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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A novel application of cosmic rays for spent nuclear fuel monitoring - 15738

Monitoring nuclear waste and controlling spent nuclear fuel at its source is one of the main strategies to minimize the risks of nuclear proliferation and reduce potential homeland threats. Storage of high level waste from a fission reactor typically involves placing spent fuel in sealed dry casks. After the spent nuclear fuel has been placed inside the dry cask, the cask is welded, not allowing for visual inspection. Conventional methods for examining the interior of materials e.g., X-rays, are limited by the fact that they cannot penetrate very dense well-shielded objects. Instead, cosmic ray muons have the potential to allow for assessment of spent nuclear fuel stored within sealed dense dry casks. Objectives: 1.Exploit the ability of cosmic ray muons to penetrate high density materials allowing the distribution of materials within the object to be inferred from the muon tracks. 2. Develop a monitoring system that tracks the path of cosmic ray muons, including their deflection and attenuation, to identify the contents of a closed dry cask. Methods: - Monte-Carlo simulations using Geant4; - Bayesian decision theory for signal classification; - Wavelet analysis for noise cancellation; - Chaotic Neural Networks for signal encryption. Results: 1. Preliminary statistical analysis proved that the method is feasible and has the ability to identify the presence of spent nuclear fuel in a dry cask. 2. Using monoenergetic muons, a dry cask with one fuel assembly missing can be correctly identified with error rate <1%. Broader merit: Once completed, the proposed project for cosmic ray muon transport will enable a new and potentially transformative approach towards efficient, inexpensive and remote cosmic ray muon applications in archaeology, medicine and safeguarding of nuclear materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

A New Safeguards Technique for Spent Fuel Verification in Dry Storage by Using He-4 Detectors - 18262

It is demanded to find key issues and critical paths for verifying and monitoring a spent nuclear fuel cask or storage with a purpose of non-proliferation of nuclear weapon. However, unfortunately there are no reliable methods to verify the containment in a spent nuclear fuel cask or storage. Therefore, research and development (R and D) has been continuously required to develop new and alternative technologies for searching and verifying undeclared nuclear materials, activities, and facilities. Therefore, the present authors have developed a new safeguards technique of passively scanning a spent fuel cask or storage using pressurized He-4 scintillation neutron detectors. A new approach is to quickly verify content in a spent nuclear fuel cask or presence of all special nuclear materials in unknown packages by comparing experimental and computational results. For that, the present authors have developed a new algorithm and user-interface coupled ORIGEN-ARP and MCNP codes, and the prototype safeguards apparatus employing the array of He-4 fast neutron detectors has been designed, built, and tested. The feasibility test of this new apparatus has been performed with Cf-252 neutron sources and the laboratory scale model cask, minimized by a factor of ten compared to the actual dimension of a KN-21 cask. This technique has several advantages over current methods based on neutron and gamma detection; (1) incident neutron energy information is preserved, (2) disturbance of gamma-ray is excellently rejected (discriminated), and (3) much higher neutron counts and operational cost and space-effectiveness are expected since no moderator is required for fast neutrons. The final goal of this study will (1) conduct a prototype instrument and employ it to monitor a nuclear spent fuel cask or storage at a commercial nuclear power reactor facilities, (2) use spectral-specific measurement data to show proof of concept for measurable data successfully converted to quantifiable signatures, such as diversion/removal of fuel, and maintenance of fuel and cask integrity of fuel, and (3) develop a cask imaging capability. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Regulatory Basis for Dry Spent Nuclear Fuel Storage Canister Inspections

Nuclear power plants in the United States began moving Spent Nuclear Fuel (SNF) from wet storage to dry storage in Independent Spent Fuel Storage Installations (ISFSIs) in 1986 at Surry station in Virginia. Dry storage represented an innovative solution to the problem created by the U.S. government's failure to fulfill its obligation to begin removing SNF from power plant sites. Today, about 70 ISFSIs have been placed into operation on the sites of current, or former nuclear power plants. There are currently over 2,200 SNF dry storage systems (DSSs) in service at these locations. Thirteen of these ISFSIs are located at sites with no operating reactor, and this number is due to increase in the coming years. Operating nuclear power stations use ISFSI storage to maintain wet storage capacity for recently discharged SNF. Permanently shutdown nuclear power plants use ISFSI storage to allow decommissioning and dismantlement of the power plant structures, systems, and components. The U.S. Nuclear Regulatory Commission (NRC) regulates the licensing, construction, and operation of ISFSIs under Title 10 of the U.S. Code of Federal Regulations, Energy, Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class-C Waste (10 CFR 72). When the regulations in 10 CFR 72 were created around 1980, 20 years was chosen by the NRC as the license term for a specific ISFSI license and later, for a DSS Certificate of Compliance (CoC). Twenty years was, at the time, believed to be a sufficient time period for on-site interim dry storage of spent nuclear fuel before the U.S. government began moving the spent fuel away from the plant site for final disposal. Little, if any consideration was given to providing a regulatory framework for extending ISFSI licenses and DSS CoCs beyond 20 years at that time. However, shortly after the turn of the century, with still no federal program for removing SNF from reactor sites on the near-term horizon it became apparent that most DSSs would be required to remain in service at the ISFSIs beyond the initial 20-year license term. Operation beyond the original license term requires the specific license or the DSS CoC (used at a general license ISFSI) to be renewed. The requirements for renewal of a specific ISFSI license or a DSS CoC include addressing age-related degradation of structures, systems, and components that perform or support safety functions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Regulatory perspective of boiling water reactor peak reactivity credit in spent fuel storage and transportation - 14592

The U.S. Nuclear Regulatory Commission (NRC) performs reviews of applications for spent nuclear fuel transportation packages and storage casks to ensure pertinent safety regulations are met. Among other acceptance criteria, these regulations require that the package be subcritical under a variety of normal, off-normal and accident conditions. Currently, all spent nuclear fuel transportation packages and storage casks assume fresh fuel (i.e., the fuel is unirradiated) in their criticality safety analyses for boiling water reactor (BWR) fuels. Reviews using this conservative assumption are less complex. The fresh fuel assumption for BWR fuel does not allow the presence of gadolinium (often called the 'fresh fuel no-Gad' assumption), an integral burnable absorber present in nearly all BWR fuel assemblies. More recently, applicants began requesting credit for the reactivity reduction due to depletion. This credit is commonly referred to as 'burnup credit' (BUC). As of today, NRC recommends that only analyses for Pressurized Water Reactor (PWR) spent fuel implement BUC. This is due to the fact there is limited directly applicable data available to benchmark codes for depletion and reactivity calculations for BWR BUC analyses, and determining the most reactive irradiation conditions is much less straightforward than for PWR BUC analyses. Similar to what was done for PWR BUC, the NRC staff and its contractors have begun identifying and prioritizing significant technical issues so that a technical basis for the allowance of BWR BUC can be developed. BWR fuel exhibits a peak in reactivity where initial gadolinium is depleted. Crediting burnup at this point is generally referred to as using a 'peak reactivity method'. This is a conservative approach and a method routinely used in BWR spent fuel pool criticality analyses. NRC is implementing a two-phased approach in investigating BWR BUC. Phase 1 investigates peak reactivity credit and Phase 2 evaluates BUC at a typical discharge exposure. This paper discusses peak reactivity credit and its history in regulatory analyses, challenges the NRC expects in reviewing an application requesting peak BWR reactivity credit for a spent fuel transportation package or storage cask, the work that is being performed by NRC contractors to address these challenges, and future work on peak reactivity credit and full BWR BUC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Evaluation of PBR Spent Fuel Criticality and Dose Rate Compliance for Storage and Transportation

Spent tri-structural isotropic (TRISO)–based fuels have a strong track record in storage and transportation without documented incidents. This work seeks to reduce uncertainty to aid in more informed spent fuel management of TRISO-based fuels by modeling both fresh and spent pebble bed reactor (PBR) fuel and comparing the results to the regulatory standards from 10 CFR 71. SCALE was used for all modeling due to it having fast and accurate methods for handling PBR fuel modeling, as well as having an efficient method for shielding calculations in monaco with automated variance reduction using importance calculations (MAVRIC), which utilizes the consistent adjoint-driven importance sampling (CADIS) and the forward-weighted consistent adjoint-driven importance sampling (FW-CADIS) methods. KENO-VI was used for all criticality calculations, TSUNAMI was used for uncertainty quantification on k-effective, TRITON and the Oak Ridge isotope generation code (ORIGEN) were both used for depletion of the fuel, and MAVRIC was used for shielding calculations. For criticality assessments, this study focused on the requirement that the value of the neutron multiplication factor, k-effective (k-eff), would not exceed a peak value of 0.95, including uncertainty, with 95% confidence. Criticality was initially examined by modeling fresh fuel from three different designs—HTR-10 fuel, PBMR-400 fuel, and demonstration fuel representative of a TRISO-fueled modern high-temperature gas reactor (HTGR) design, henceforth referred to as Demo HTGR—and placing them into various sized containers with conditions described in 10 CFR 71 to quantify the peak k-eff state. When the peak value of 0.95 k-eff was exceeded, mitigation methods were examined in those scenarios. Burnup credit, pebble displacement in areas of strong neutron multiplication, and random pebble replacement using pebbles of various compositions and replacement fractions were examined. In summary, the criticality of PBR fuels can be well accounted for by restricting container size, taking credit for burnup, or by displacing/replacing pebbles. Uncertainty of the k-eff due to nuclear data uncertainties was recorded at ~0.6644%Δk/k, or roughly 664% mil (pcm). The nuclear data–induced uncertainty was relatively small and should not require significant modification in the design to be accounted for. Revisions to the evaluated nuclear data file values have been shown to have a larger impact than nuclear data–induced uncertainty. For dose rate aspects, U.S. Nuclear Regulatory Commission regulations require a maximum dose rate of 10 millirem per hour (mrem/h) at 2 meters. In examining the dose rate behavior of spent PBR fuel, the representative Demo HTGR fuel was modeled exclusively due to it possessing the highest target burnup of the examined fuels. Equilibrium cycle modeling methods were used to produce a higher-fidelity discharge isotopic composition than simple assumptions, such as reflected pebbles. The discharge composition was used as a source term in the fixed-source transport shielding calculations, and dose rates were calculated at 2 m for the shortest possible cooling time. The low concentration of fuel material led to dose rates that were in line with regulatory limits, despite the high burnup when compared to traditional light water reactor fuels. In conclusion, the methods employed in this study would require more work to further verify and validate and are limited to the criticality and dose rate analyses performed.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Integral nuclear data validation using experimental spent nuclear fuel compositions

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors and representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

An Evaluation of a Neutron Time Correlated Interrogation Method for Measurement of Fissile Content in Research Reactor Spent Fuel Assemblies

There are many research reactors worldwide that have been, or are being converted from the use of high enrichment uranium (HEU) to low enrichment uranium (LEU, < 20% enriched). The verification of fissile content and initial enrichment of the spent fuel is needed for the effective safeguards of the fuel. The advanced experimental fuel counter (AEFC) was developed for the measurement of spent fuel rods and assemblies from research reactors for safeguards verification. This measurement system contains components for active neutron interrogation, passive neutron totals counting, neutron coincidence counting, and gross gamma-ray counting. This report presents the first application of the time correlated interrogation technique for the measurement of the 235 U content in research reactor spent fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252 Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by coincidence counting. The doubles rates are enhanced by having the neutron trigger events from both the 252 Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission of the 252 Cf source is 3.76 and the induced fission neutrons for 235 U is 2.44, so the number of neutrons that are produced is higher than for random neutron interrogation. This high effective neutron number increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252 Cf are reduced by the water in the sample cavity and the polyethylene surrounding the 3 He detector tubes. This method of active neutron interrogation has been applied to the measurement of spent research reactor (IRT) fuel assemblies. The advanced experimental fuel counter (AEFC) was used to compare the TCIF method with the typically used AmLi neutron interrogation source that emits neutrons that are random in time. The statistical uncertainty for the use of the random neutron source (AmLi) and the time correlated source ( 252 Cf) for spent fuel interrogations was evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

An Alternative Conceptual Model for the Spent Nuclear Fuel–Water Interaction in Deep Geologic Disposal Conditions

For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO 2 matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H 2 may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (10 -7 year -1 ) is over estimated because it assumes that there is net UO 2 oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel–water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Sixth National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

The United States of America (U.S.) ratified the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention) on April 9, 2003. The Joint Convention establishes an international peer review process among Contracting Parties and provides incentives for nations to take appropriate steps to bring their spent fuel and radioactive waste management activities into compliance with general safety standards and practices. An extensive U.S. legal and regulatory structure ensures the safety of spent fuel and radioactive waste management. This report describes radioactive waste management in the U.S. in both the commercial and government sectors, and provides annexes with information on spent fuel and waste management facilities, inventories, and ongoing decommissioning projects. It also provides detailed information on spent fuel and radioactive waste management safety, as well as transboundary movements (imports/exports) and disused sealed sources, as required by the Joint Convention.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The Recommendation of Public Engagement Commission on Spent Nuclear Fuel in Korea - 16074

The Public Engagement Commission on Spent Nuclear Fuel Management was lunched and embarked on public engagement activities with an aim of presenting recommendations on the method for managing spent fuel to the government by June 2015. Therefore, a national policy, strategy and management plan for spent fuel will be presented in detail in the Basic Plan for Radioactive Waste Management where the results of public engagement activities will be incorporated. In consideration of Korea's circumstances, permanent disposal facilities should be constructed and operated with the goal of 2051. For this reason, Underground research laboratory (URL) in the same area and on-site and site conditions of disposal facilities by selecting the site and construction process in 2020 should be started and the empirical study should be initiated in 2030. This study is expected to transfer experience of and lessons from determining national policy to countries that have not established their final management policy for spent fuel, based on Korea's solutions to shortage of spent fuel pools and the formation of national consensus. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Multilateral Spent Fuel Management Efforts: Nonproliferation and Energy Implications in the Asia Pacific

Management of spent nuclear fuel in the Asia Pacific will be a growing challenge in coming decades as the region continues its rapid economic and nuclear energy expansion. Accompanying this growth has been skyrocketing energy demand. The capacity of nuclear energy in the Asia Pacific is expected to increase by approximately 50% by 2035. In fact, of the 66 reactors under construction worldwide, 47 are being built in the Asia Pacific region alone. In a sense, the area is experiencing the so-called ‘Nuclear Renaissance’, which could simultaneously increase proliferation risks while fueling economic growth and political stability. And because ‘Asia Pacific’ includes North America, Oceania, East Asia, Southeast Asia and, by some definitions, South and Central Asia as well as the Middle East, the demand for nuclear energy in particular is expected to continue. Consequently, the Asia Pacific region will soon face important decisions about how to support and implement the necessary infrastructure for nuclear energy expansion and deal with the waste, while also meeting all nonproliferation obligations under the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). One option for dealing with spent nuclear fuel that should be considered is multilateral spent fuel management. Such an effort could help reduce risks associated with safety, proliferation and security of nuclear energy. While the original idea of multilateral spent nuclear fuel management is a decades-old concept, no official plan has been developed because of challenges related to cost, liability, transportation, export controls, international commerce, and nonproliferation, to name a few. Nonetheless, as more and more newcomer countries decide to develop nuclear energy, but do not yet have a strategic plan or the financial means to deal with the waste, the international community may need to take a more proactive approach in developing a multilateral effort, particularly with respect to geological repositories and interim dry storage. Risks and benefits will need to be analyzed, while the role of such an effort in the revival of the Nuclear Renaissance will need to be evaluated. This paper provides an overview of spent nuclear fuel management efforts in the Asia Pacific region and attempts to analyze the possible benefits and feasibility of a multilateral approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Numerical Estimation of the Spent Fuel Ratio

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Integrating Management of Spent Nuclear Fuel from Generation to Disposal

The current management system in the United States for commercial spent nuclear fuel does not emphasize integration among storage, transportation, and disposal. Unless a path can be implemented that addresses the long-term needs for integration, the United States could end up leaving substantial quantities of stranded commercial spent nuclear fuel stored at decommissioned reactor sites in an increasingly wide variety of containers. This lack of integration does not cause safety issues, but may delay transporting the spent fuel and complicate options for permanent disposal. The large containers now in use for dry storage remain at high temperatures for decades, thereby delaying transportation from decommissioned reactors. The large containers also have no easy path to disposal unless (1) disposal is further delayed (up to 150 years or more for some mined repository concepts); (2) the contents are repackaged into smaller, cooler packages; or (3) the high temperatures are used as de facto site-selection and design criteria for a repository. Implementing consolidated interim storage could address many issues that exist because of this lack of integration. A consolidated interim storage facility that includes appropriate capabilities can allow existing disparate parts to integrate as a system. Previous agencies and commissions have noted this theme before as a way to provide flexibility in the waste management system. This report uses the rationale for such an approach as a framework to discuss the complexities of reconfiguring the waste management system to include consolidated storage. However, concerns that increased storage capacity will reduce the national urgency for a repository are unavoidable, and continued effort will be necessary in public dialogues on the societal aspects of moving commercial spent nuclear fuel into consolidated interim storage. A single optimal solution for integrating current storage and planned transportation with disposal is unlikely. Rather, efforts to integrate various phases of spent fuel management should begin promptly and continue throughout the remaining life of the current fuel cycle. These efforts will need to adapt continuously to evolving circumstances.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Thermal Modeling of the TN-32B Cask for the High Burnup Spent Fuel Data Project

This report describes thermal modeling done for the High Burnup Dry Storage Research Project that is being performed by the U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) research and development (R&D) program. The purpose of this project to investigate the performance of high-burnup spent nuclear fuel in dry storage. Part of this project is a demonstration test being performed with a storage module in the North Anna Nuclear Power Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an Orano TN-32B High Burnup cask (Fig. S-1) which has been loaded with fuel assemblies from the North Anna spent fuel pool. The main goals of this test are to provide confirmatory data for model validation and potential improvement, support license renewals and new licenses for ISFSIs, and support transportation licensing for high burnup spent nuclear fuel (EPRI 2014).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Validation of Spent Fuel Dose Rate Calculations Using NDA Measurement Campaign Data

The predicted radiation gamma dose rate is a key parameter to determine if a spent nuclear fuel assembly is self-protected from misuse as well as an indicator for worker/inspector safety. Recent studies showed that using state-of-the-art techniques for dose rate calculations, including the one adopted for this study, the numerical calculations of the gamma dose rates are three times lower than those from the most commonly-cited reference values for spent LWR fuel used by domestic and international agencies. With only a few publically available experimental data points worldwide for this type of benchmark study, the gamma flux measurements from the Spent Fuel Non Destructive Assay (NDA) measurement campaign can be used to help validate the physics approach used in calculating gamma dose rates. Specifically, this data would validate the gamma flux calculation from spent fuel, which is a key component of these multi-physics techniques for dose rate assessments; dose rate measurements were not part of the NDA campaign. These data are primarily the passive gamma measurements of the spent nuclear fuel assemblies from the Swedish Nuclear Fuel and Waste Management Company (SKB).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Corrosion-Resistant Coatings on Spent Nuclear Fuel Canisters to Mitigate and Repair Potential Stress Corrosion Cracking (FY22 Status)

This report summarizes the activities performed by Sandia National Laboratories in FY22 to identify and test coating materials for the prevention, mitigation, and/or repair of potential chloride-induced stress corrosion cracking in spent nuclear fuel dry storage canisters. This work continues efforts by Sandia National Laboratories that are summarized in previous reports in FY20 and FY21 on the same topic. The previous work detailed the specific coating properties desired for application and implementation to spent nuclear fuel canisters (FY20) and identified several potential coatings for evaluation (FY21). In FY22, Sandia National Laboratories, in collaboration with four industry partners through a Memorandum of Understanding, started evaluating the physical, mechanical, and corrosion-resistance properties of 6 different coating systems (11 total coating variants) to develop a baseline understanding of the viability of each coating type for use to prevent, mitigate, and/or repair potential stress corrosion on cracking on spent nuclear fuel canisters. This collaborative R&D program leverages the analytical and laboratory capabilities at Sandia National Laboratories and the material design and synthesis capabilities of the industry collaborators. The coating systems include organic (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin), organic/inorganic ceramic hybrids (silane-based polyurethane hybrid and a quasi-ceramic sol-gel polyurethane hybrid), and hybrid systems in conjuncture with a Zn-rich primer. These coatings were applied to stainless steel coupons (the same coupons were supplied to all vendors by SNL for direct comparison) and have undergone several physical, mechanical, and electrochemical tests. The results and implications of these tests are summarized in this report. These analyses will be used to identify the most effective coatings for potential use on spent nuclear fuel dry storage canisters, and also to identify specific needs for further optimization of coating technologies for their application on spent nuclear fuel canisters. In FY22, Sandia National Laboratories performed baseline testing and atmospheric exposure tests of the coating samples supplied by the vendors in accordance with the scope of work defined in the Memorandum of Understanding. In FY23, Sandia National Laboratories will continue evaluating coating performance with a focus on thermal and radiolytic stability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2014

Effects of Spent Cooling and Swirler Angle on a 9-Point Swirl-Venturi Injector

This paper presents multipoint Lean-Direct-Injection (LDI) emissions results for flame tube combustion tests at an inlet pressure of 1034 kPa and inlet temperatures between 835 and 865 K; these are the combustor inlet conditions that the High Speed Research (HSR) program used for supersonic cruise. It focuses on one class of LDI geometry, 9-point swirl-venturi LDI (SV-LDI). Two parameters are compared in this paper: the use of dome cooling air and the swirler blade angle. Dome cooling air is called "spent cooling" and is at combustor inlet conditions. Three cooling variations are studied: cooling at the venturi throat, cooling at the dome face, and no cooling at all. Two swirler blade angles are studied: 45deg and 60deg. The HSR 9-point SV-LDI emissions are also compared to a similar 9-point SV-LDI design which was used in the later ultra-efficient engine technology (UEET) program. The HSR and UEET designs cannot be compared directly due to different UEET combustor conditions. Therefore, this paper uses previously published UEET correlation equations to make comparisons. Results show that using a 45deg swirler produces lower NOx emissions than using a 60deg swirler. This is consistent with the later UEET results. The effects of spent cooling depend on swirler angle, spent cooling location, and the test conditions. For the configuration with 45deg swirlers, spent cooling delivers lower NOx emissions when it is injected at the throat. For the 60deg swirler, spent cooling does not have much effect on NOx emissions. These results might be caused by the location and the intensity of the flame recirculation zone.

He, Zhuohui J.↗