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DOE OSTI2018

Design and Development of the R79: A new Type B and IP-1-2-3 Transportation Cask for Historic Waste - 18570

In October 2014, Robatel began the design and development of a new Type B and IP-1-2-3 transportation cask, the R79. This cask is designed to transport 'historical' waste from nuclear power plants, nuclear research labs, and nuclear medical isotope production facilities. This 'historical' waste may consist of activated and/or contaminated components including irradiated metal from decommissioning activities, fuel assembly containers, consumables, dried residues, etc. It may also include dewatered ion exchange resins, dried concentrates (salt, metal oxide, sludge, etc.), and hydrogenous materials. The main isotopes in this waste are Co-60 and Cs-137. The cask is not designed to transport fissile material but a trace quantity of up to 22.5 g is allowed. For ease of loading and unloading, the cask is designed to handle waste prepackaged into standard drums. With a loading capacity of 300 kg (661 lb) and a volume of 350 L (12.3 ft{sup 3}), the R79 cask is well suited to carry small to medium-sized waste shipments. The cask body is constructed of 304L stainless steel, its impact limiters are filled with Robatel's FENOSOL{sup TM} foam, and the shielding material includes Robatel's PNT7{sup TM} compound for its thermal and neutron absorbing properties. The cask is loaded in a dry environment and is designed for road transportation. Some novel features were integrated into the design of this cask. With its impact limiters installed, the cask may be treated as a Type B container. Depending on the result of the gas generation analysis, the cask can be treated as either Type B(U) (no limit of the transport duration) or Type B(M) (transport duration limited to 2 weeks). With the impact limiters removed, the R79 cask functions as an industrial-type package (either IP-1, IP-2, or IP-3 depending on the content classification). This can be advantageous for a number of reasons. First, it takes less time for loading and unloading operations when the impact limiters are not installed. Secondly and more importantly, three industrial-type packages can be transported on the same conveyance compared to two Type B packages. These options allow the shipper more flexibility: if the content meets the lower radiation limits for an industrial type package, more waste can be transported on one shipment. If the content must be classified as Type B, the user only must install the impact limiters prior to shipment. The design and development of this cask was performed in 2015, with the first drop testing beginning in July 2015. The complete design documents were first submitted September 2015. Robatel completed the foam for the impact limiters in October 2016, with the completed impact limiters finished in March 2017. The end of manufacturing, delivery of the cask to our customer, and receipt of the regulatory certificate of approval is scheduled before the end of 2017. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Multimodal Nondestructive Dry Cask Basket Structure and Spent Fuel Evaluation

This NEUP-IRP assembled a team from diverse institutions and a diverse set of expertise to examine the problem of assessing the structural health of dry nuclear fuel storage casks, specifically the TN-32 design. With no physical access designed into these storage casks and the high cost and risk associated with opening casks for direct inspection, the technical difficulty of structural assessment of the basket structure and fuel assemblies is high. In response to this challenge, this IRP brought together experts in both traditional and novel NDE methodologies: active acoustics (University of Mississippi), emissions tomography (University of Florida), muon tomography (Oregon State University), and passive-active ultrasonics (University of South Carolina). Further, our industry partners in the Electric Power Research Institute (EPRI) and AREVA provided invaluable insight into technical design details for the cask, practical knowledge of operating parameters at a storage facility, and key relationships that enabled activities unforeseen at the time of the proposed work but were critical in guiding the path of the research. The team decided early in the project that we had a collective goal of laying the technical ground work for practical solutions that could be implemented in the field with reasonable cost, time, and reliability. A fortunate opportunity arose early in this IRP project to obtain access to a new TN-32 cask at Columbiana High Tech. Both acoustics teams were given access for testing concepts in the first year of the project. This ended up providing critical guidance in the design of subsequent laboratory experiments. Specifically, it became clear that the only reliable acoustic pathways to the interior were through the trunnions, base plate, and top ring of the cask. Based on these field tests and discussions about what would be of practical interest, the team as a whole decided a realistic goal was to detect catastrophic failure of a single fuel assembly, or one that was missing. It is clear from this work that the active acoustics / resonance and muon tomography methodologies would be capable of detecting interior structural integrity problems at this level. The other two areas of investigation (emissions tomography and passive/active ultrasonics) would need more testing to determine a successful implementation, but each may also have other applications such as long term monitoring and monitoring during transportation. This IRP project resulted in technical advances with applications that extend beyond the problem at hand, supported the training of 10 graduate students + 1 post doc, 21 publications, and 13 conference presentations. The remainder of this summary highlights significant activities and achievements in each of the methodologies and a brief discussion on how they can be combined to leverage strengths. Active Acoustics The active acoustics team at UM used both linear and non-linear structural resonance methods to infer information about the internal structure of a cask. Significant activities were a set of four field test measurements on a full scale TN-32 cask at the Columbiana Hi Tech facility in Greensboro, NC and the design and fabrication of a 1/6th scale model TN-32 cask for laboratory measurements. A significant set of acoustic data has been acquired on both the full scale and laboratory scale casks. It was determined early in the project that mechanical resonance techniques were not going to be sensitive to small scale (centimeter) cracks and structural defects in the basket structure or fuel assemblies. We were able to ascertain that both the energy and complexity of the resonance spectra were sensitive to “bulk state” of the fuel assemblies – an intact assembly versus a severely degraded assembly. By developing a novel combination of metrics, we are able to measure a statistically significant shift due to one out of the 32 fuel assemblies being either missing or degraded to a point of complete loss of structural coherence (i.e. “rubble”). We believe this combination of resonance metrics is new to the field and will have applications beyond the fuel storage problem. Emissions Tomography The Emissions Tomography team at the University of Florida explored the concept of using scattering of radiation emissions from the stored fuel as method of evaluating the internal structural health of the cask. Due to the need for emissive fuel assemblies and the difficulties of gaining access to storage facilities, this team primarily relied on simulations to inform feasibility. A key question was the flux estimates of neutrons with a small enough number of scattering events to retain spatial information. Some key conclusions from the transmission component of this work are that the number of scattering events falls rapidly with energy – setting a 11MeV threshold, the average number of scattering events is 8∓4. Rough estimates of scanning times using an ideal detector array would be 1 week. The team also explored the case of significant structural failure resulting in a shift of the source material (fuel). The result is that a +/- 2% mass shift results in a detectable (5.13%) shift in the correlated 2 neutron emission counts and thus may be a candidate as an initial survey. Finally, simulations for predicting the azimuthal dependence of the flux based on specific loading patterns were conducted. Due to self-shielding effects of the assemblies, a periodic pattern results and could potentially be used to identify missing or severely degraded assemblies. This work forms a basis for proper detector placement with the goal of optimizing sensitivity. Muon Tomography Muon tomography (MT) is a relatively novel method of interrogating interior structure on a large scale. One of the advantages of this technique is that the exciting energy source (cosmic muons) is ubiquitous and relatively stable in time. Challenging aspects are relatively low flux rates and appropriate detector technologies. The efforts of the muon group at Oregon State University focused on two areas: (1) simulation of various combinations of muon scattering events and detection schemes to assess sensitivities and time scales for image construction and (2) building a prototype muon detection system applicable to the dry storage cask problem as well as many other similar systems. Significant accomplishments in this area are: (1) demonstration that MT is a viable method for detecting missing or severely degraded rod assemblies, (2) a scintillation based detector with 4 panels and 32 channels of position data each were constructed, tested, and calibrated. Testing of the panels using natural muon background and no scattering objects yielded largely expected results – a straight line path for the muon through the four panels. However some anomalies (paths deviating from a line) were also detected. A redesign of the board improved the signal which greatly improved the muon tracking through the panels. Testing with lead bricks as scattering bodies showed a reduction in the coincidence rate with increasing bricks indicating increased scattering events. A refinement of the scintillator detector design resulted in a spatial image resolution of around 2 cm based on lead brick spacing tests lasting 24 hours. Passive/Active Ultrasonics This team at the University of South Carolina investigated the feasibility of combining passive ultrasonics, in which an elastic pulse is generated by a degradation process (such as a crack extending), and active ultrasonics, in which a known excitation is introduced into the structure and detected by an array of transducers. Significant accomplishments in this area include the development of elastic wave propagation simulation in relevant materials, development of diagnostic algorithms, and laboratory testing and validation using laser doppler vibrometry. This group demonstrated the applicability of various non-contact vibration detection methodologies such as air-coupled ultrasonic transducers and scanning laser doppler vibrometry. While passive acoustic emissions monitoring will likely have a role during the transportation of a fuel storage cask, the level of detailed information about the defect generating the event which can be ascertained from the detections in a complex structure remains for future work. Thermography The EPRI team contributed early in the project a series of thermography measurements in which various levels of water pooled in a mock-up canister were analyzed using commercially available infrared imaging technologies. While large levels of water (100 L, ~10 cm deep) were observable in the thermograms, more realistic levels of pooled water (1 L) were not detectable whether localized or distributed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Computed Tomography of Dry Cask Storage for Used Fuel

The demand for dry cask storage for used nuclear fuel is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. Dry cask storage allows used fuel that has already been cooled in the used fuel pool for at least five years to be surrounded by an inert gas inside a specially designed and licensed cask. Currently, 75 of the 78 U.S.A reactor sites either have or are pursuing an independent spent fuel storage installation (ISFSI) for dry cask fuel storage [1] with over 2,000 casks already loaded [2]. Cask loading will only continue to increase until a final repository solution is operational or the U.S. begins to recycle fuel. The challenge of monitoring the condition of the used fuel inside the cask without exposing the used fuel has been apparent since the 1980s. The objective of the University of Florida (UF) project is ultimately to investigate the feasibility of imaging the used fuel assemblies’ location and evaluating the inner structural condition using emission source computed tomography (CT). A combination of radiation transport models will be used to simulate gamma-rays and neutrons traversing through the used fuel, structure, and cask body materials. In the simulations, multiple linear detector arrays will record the radiation outside the cask wall and will be placed at various angles in order to model rotating a linear detector array around the entire cask. The combination of data from all the detectors will give the planar radiation projection onto the detectors, more commonly called the Radon transform, of the structure of used fuel assemblies. This project can be processed with filtered back projection or other more advanced reconstructive imaging algorithms to recreate the fuel structures. A majority of the historical work done for used fuel has been on verification, diversion detection, and burnup credit validation in order to meet regulatory guidelines. One of the more commonly used systems for single assembly verification is the Fork Detector developed at Los Alamos National Laboratory in 1988 [3]. A group at the Idaho National Laboratory developed a Compton Dry-Cask Imaging Scanner that has the ability to detect empty vs full storage positions but it suffered a setback when transported to the Doel Nuclear Power Plant near Antwerp, Belgium; the system was unable to detect full energy peaks because of the additional scattering caused by a thick ballistic steel shield [4]. As detectors, computational power, and modeling improves, a few groups have leveraged these advancements and started to investigate CT hoping, that despite the difficulties, the technique can give information on fuel condition. So far the groups have mostly limited the tomographic work to a per assembly basis [5] [6] [7], rather than the full dry cask as the University of Florida plans to investigate. The downsides of CT versus simpler verification and validation are the amount of data required and the time it takes to acquire that data. Given that dry casks are intended to sit undistributed for up to 60 years, a scanning time of several weeks to several months is not unreasonable if such systems do not require human operators beyond the initial setup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

HBU R and D Cask Transportation Options - 20436

The TN-32B HBU Demonstration Cask was successfully loaded in November 2017 with 32 high burnup (HBU) fuel assemblies at Dominion's North Anna Nuclear Plant and is collecting thermal data as intended. The first two phases of this very successful project sponsored by DOE and EPRI was designed to collect thermal data of HBU fuel to support extended storage. The next phase of the project requires the cask be licensed for transportation so the cask can be relocated to a site (to be determined by DOE) to facilitate long term monitoring and additional research on the fuel. This is a unique challenge given the constraints that exist for this particular bolted metal cask. The TN-32B cask is a 234,288 lb. (storage configuration) dry storage system that was selected for this project and it was modified to provide thermal data and the ability to collect gas samples after loading. Several major issues must be addressed to obtain the transport license including protecting the seven lid penetrations that were made to allow for the insertion of the thermocouples into the selected fuel assemblies after the cask was loaded. In order to protect the integrity of the research, the payload must remain as is. Therefore, the cask system must be placed into the transportation configuration without rewetting the fuel, replacing the lid and without removing the thermocouple instrumentation that is currently installed. A critical portion of the design and engineering that was performed in the first two phases of the project was to ensure that future transportation of the cask was integrated into the approach. Efforts were made to accommodate requirements such as containment boundary leakage testing, installation of impact limiter brackets and lid bolts suitable for transportation. Due to the science requirements of the project, the fuel selected was intended to yield the highest payload temperatures permitted to provide the greatest value for the researchers. This resulted in higher than usual external dose rates and surface temperatures that warranted performing as much preparatory work prior to cask loading to ensure ALARA and personnel safety to the greatest extent practical rather than working on a loaded cask. In addition to the work performed prior to cask load, there are remaining challenges such as protecting the modified lid, design and fabrication of impact limiters that accommodate the modified cask lid and preserving the thermocouple connections for continued monitoring. The project team is currently performing the necessary calculations, engineering, and design to support the licensing as required by the US NRC. Pre-application meetings have been held to inform the regulator of the planned technical approach to ensure compliance with the applicable regulations. The current plan is to obtain the transportation C of C by the end of 2021. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Evaluation of Neutron Flux on the Surface of a PWR Dry Storage Cask - 17574

Since the construction of a dry storage facility for spent nuclear fuel is expected in Korea, several methods have been studied for assuring the safety of pressurized-water reactor (PWR) spent fuels in dry storage casks. Existing evaluation methods are mostly based on detection technology of gamma-rays. However, the gamma-ray detection methods have high uncertainties due to background radiation. Therefore, we developed an evaluation method for the spent nuclear fuel monitoring based on neutron detection. In this study, we evaluated neutron flux on the surface of a PWR dry storage cask using computational simulation. Neutron release rate and energy spectra from a Westinghouse (WH) 17x17 type spent fuel with 10 years cooling time were calculated using ORIGEN-ARP code. MCNP code was used to simulate geometry and material properties of the spent fuel assemblies and casks. Neutron flux after 10 years of cooling time of the spent fuels was calculated on the surface of top, side, and bottom of the casks. The metal cask and the concrete cask, called as KORAD-21 casks which newly developed by Korea Radioactive Waste Agency (KORAD), were selected for this study. As a result, for the metal cask, average neutron flux were 4.79x10{sup 2} neutrons/cm{sup 2} sec, 5.29x10{sup 2} neutrons/cm{sup 2} sec, and 1.47x10{sup 3} neutrons/cm{sup 2} sec on the surface of top, side, and bottom of the cask, respectively. For the concrete cask, average neutron flux on the surface of top, side and bottom of the casks were 5.07x10{sup 1} neutrons/cm{sup 2} sec, 1.18x10{sup 1} neutrons/cm{sup 2} sec, and 1.46x10{sup 3} neutrons/cm{sup 2} sec, respectively. The study results can be applied to evaluate characteristics of PWR spent fuels and the neutron flux of dry storage cask. (authors)

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DOE OSTI2018

Final Scientific/Technical Report, Multi-Sensor Inspection and Robotic Systems for Dry Storage Casks

Once removed from service, fuel rod assemblies are cooled in a pool and then placed in dry storage until they can be transported to a repository for disposal. The extended duration that used nuclear fuel spends in dry storage makes it prudent to be able to nondestructively inspect the dry storage casks. The degradation mechanisms of most concern are chloride-induced stress corrosion cracking of the stainless steel canister that confines the spent fuel and distributed cracking of the concrete in the overpack structure. Multi-sensor inspection and robotic systems were researched, developed, built, and tested. One system operates completely outside the cask to inspect the concrete. The other system, known as the PRINSE system, for Proactive Robotic Inspection of Nuclear Storage Enclosures, enters the cask through an outlet vent and navigates around the outer surface of the stainless steel canister. The PRINSE system delivers sensors for: (i) real-time characterization of the canister surface in order to assess whether chlorides are present, (ii) detecting the presence of cracks using shear-horizontal ultrasonic guided waves, (iii) measuring the temperature of the canister surface, and (iv) a gamma radiation counter. The four sensing modalities are applicable to any cask containing a stainless steel canister, however the robotic delivery system is designed specifically for Holtec’s HI-STORM family of casks. Horizontal axis casks have exposed concrete surfaces. A fully contactless air-coupled scanning ultrasonic test platform and a frequency-wavenumber domain data analysis method are used for these concrete structures. Analysis, numerical simulation and experimental results demonstrate that distributed cracks in concrete can be detected and cracked regions localized within a scanned region of concrete. Vertical axis casks (like the HI-STORM family) have steel-clad concrete overpacks. In a separate effort, a semi-coupled test platform and a guided wave attenuation-based data analysis method are used for disbond detection. Analysis, numerical simulation and experimental results show that the bonding conditions of steel-concrete interfaces can be quantitatively characterized using the guided wave signal analysis scheme. The PRINSE system is designed for vertical axis dry storage casks, specifically the HI-STORM family. The double pulse laser-induced breakdown spectroscopy (LIBS) system takes advantage of an optical fiber to deliver intense light to the canister surface that enables surface composition characterization in-situ in real time. Chlorine concentrations as low as 10 mg/m2 were detected on stainless steel. Additionally, a simpler laser setup was created to detect sodium as a surrogate for chlorine, and sodium was detectable when the chlorine concentration was as low as 0.5 mg/m2. The electromagnetic acoustic transducers (EMATs) enable detection of stress corrosion cracks. The EMATs generate shear-horizontal waves that propagate around the canister and are reflected by cracks enabling 100% of the canister weld lines to be inspected. The EMAT-based crack detection system was tested at the Pacific Northwest National Laboratory on a canister containing blind defects and was shown to detect the crack sizes it was designed to detect. The PRINSE system is a novel marsupial robot. The command center and winch outside the cask control the delivery arm and train of cars that traverse the outside of the canister from the cask ventilation system. The sensors are housed in the cargo bays of three cars that comprise the train, which is lowered to the bottom of the canister by the winch. Encoders on the winch enable the sensors to be delivered to a prescribed location, e.g., for re-inspection of a problem area. The PRINSE system does not mar the canister surface, can be extracted manually in the case of a power failure, and is designed for an elevated temperature and gamma radiation environment, although not all components of the prototype are made from the final materials. The PRINSE system was tested on a full canister that was placed inside a HI-STORM cask at the Holtec Manufacturing Division. It was also demonstrated on a full-height mockup at Penn State to a live audience.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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 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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Validation of the Computational Fluid Dynamics Method using the Aboveground Configuration of the Dry Cask Simulator

Dry cask applicants submit their spent nuclear fuel dry storage and transportation cask designs to the U.S. Nuclear Regulatory Commission (NRC) for certification under 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'. The NRC staff performs its technical review of these designs in accordance with 10 CFR Part 72 and NUREG-1536, 'Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility,' Revision 1, issued July 2010. To ensure that the cask and fuel material temperatures of the dry cask storage system remain within the allowable limits or criteria for normal, off-normal, and accident conditions, the NRC staff performs a thermal review as part of the technical review. Recent applications increasingly have used thermalhydraulic analyses using computational fluid dynamics (CFD) codes (e.g., ANSYS FLUENT) to demonstrate the adequacy of the thermal design. The NRC performs validation studies of the FLUENT CFD code to assist it in making regulatory decisions to ensure adequate protection for storage and transportation casks. 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. In efforts to further validate these designs, the NRC in conjunction with the Department of Energy and Sandia National Laboratories (SNL) are conducting an experiment to capture the effects of elevated helium pressures and accurately portray the external convection of aboveground and belowground canister dry cask systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Developing Design Criteria for Safeguards Seals for Spent Fuel Transportation Casks

Current designs for spent fuel transportation casks cannot ensure a cask's integrity during shipment, nor is there any verifiable means of maintaining continuity of knowledge (CoK) on a cask's contents. Spent fuel destined for encapsulation plants or geological repositories requires additional containment and surveillance (C/S) measures during shipment. Following final safeguards accountancy measurements on spent fuel assemblies, the shipment of verified assemblies will require unprecedented reliance on maintaining CoK on the fuel inside transport casks. Such increased reliance is due to the lack of reverification of spent fuel following encapsulation into disposal canisters and by meeting the requirement of dual C/S measures during such fuel shipments according to recommendations made by the Application of Safeguards to Geological Repositories (ASTOR) International Atomic Energy Agency (IAEA) expert group. By designing spent fuel transportation casks with effective seals integrated into their design, CoK can be more effectively maintained than by ad hoc C/S measures because seal integration ensures that a cask has not been tampered with. Externally applied seals might not be able to provide such assurance for currently designed spent fuel transportation casks, although some combination of seals, detectors, and/or a technology that can verify canister integrity might provide this assurance. This paper examines the design criteria for integrating safeguards seals into transportation casks and provides recommendations for near-term applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Signatures of Extended Storage of Used Nuclear Fuel in Casks

As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each cask's neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Extended Storage of High Burn-up Fuel Cask R and D Project - 17145

Under contract by EPRI and DOE, AREVA Federal Services LLC (AFS) is leading a team consisting of AREVA Inc. and Dominion Virginia Power (Dominion) to develop and conduct a full-scale confirmatory Research project with HBU fuel. A test plan was developed using an existing TN Americas (TN) TN-32 bolted storage cask with a modified, specially instrumented lid to be engineered and then loaded with multiple cladding types of HBU pressurized water reactor (PWR) UNF from Dominion's North Anna Power Station (this modified cask is identified as TN-32B). Seven specially designed thermocouple lances, each containing nine detectors, will be inserted into the designated fuel assembly guide tubes through the modified cask lid. The loaded cask with the monitoring system will be placed into dry storage on the existing North Anna Independent Spent Fuel Storage Installation (ISFSI) pad for a period of extended monitoring. In addition, design modifications to the lid have been incorporated to permit periodic gas sampling of the cask internals during the storage period on the ISFSI pad. The fuel selected for this research project ranges in average assembly burnup between 50 GWD/MTU to 67 GWD/MTU. The instrumentation used to collect the thermal data is very similar to in-core thermocouple detectors used at nuclear power plants and is considered very reliable for this application. Prior to cask loading at North Anna, a series of 'fit up' tests will be performed at the fabricator to ensure proper alignment, fit and function of the modified cask and related instrumentation and equipment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Vibrational metrics for the evaluation of internal conditions in a scaled nuclear dry storage cask

In many countries, permanent repositories for high burnup spent nuclear fuel assemblies have yet to be established. As a result, much of the spent fuel from nuclear reactors is contained within dry storage casks, with many being beyond their designed service life. The assessment of the internal structural integrity of these casks and the fuel assemblies contained therein is of critical importance for both extended storage and transport to permanent repositories. The large size, structural complexity, and inaccessibility of the interior of the casks make this task challenging. To address these difficulties, a 1:6 scaled model based closely on the design of a Transnuclear (TN-32) dry storage cask was fabricated to facilitate controlled studies of these structures in the laboratory. Vibrational spectroscopy was used to evaluate the state of the cargo and internal structures within the cask utilizing only measurements on the outer surface. Using modes identified through Finite Element modeling corresponding to those previously measured on a full-scale TN-32 cask, we report on the development of amplitude- and phase-based metrics that are sensitive to internal conditions in the lab cask. Steel rod bundles and steel shot were used as surrogates for intact and damaged fuel assemblies, and various internal configurations of these materials were investigated. Finally, the metrics were based on acquired spectra involving the (1, 2) global bending mode and the (2, 1) radial-with-shearing mode. The results show that the metrics are sensitive to the condition of a single assembly and have some ability to determine the locations of damaged and empty slots.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗