Spent Nuclear Fuel Transportation Research Articles

Computation of Particulate Deposition on Dry Storage Canisters

This paper provides an overview of ongoing work aimed at developing spent nuclear fuel (SNF) canister deposition models. Currently, it is known that stainless steel canisters are susceptible to chloride-induced stress corrosion cracking (CISCC). However, the rate of CISCC degradation and the likelihood that it could lead to a through-wall crack is unknown. While it is currently unknown if there is a threshold chloride surface concentration for CISCC initiation, it can be assumed that the onset and progress of material degradation will depend on the local contaminant concentration, the properties of the contaminant species, and the synergistic effects when multiple contaminants are present. This study uses well-developed computational fluid dynamics and particle tracking tools and applies them to SNF storage to determine the rate of deposition on canisters. Understanding the rate of deposition on SNF canisters could be important for making canister aging management predictions. This study is a part of an ongoing effort funded by the U.S. Department of Energy, Office of Nuclear Energy, Office of Spent Fuel and Waste Science and Technology, which is tasked with doing research relevant to enhancing the technical basis for ensuring the safe extended storage and subsequent transport of SNF. This work is being presented to demonstrate a potentially useful technique for SNF canister vendors, utilities, regulators, and stakeholders to utilize and further develop for their own designs and site-specific studies.

Rolling Test Simulation of Sea Transport of Spent Nuclear Fuel Under Normal Transport Conditions

Rolling Test Simulation of Sea Transport of Spent Nuclear Fuel Under Normal Transport Conditions

Package Performance Study: A Historical Perspective and Planning for a Path Forward

The topic of spent nuclear fuel (SNF) transportation often draws public attention and concerns regarding safety, and with large-scale shipping campaigns expected in the future, public interest is anticipated to substantially increase. In 1999, the U.S. Nuclear Regulatory Commission (NRC) began an initiative called the Package Performance Study (PPS) that initially focused on SNF transportation cask responses to severe transportation accidents. A public participation approach was to be used to help guide the scope and parameters of the study, which was also to consider using full-scale physical testing of a SNF transportation cask, where appropriate. In 2010, the NRC decided not to go forward with physical testing as proposed. In 2006, the National Academy of Sciences (NAS) conducted a study on SNF and high-level radioactive waste transportation in the United States. The NAS study recommended that full-scale cask testing, as well as other accepted methodologies, continue to be used as part of the package performance evaluation. In 2012, the Blue Ribbon Commission on America’s Nuclear Future (BRC) recommended conducting the PPS with a full-scale rail transportation cask for the purpose of building public trust and confidence in the safety of SNF transport. In 2021, the U.S. Department of Energy (DOE) restarted a consent-based siting process, focusing on how to site federal facilities for the interim consolidated storage of SNF using a consent-based approach. Large-scale SNF transportation would be necessary for any such interim storage facility. As the planning activities begin for future large-scale shipping campaigns, the DOE is considering options for conducting a DOE-led PPS to help build public trust and confidence in the transportation of SNF. Therefore, the DOE is gathering information related to the previous NRC-led PPS efforts, as well as applicable international experience with the testing of SNF packages. This paper discusses previous NRC efforts, as well as the current status of a DOE-led PPS, which is in the early planning stages.

Neutron background of composite low-enriched uranium fuel of the IVG.1M research reactor

IVG.1M is a research pressurized water reactor designed to use high-enriched fuel. As part of the core conversion program, the reactor will be switched to a new low-enriched composite uranium fuel. Further operation of the reactor is determined by the availability of fresh fuel to replace the core after the next campaign and the possibility of ensuring safe storage of irradiated spent nuclear fuel (SNF) unloaded from the core. The SNF storage conditions are assessed in terms of ensuring nuclear and radiation safety. Radiation safety of the research reactor fuel storage is achieved, first of all, by solving problems of protection against γ-radiation, while neutron radiation, as a rule, is not considered due to its significantly lower intensity compared to γ-radiation. As for the new low-enriched fuel of the IVG.1M reactor, which is characterized by a set of elements with low and medium atomic masses, on which the (α, n) reaction is possible, the assessment of the neutron component is a necessary procedure to ensure safe fuel storage. The authors of the article propose a procedure for calculating the neutron component of the radiation characteristics of fresh and irradiated composite fuel of the IVG.1M reactor, and also estimate the (α, n)-component. The results of the research will be useful in selecting SNF storage and transportation technologies as well as in providing scientific justification for the possibility of using neutron radiation to control burnup. The research was carried out using verified computational codes MCNP5 and Sources-4C, high-precision experimental EXFOR and computational ENDSF data, as well as evaluated nuclear data libraries.

Sensitivity of SNF transport cask response to uncertainty in properties of wood inside the impact limiter under drop accident conditions

It is essential to ensure the safety of spent nuclear fuel (SNF) transport cask in drop situation that is included in transport accident scenarios. The safety of the drop situation is affected by the impact absorption performance of impact limiters. Therefore, when designing an impact limiter, the uncertainty in the material properties that affect the impact absorption performance must be considered. In this study, the material properties of the wood inside the impact limiter were selected as the variables for a parametric study. The sensitivity analysis of the drop response of the SNF transport cask with impact limiter was performed. The minimum wood strength required to prevent a direct collision between the cask and floor was derived from the analysis results. In addition, the plastic strain response was analyzed and strain-based evaluation was performed. Based on this result, the critical values of wood properties that change the impact dynamic characteristics were investigated. Finally, the optimal material properties of wood were obtained to secure the structural safety of the SNF transport cask. The results of this study can contribute to the development of SNF transport cask, thereby ensuring safety in transport accident conditions.

Preliminary Modeling of Chloride Deposition on Spent Nuclear Fuel Canisters in Dry Storage Relevant to Stress Corrosion Cracking

This study presents a method that can be used to gain information relevant to determining the corrosion risk for spent nuclear fuel (SNF) canisters during extended dry storage. Currently, it is known that stainless steel canisters are susceptible to chloride-induced stress corrosion cracking (CISCC). However, the rate of CISCC degradation and the likelihood that it could lead to a through-wall crack is unknown. This study uses well-developed computational fluid dynamics and particle-tracking tools and applies them to SNF storage to determine the rate of deposition on canisters. The deposition rate is determined for a vertical canister system and a horizontal canister system, at various decay heat rates with a uniform particle size distribution, ranging from 0.25 to 25 µm, used as an input. In all cases, most of the dust entering the overpack passed through without depositing. Most of what was retained in the overpack was deposited on overpack surfaces (e.g., inlet and outlet vents); only a small fraction was deposited on the canister itself. These results are provided for generalized canister systems with a generalized input; as such, this paper is intended to demonstrate the technique. This study is a part of an ongoing effort funded by the U.S. Department of Energy, Nuclear Energy Office of Spent Fuel Waste Science and Technology, which is tasked with doing research relevant to developing a sound technical basis for ensuring the safe extended storage and subsequent transport of SNF. This work is being presented to demonstrate a potentially useful technique for SNF canister vendors, utilities, regulators, and stakeholders to utilize and further develop for their own designs and site-specific studies.

Thermal management associated with geologic disposal of large spent nuclear fuel canisters in tunnels with thermally engineered backfill

Coupled thermal-hydro-mechanical (THM) numerical modeling is conducted to study thermal management associated with geologic disposal of spent nuclear fuel (SNF) in large dual-purpose canisters (DPCs). DPCs, are containers designed for SNF storage and transportation and if determined to be feasible for permanent geological disposal could provide a cost effective disposal solution. However, one of the challenges to direct disposal of DPCs is thermal management to avoid overheating of the Engineered Barrier System (EBS), including bentonite-backfill used as a protective buffer. The model simulations show that the use of a backfill that is thermally engineered for high heat conduction can reduce the EBS temperature to acceptable levels for disposal of large waste canisters in backfilled tunnels. On the other hand, the use of high heat conduction backfill will not reduce the far field rock peak temperature that can occur several thousand years after closure of the repository. This longer term host rock peak temperature generates thermal-poro-elastic stress and geomechanical changes that must be considered in the thermal management and design of a repository.

Development of Equivalent Beam Model of High Burnup Spent Nuclear Fuel Rods under Lateral Impact Loading

Spent nuclear fuel (SNF) is nuclear fuel that has been irradiated and discharged from nuclear reactors. During the whole management stages of SNF before it is, in the end, disposed in a deep geological repository, the structural integrity of fuel rods and the assemblies should be maintained for safety and economic reasons. In licensing applications for the SNF storage and transportation, the integrity of SNF needs to be evaluated considering various loading conditions. However, this is a challenging task due to the complexity of the geometry and properties of SNF. In this paper, a simple and equivalent analysis model for SNF rods is developed using model calibration based on optimization and process integration. The spent fuel rod is simplified into a hollow beam with a homogenous isotropic material, and the model parameters thus found are not dependent on the length of the reference fuel rod segment that is considered. Two distinct models with different interfacial conditions between the fuel pellets and cladding are used in the calibration to account for the effect of PCMI (Pellet-Clad Mechanical Interaction). The feasibility of the models in dynamic impact simulations is examined, and it is expected that the developed models can be utilized in the analysis of assembly-level analyses for the SNF integrity assessment during transportation and storage.

Problems of radiation safety calculations related to spent nuclear fuel transport casks

The paper discusses the stages of calculating the radiation safety of spent nuclear fuel (SNF) transport packages, in particular, transport casks and some related problems. The problem of describing the source of neutrons and gamma radiation of spent nuclear fuel is shown. For individual designs of fuel assemblies, data are given on isotopes that make the main contribution to the neutron source as well as on gamma rays in nuclear fuel material and structural materials. The authors emphasize the necessity of analyzing the influence of the initial spent fuel parameters on the formation of the radiation spectrum and, therefore, on the radiation situation around the transport casks. Consideration is given to the problem of assessing the attenuation of gamma radiation in calculating protection analytically and using software. Due to the ambiguity of the position of the zone with the highest effective dose value on the SNF transport cask surface, it is indicated that preliminary estimates are required to take into account all radiation sources and their nonuniformities. All the problems presented in the paper are currently being solved by means of rather complex and voluminous calculations that take a long time. In order to be able to conduct a preliminary assessment of the radiation situation around the transport casks, the authors propose to create a methodology that will determine the type of interrelations between the maximum effective dose and input parameters, such as fuel burnup, decay, fuel composition, protection material in the SNF transport cask, etc. This methodology will make it possible to improve the efficiency of the process of designing the SNF transport casks, avoid possible design errors and, in particular, when used as intended, resolve the issue of the SNF cask loading configuration.

Stress-based vs. Strain-based safety evaluations of spent nuclear fuel transport casks in energy-limited events

The stress-based criteria specified in ASME Section III (American Society of Mechanical Engineers, 2010) have long been used to evaluate the structural safety of mechanical structures, systems, and components (SSC) associated with nuclear power plants. These evaluation criteria, however, have inevitable limitations in accounting for the structural behaviors of spent nuclear fuel (SNF) transport casks, which likely undergo significant plastic deformations in extreme accidents. Hence, the stress-based evaluation criteria are considered inappropriate for energy-limited events that involve a high impact load, such as an accidental drop.This study evaluates the effectiveness of strain-based criteria, which were recently added in ASME Section III, Nonmandatory Appendix FF (American Society of Mechanical Engineers, 2015) and provides critical comparative study over traditional stress-based criteria for assessing the structural safety of SNF transport casks. Horizontal and vertical accidental drop events specified in NUREG-1536 (US Nuclear Regulatory Commission, 1997) are simulated using nonlinear dynamic finite-element analysis (FEA). Four SNF cask models with different impact-limiter designs are numerically simulated. The design of the impact limiter varies with wood types (redwood and balsa wood) and grain directions (transverse and longitudinal with respect to the loading direction). The FEA stress and strain results for the basket and canister parts are examined according to both stress-based and strain-based criteria. The strain-based criteria provide less conservative and more sensitive estimation than the stress-based criteria.

A probabilistic assessment of PWR SNF pinching failure considering hydride-related cladding embrittlement

A probabilistic methodology to investigate pinching failure of pressurized water reactor (PWR) spent nuclear fuel (SNF) rod cladding is presented. The considered accident scenario is a 9-m SNF transportation cask drop after long-term SNF dry storage. The offset strain was used to evaluate cladding failure. A set of 3000 pinching scenarios was generated via Latin Hypercube Sampling (LHS) of geometrical, material, and loading parameters and finite element (FE) models were developed to simulate each scenario. The methodology considers hydride-related cladding embrittlement related to SNF dry storage cool-down. Regression and interpolation models were developed to estimate the cladding offset strain capacity as a function of cladding hydrogen content, peak vacuum drying cladding hoop stress, and cladding temperature at the moment of load application. The probabilistic results indicate a small likelihood of hydride-related pinching failure in the considered accident scenario, after 30 or 300 years of SNF dry storage.

Using Systems Theory to Address Complex Challenges to International Spent Nuclear Fuel Transportation

AbstractSimply looking at a world map suggests new, more complex set of risks and threats will challenge successful international spent nuclear fuel (SNF) transportation operations. Whether related to multimodal transfers of SNF casks or inconsistency in multijurisdictional control measures, international SNF transportation represents a clear example of new threats and risk stemming from a multifaceted, globalized operating environment. In response to recent work out of Sandia National Laboratories (SNL) suggesting that new, system‐theoretic analysis techniques better ensure safe, secure and safeguarded international transportation of SNF, this paper explores the ability of basic systems theory concepts—interdependence, hierarchy and emergence—to better understand and address these complexities in system analysis.

High burn-up spent nuclear fuel transport reliability investigation

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which has provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. These data can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.

Spent Nuclear Fuel Transport: Problem State and Analysis of Modern Approaches

The paper presents the review of international and national experience related to transport of spent nuclear fuel (SNF) and trends in the development of transport containers. The analysis covers the vectors for the future improvement of packaging and the regulatory framework on SNF transport in Ukraine and other countries. The tasks for future research were identified. The results of this research will be used during the operation of the CSNSF.

As-Loaded Criticality Margin Assessment of Dual-Purpose Canisters Using UNF-ST&DARDS

A novel assessment has been completed to determine the previously unquantified and uncredited criticality margin available in as-loaded commercial spent nuclear fuel (SNF) canisters. This assessment was performed as part of a broader effort to assess issues and uncertainties with storage, subsequent transportation, and final disposal of SNF canister systems. Detailed analyses crediting the burnup, initial enrichment, and postirradiation cooling time of actual SNF inventory were performed for 554 SNF canisters stored at 23 commercial reactor sites to determine realistic criticality safety margins. These detailed analyses were automated by the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS), a comprehensive, integrated data and analysis tool. Calculated, uncredited criticality margins (Δkeff) with respect to the safety analysis results range from 0 to almost 0.30 Δkeff for normal storage and transportation cases. Calculated eigenvalues (keff) range from 0.72 to 1.11 assuming a degraded neutron absorber disposal condition, and they range from 0.94 to 1.20 assuming a degraded basket disposal condition. Calculations with NaCl present in the moderator (which is possible for certain disposal geologies) were used to demonstrate the possibility for subcriticality for degraded cases with a keff above 0.98 with freshwater. The methods used to calculate keff for the canisters analyzed in this work are discussed in detail.The results demonstrate that, for the majority of canisters analyzed here, significant uncredited safety margin is available that could be used to compensate for uncertainties in the SNF assembly and canister internal components. These uncertainties are associated with long-term storage and subsequent transportation and disposal. Results also suggest that the inherent margins associated with how canisters are loaded could support future changes in licensing SNF storage and transportation systems to directly or indirectly credit the margins associated with actual SNF characteristics. Ongoing research continues to gather additional data to quantify uncredited safety margins for SNF canisters loaded at other nuclear reactor sites and to explore potential methods for applying this uncredited margin.

Spent nuclear fuel system dynamic stability under normal conditions of transportation

In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. Dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly.To further evaluate the intensity of contact interaction induced by the local contacts’ impact loading at the spacer grid, detailed models of the actual spring and dimples of the spacer grids were created. The impacts between the fuel rod and springs and dimples were simulated with a 20g transient shock load. The associated contact interaction intensities, in terms of reaction forces, were estimated from the finite element analyses (FEA) results. The bending moment estimated from the resultant stress on the clad under 20g transient shock can be used to define the loading in cyclic integrated reversible-bending fatigue tester (CIRFT) vibration testing for the equivalent condition. To estimate the damage potential of the transient shock to the SNF vibration lifetime, drop tests were performed on the CIRFT specimens. FEA was used to investigate the contact reaction at CIRFT test samples during impact loading induced by drop tests, and the result was compared with that from a 20g acceleration transient shock load.

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (keff) Predictions

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. This paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methods and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth within the application model. This paper provides a detailed description of the approach and its technical bases, describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models, and provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data.

Reversal bending fatigue test system for investigating vibration integrity of spent nuclear fuel during transportation

AbstractTransportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading during road or rail shipment. Oak Ridge National Laboratory (ORNL) has been developing testing capabilities that can be used to improve the understanding of the impacts on SNF integrity due to vibration loading, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety and security of SNF storage and transportation operations. The ORNL developed test system can perform reversal bending fatigue testing to evaluate both the static and dynamic mechanical response of SNF rods under simulated loads. The testing apparatus is also designed to meet the challenges of hot cell...

Methodology for Experimental Verification of Safety of Packages for Transport of Spent Nuclear Fuel

Experimental methods of mechanics can be used for verification of lifespan of structures and for assessment their safe operation. For transport of radioactive materials is necessary to use packages that fulfill severe rules declared by laws and regulations. In the paper are given experimental procedures used for verification of safety of packages used for transport of spent nuclear fuel.

Quality control of metal-concrete flasks for spent nuclear fuel

The safety operation of a serial metal-concrete flask intended for spent nuclear fuel (SNF) storage requires flaw inspection of its concrete filler and testing of the radiation protection parameters of the casing for compliance to design specifications. A test bench for γ flaw detection of flasks intended for storage and transportation of SNF in an amount of (5–7) × 103 kg with an activity of ∼106 Ci is described (including its new version). The kinematics of the bench, as well as its electronics, software, and hardware solutions that improve the survivability and reliability of the instrumentation and raise the validity of measurement results, are presented.