and imaging applications has been investigated over the years and includes applications to archaeology, volcano imaging, material identification and medical diagnosis. Novel applications of cosmic ray muons have been proposed for examination of comatose patients towards bone density monitoring and also for determination of the molten nuclear fuel location in nuclear reactors having suffered from the effects of a severe accident similar to the one happened in Fukushima. There is a need to develop a capability to monitor nuclear waste and investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Monitoring nuclear waste and controlling nuclear material at its source is one of the main strategies to minimize the risks of nuclear proliferation and reduce potential homeland threats. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of spent nuclear fuel have been generated, 25% of which are under dry storage conditions. After the spent nuclear fuel has been placed inside the dry cask, the cask is welded, not allowing for visual inspection. During the long storage life of the cask, intermediate handling and transportation activities take place which does create concerns regarding fuel integrity. Accidents and natural disasters also contribute to this concern. Previous work on spent nuclear fuel dry cask monitoring using muons has shown that the muon scattering distributions are adequately separated and a and a decision rule derived from Bayes' Theorem rendering the classification of different dry casks feasible. Muons present significant advantages over the existing imaging techniques such as the utilization of the passive nature of muons, the lack of radiological sources and consequently the absence of any artificial radiological dose. In the present work, a methodology is developed to demonstrate the applicability of muons for imaging of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amounts of loading. Monte Carlo simulations of dry casks with different spent nuclear fuel loading are performed to simulate the passage of polyenergetic muons having energies in the range of 1-60 GeV through matter. A 'Muon Generator' is developed to generate muons that follow the actual sea level energy distribution. The 'Muon Generator' is validated against a series of actual spectrum measurements and then coupled with the Monte Carlo particle transport code GEANT4. The accumulated dataset of scattering angles were processed with the Point-of- the-Closest-Approach (PoCA) imaging algorithm and the obtained images demonstrate that an empty dry cask can be clearly distinguished from a half-loaded or a fully loaded one. The apparent separation of the images demonstrates that the muon scattering can be used as a feature for imaging spent nuclear fuel dry casks. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Classification and imaging of spent nuclear fuel dry casks using cosmic ray muons - 132

The present work explores alternative approaches for in situ monitoring with near real-time data analysis and imaging of dry casks using muons. Monitoring can provide useful signals about the cask contents in a yes/no decision format faster and with fewer muons than traditional imaging. Imaging can be performed as a second step in this process if specific information about the location of the fuel assemblies is needed. To explore and analyze these two modalities, simulations of muon scattering and displacement were performed for various scenarios, including vertical and horizontal fully loaded dry casks, half-loaded dry casks, dry casks with one row of fuel assemblies missing, dry casks with one fuel assembly missing, and empty dry casks. This work demonstrates monitoring can provide fast and reliable indication about the contents of the cask. For 100,000 muons, a few minutes of measurement time with a 3.6 m{sup 2} detector at a 45 deg. zenith angle, 90% detection probability for missing fuel assemblies can be achieved with a false alarm rate lower than 1%. If a positive signal is detected, then longer measurements can be performed and coupled with muon imaging algorithms to determine the location of missing nuclear material. In these cases, transmission muon radiography can produce images of acceptable resolution for all scenarios. In the horizontal dry cask case, all scenarios can be identified and separated adequately. In the vertical dry cask case, similar conclusions can be drawn, but because of shadowing from other fuel assemblies, the location of a missing fuel assembly cannot be identified in the 2-D transmission map. However, views from different directions may resolve this. (authors)

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DOE OSTI2017

Thermal-Hydraulic Results for the Aboveground Configuration of a Dry Cask Simulator

The thermal performance of a commercial nuclear spent fuel dry storage cask is evaluated through detailed analytical modeling. These modeling efforts are performed by the vendor to demonstrate the performance as well as regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). The majority of commercial dry storage casks in use today are aboveground. Cooling of the assemblies located inside the sealed canister is achieved by naturally-convective air flowing in the bottom of the cask and exiting out the top. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating design and performance models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation efforts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. Thus the enhanced performance of modern dry storage casks cannot be fully validated using previous studies.

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DOE OSTI2015

Used Fuel Cask Identification through Neutron Profile

Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Viability of Existing INL Facilities for Dry Storage Cask Handling

This report evaluates existing capabilities at the INL to determine if a practical and cost effective method could be developed for opening and handling full-sized dry storage casks. The Idaho Nuclear Technology and Engineering Center (INTEC) CPP-603, Irradiated Spent Fuel Storage Facility, provides the infrastructure to support handling and examining casks and their contents. Based on a reasonable set of assumptions, it is possible to receive, open, inspect, remove samples, close, and reseal large bolted-lid dry storage casks at the INL. The capability can also be used to open and inspect casks that were last examined at the TAN Hot Shop over ten years ago. The Castor V/21 and REA-2023 casks can provide additional confirmatory information regarding the extended performance of low-burnup (<45 GWD/MTU) used nuclear fuel. Once a dry storage cask is opened inside CPP-603, used fuel retrieved from the cask can be packaged in a shipping cask, and sent to a laboratory for testing. Testing at the INL’s Materials and Fuels Complex (MFC) can occur starting with shipment of samples from CPP-603 over an on-site road, avoiding the need to use public highways. This reduces cost and reduces the risk to the public. The full suite of characterization methods needed to establish the condition of the fuel exists and MFC. Many other testing capabilities also exist at MFC, but when those capabilities are not adequate, samples can be prepared and shipped to other laboratories for testing. This report discusses how the casks would be handled, what work needs to be done to ready the facilities/capabilities, and what the work will cost.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Viability of Existing INL Facilities for Dry Storage Cask Handling

INL, dry storage cask, Idaho National Lab, Long Te↗

DOE OSTI2015

Used nuclear fuel dry storage cask materials monitoring system: a proof of concept to deploy retrieve monitoring equipment - 15504

The cask materials monitoring system is a proof-of-concept to deploy and retrieve monitoring equipment. The system is an airlock-type system that is bolted to the cask fill/drain port. It consists of the following major components: (1) a monitoring module that contains the equipment for monitoring the cask environment, (2) a sliding shield plug that provides radiation shielding throughout the deployment and retrieval of the monitoring equipment module, (3) a primary seal flange that is permanently installed on the cask that provides cask boundary confinement, and (4) a containment tube that provides the airlock function providing cask confinement when the primary seal is broken. In this proof-of-concept development, corrosion coupons were selected as the monitoring equipment for deployment into a spent nuclear fuel cask. However, the system has the potential for a variety of uses. The integration of a camera deployment system into the cask materials monitoring system will permit inspection of used nuclear fuel while maintaining cask atmosphere. (authors)

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DOE OSTI2019

Test Plan for the Horizontal Dry Cask Simulator

The thermal performance of commercial spent nuclear fuel dry storage casks is evaluated through detailed numerical analysis. These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission(NRC). Canistered dry storage cask systems rely on ventilation between the inner canister and the overpack to convect heat away from the canister to the surrounding environment for both horizontal and vertical configurations. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of the investigation described in this test plan is to produce data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator(HDCS) has been designed to generate this benchmark data and add to the existing knowledgebase. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure(MAWP)rating of 2,400 kPa at400 °C. An existing electrically heated but otherwise prototypic boiling water reactor(BWR), Incoloy-clad test assembly will be deployed inside of a representative storage basket and canister. An insulated sheet metal enclosure will be used to mimic the thermal properties of the concrete vault enclosure used in a modern horizontal storage system. Radial and axial temperature profiles along with induced cooling air flow will be measured for a wide range of decay powers and representative(and higher)cask pressures using various backfills of helium, argon, or air. The single assembly geometry with well-controlled boundary conditions simplifies computational requirements while preserving relevant physics. The proposed test apparatus integrates all the underlying thermal-hydraulics important to defining the performance of a modern horizontal storage system. These include combined-mode heat transfer from the electrically-heated assembly to the canister walls and the primarily natural-convective heat transfer from the canister to the cooling air flow passing through the horizontal vault enclosure. The objective of the HDCS is not to reproduce the performance of a commercial dry storage system for any given set of operational parameters. Rather ,the objective is to capture the dominant physics in a well-characterized test apparatus. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate is of particular importance. While incorporating the best available information based on thermal-hydraulic scaling arguments as well as previous vertical testing, this test plan is subject to changes due to improved understanding or from as built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

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DOE OSTI2013

Fundamental study on mitigation of rocking motion of a cask system by a gyro system

In Japan, ensuring the structural integrity of cask systems during seismic events is becoming increasingly important. Cask systems, which are free-standing cylindrical structures that contain spent fuel assemblies, are believed to exhibit rocking motions under strong seismic excitations. Thus far, analytical studies conducted by the authors have indicated that cask systems subjected to strong seismic motions, undergo large rocking motions, and, in the worst case, may overturn and collapse. Therefore, reducing the rocking motions of casks to avoid overturning and consequent contamination of radioactive substances is critical. To suppress rocking motions for heavy free-standing structures such as cask systems, we propose a rocking motion suppression system that employs a gyro system. This system is installed in the free-standing structure. A previous analytical study showed that this system largely mitigates rocking motion. In the present study, we fabricated a fundamental cask system model and a gyro system. By using the cask system model without a gyro system, free vibration tests and shaking table tests were conducted to understand the basic characteristics of rocking motion and responses under base excitations. Analyses were also conducted to confirm the validity of the analytical model for rocking motion comparing with the experimental data. Moreover, analyses for the cask system with the gyro system were conducted. From these results, we evaluated the ability of the gyro system to mitigate rocking motion. (authors)

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DOE OSTI2016

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Results for the Aboveground Configuration of the Boiling Water Reactor Dry Cask Simulator

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Dry coupled magnetostrictive transducers for robotic inspection of dry storage casks - 15304

The lack of a national repository for used nuclear fuel has lead to the usage of dry storage casks as the de facto long term storage solution, the usage of which is prevalent throughout the United States. Furthermore, as shown in figure 3, the usage of dry storage is only expected to increase over time. Because these casks were originally intended as only an interim measure, there is an urgent need to develop methods to monitor their structural health over the long-term. Additionally, when concerns about the susceptibility of these casks to cracking is considered, the need for inspection becomes even more apparent. The use of robotic devices to inspect the interior of these casks for signs of damage has been proposed, but in order for this to be realized, advances in sensing technology must occur. We develop a novel type of transducer for non-destructive evaluation to aid in these efforts. We construct a prototype sensor, demonstrate simple defect detection with it, and present a design for an inspection robot equipping these transducers. Ultrasonic inspection is a well known method of non-destructive evaluation which works by measuring the response of ultrasonic waves propagating through a material. In through-transmission inspection, a transmitter and receiver are used to generate and detect waves in the material to be inspected. The presence of a crack or defect in the material will scatter some of the transmitted wave's energy, reducing the strength of the response at the receiver. Magnetostrictive transducers are often used to generate and detect guided waves. These transducers depend on the magnetostrictive effect, where certain materials will deform in the presence of a magnetic field. The typical construction of one of these transducers is shown in figure 5, consisting of a magnetostrictive foil, meandering coil, and a biasing magnet. Materials typically used for the foil include Fe-Co and Terfenol-D. For wave generation, a high-frequency alternating current is supplied to the coil of the transducer, which generates an oscillating magnetic field that causes the foil to deform. For reception, the inverse occurs: deformation of the foil results in a current in the coil that can be detected. The deformation of the foil is usually coupled to the material to be inspected with an adhesive, and waves are transmitted through this bond. Obviously, an inspection robot cannot have its sensors bound to the surface of the cask. A robot is being designed for inspection, and it must deal with many challenges inside the cask, including tight confines, right-angle transitions, high temperatures, and radiation. For our initial design, we consider only the geometrical constraints, and present a design for a prototype that can be assembled from COTS components. The robot needs to be articulated to navigate the bends in the cask, so each wheel is individually actuated. The robot is controlled via a tether, which also supplies electrical power and compressed air. For localization, ultrasonic range-finders and a camera are equipped. Using the basic transducer, we experimented with various configurations of materials and geometry. In order to gather relevant data, we used force sensitive resistors to measure applied force, and a Matec computer system to generate and measure ultrasonic waves. A testing rig was constructed to simulate the geometrical constraints of the interior of the cask, and we experimented using deadweights, a pneumatic cylinder, and an air bladder to apply the force. It was discovered early in the testing process that the performance of the transducer was very sensitive to the alignment of the components, and the method by which the foil was attached. To maximize performance, we designed an enclosure that would hold the components in alignment while still allowing the foil to float. To test the effectiveness of our new transducer, we cut circular holes of varying diameter in 2 mm aluminum plate, and measured the response across them. It can be seen that there is a clear relationship between the strength of the received signal and the force applied, and to a lesser extent, the size of the defect. This design performs adequately as a proof of concept, and a more precisely fabricated enclosure should further increase the accuracy of its measurements.

42 ENGINEERING↗

DOE OSTI2016

Shipping Grossly Dewatered resins in the RT-100 Type B Transportation Cask - 16068

Ion-exchange resins are used in the purification of coolant water in boiling water reactor nuclear power plants. The spent resins are a form of low-level or intermediate-level radioactive waste. These resins may have a relatively high specific activity level and are normally transported to a disposal facility within a secondary container placed in a Type B transportation cask. During transportation, hydrogen gas may build up in the cask due to radiolysis of the hydrogenous material in the contents. In the licensing process, a Type B cask is carefully analyzed to ensure that the maximum amount of hydrogen gas generated does not exceed 5%. The calculation to determine the mole fraction of hydrogen gas generated is a function of many variables, including the properties of the contents and the volume ratios of the contents. Because water generates hydrogen gas when it undergoes radiolysis, the amount of water present in the waste is normally limited. Typically, Type B transportation casks are allowed to carry dewatered resins, meaning less than 1% free-standing water by volume. To ensure compliance with this limitation, the dewatering process requires specialized equipment to pump the water from the secondary container. In an effort to enhance the usability of the RT-100 Type B transportation cask, Robatel Technologies re-analyzed the hydrogen gas generation calculation for contents including 'grossly dewatered' resins. Grossly dewatered resins are defined as resins containing less than 20% free-standing water by volume. This higher limit is made possible by a more thorough hydrogen gas generation calculation, which ensures that the flammability limit is not reached during shipment. In July 2015, The USNRC issued Revision 1 of the Certificate of Compliance for the Robatel Technologies RT-100 Type B transportation cask. This revision allows the cask to be used for grossly dewatered resins. With the revised Certificate of Compliance, the RT-100 is the first of its kind whose authorized contents include greater than 1% free liquids by volume. As a result of this change, the cask user can meet the 20% maximum free-standing water volume with a much simpler gravity-driven dewatering process. It is no longer necessary to have a complex dewatering system to ensure compliance with regulatory requirements. (authors)

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