Dry Cask Research Articles

Dry cask radiation shielding validation and estimation of cask surface dose rate with MAVRIC during long-term storage

The dry storage cask became a semi-permanent solution for spent nuclear fuel since no permanent solution has been finalized over the past three decades. Considering the difficulty of deep geological repository projects in many aspects, prolonging the service life of dry casks longer than a few decades has become an option for spent nuclear fuel storage. It must be ensured that dry cask safety objectives, including radiation shielding, criticality safety, structure health, decay heat removal, and containment be fulfilled during the prolonged service life. This paper presents numerical modeling results of the radiation dose rate distribution on a TN-32 cask surface aged up to 300 years. The numerical modeling method was validated with measurement data, and the difference between the calculated results and measurement data in the total dose rate is 7%. The detailed radiation distribution variations over cask surfaces along the axial and radial directions for 300 years of storage were obtained, and their azimuthal angular dependence was established. The maximum total dose rate over the cask surfaces is less than 10 mrem/hour (9.258 ± 0.212 mrem/hour) after 40 years, and this outcome fulfills all dose rate requirements from 10 CFR 71.47(b). After 300 years of storage, the average total dose rates on the side surface and top surface are 0.150 ± 0.007mremmrem/hour and 0.057 ± 0.003mremmrem/hour, respectively, while the maxima are 0.601 ± 0.014 mrem/hour and 0.112 ± 0.006 mrem/hour, respectively. The variations in the fuel gamma ray, structure gamma ray, neutron and secondary gamma-ray contributions to the total dose rate are presented and analyzed in this work. The neutron and gamma-ray dose rate contributions by energy group are presented and analyzed at the cask top, mid-plane, and bottom at the selected storage times.

Verification of dry storage cask loading using monoenergetic photon sources

Safeguarding the contents of spent nuclear fuel dry storage casks requires the verification of cask contents, especially if continuity of knowledge has been lost. The high attenuation of the cask’s thick shielding and stored fuel assemblies makes verification difficult using conventional photon radiography. Thomson-scattering quasi-monoenergetic photon sources offer narrow angular and energy spread for increased penetration. This paper demonstrates the advantages of such a source by examining effects of source properties on transmission through thick, highly attenuating objects. This information is then applied to calculations and simulations of transmission for a variety of photon sources and cask configurations. The performance of the source is compared to traditional bremsstrahlung sources and other advanced methods, in the context of typical nuclear fuel dry storage casks. It is shown that the use of a Thomson scattering based quasi-monoenergetic photon source enables determination of fuel assembly occupancies faster and more accurately than other methods.

A multi-scale review of the effects of gamma radiation on concrete

For concrete used in dry cask nuclear storage, gamma radiation on the order of MGy is an important weathering mechanism. At such doses, previous studies have observed microstructural effects such as changes in carbonate speciation, which have correlated with a decrease in bulk porosity. However, the mass, dimensions, and mechanical properties of concrete have not been substantially affected, except in cases of radiogenic heating or extremely high radiation doses (≥1 ​GGy). Based on a review of the literature, a multi-scale mechanistic framework is presented to describe and model the effects of gamma radiation on concrete, including the interactions of photons with the various cementitious phases, aggregates, additives, admixtures, and forms of water. The durability of concrete to radiation is not apparently affected by the type of cement or the use of secondary cementitious materials. However, the use of Si-containing aggregates or of certain organic additives or admixtures (notably fibers or latex) may be detrimental to the durability of concrete in nuclear applications.

Feasibility assessment of using external neutron and gamma radiation measurements for monitoring the state of fuel assemblies in dry storage casks

This study assesses the feasibility of using external neutron and gamma flux measurements to detect the state of spent fuel assemblies inside a sealed cask. Several Monte-Carlo simulations were performed to evaluate the neutron and 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 positions of the damaged fuel assembly and different storage times. The results show that in case the damaged fuel assembly is located near the inner wall of the cask, it can be detected by both of the radiation modes. The detection of the inner damaged fuel assemblies, however, can be feasible only by the external neutron flux.

Vibrational modes for the internal characterization of a full-scale Transnuclear-32 dry storage cask for spent nuclear fuel assemblies

The assessment of the internal structural integrity of dry storage casks with high burnup nuclear fuel assemblies is of critical importance for their transport to permanent repositories and during their extended storage service lives. The large size, structural complexity, and inability to access the interiors of the casks make this task challenging. The aim of this work is to evaluate the ability of vibrational characterization to assess the state of the internal cargo (i.e., fuel assemblies) in a cask strictly using measurements on the outer surface. In this study, we report on the identification of five modes of an unloaded full-scale Transnuclear-32 (TN-32) cask and examine the potential of using these modes to assess the internal configuration in a loaded condition. Vibrational spectra were acquired using impulse and continuous-wave techniques on both storage and transportation configurations from which the modal frequencies and the associated quality factors were determined. A finite element model of the TN-32 was constructed and used to identify the observed modes. The model was extended to include loaded configurations to assess how the presence of assemblies impacts the modal structure of the cask and its surface vibrations.

MOX Fuel Alternatives for Minor Actinides Burnup in Thermal Spectrum

The spent fuel from the nuclear power reactors is being stored in repositories and dry casks in many countries of the world. However, it does not solve the main problem of the spent fuel, which is the high radiotoxicity of fuel for a long term. A new strategy is required to close the nuclear fuel cycle, and for the sustainability of the nuclear power generation. This strategy could be the plutonium recycling to obtain more energy and recycle the Minor actinides produced during the irradiation, to transmute them into less radioactive products.

Research on Shielding Performance of Dry Storage Cask for Spent Fuels

The radiation shielding design of spent fuel storage containers is one of the important factors affecting the radiation safety of spent fuel dry storage facilities. Taking the dry storage container independently researched and developed by China as an example, the paper focuses on the reasonable determination of the shielding performance target, the shielding calculation method, the selection of the calculation hypothesis and the analysis of the calculation results. The calculation analysis shows that the designed spent fuel dry storage container is with good shielding performance and meets the radiation safety requirements.

Destructive and non-destructive evaluation of reinforced concrete dry casks affected by alkali-silica reactivity damage

This article investigates the effect of alkali–silica reaction (ASR) damage on the structural performance of reinforced concrete (RC) dry casks for spent nuclear fuel (SNF) storage. Dry casks have an initial licensed lifetime of 20 or 40 years with the possibility of an extension of another 40 years. Due to lack of a permanent storage facility in the United States, an extension of the service life of dry casks that are currently in use and that were designed for shorter service lives is a critical step. Deterioration of concrete due to ASR is a major concern as has been reported for some nuclear facilities. In this study, the effect of ASR aging on the tip-over response of a vertical RC cask is experimentally studied for the first time using 1/3-scale physical models. An accelerated aging method was used to represent the long-term degradation of the structure. A series of non-destructive tests were performed to monitor the aging progress. At the end of the aging period, the structural performance of the casks is evaluated experimentally under tip-over impact. The results are compared with data obtained from a tip-over test on an undamaged cask. In addition, a detailed finite element model of the cask is developed to simulate the tip-over impact in aged conditions.

Surrogate modelling to enable structural assessment of collision between vertical concrete dry casks

Vertical concrete dry casks, used for interim storage of spent nuclear fuel, are susceptible to large horizontal displacements caused by lateral loads. This study explores the collision between adjacent casks in the case of such large motions. To estimate an upper bound for the probability of failure, a common collision scenario that causes severe structural damage is determined through a comparative deterministic analysis. Then the effects of uncertain parameters on the cask’s performance in the specific collision scenario are studied with a probabilistic approach. Numerical simulations are conducted for 200 realisations of the collision problem to evaluate the canister maximum strain and the overpack accelerations. Then surrogate models, such as polynomial response surface models, multivariate adaptive regression splines, regression trees, and support vector machines, are trained on the response data to derive efficient approximating functions for these responses. Utilising the best developed surrogate models, fragility and the failure probability given collision are estimated. The findings show that the structural integrity of the dry cask is generally maintained in case of the collision. That is, the canister might yield due to the impact loads, but the probability of the canister fracture is negligible for the cask and parameter space considered here.

Methodology for thermal analysis of spent nuclear fuel dry cask using CFD codes

As the capacity in spent fuel pools is depleted, dry storage systems have been increasingly used. The key parameter for fuel cladding behaviour during dry storage is the temperature; that cannot surpass 400 °C based on NRC regulation. This paper proposes a new methodology to realistically calculate the thermal behaviour of cask/fuel system.In order to obtain realistic values, the methodology is divided in two steps. In the first step the outside part of the cask is simulated (from the inner surface of the interior cavity to the surrounding air). The second step is a fully fledge simulation of the inner cask (from the clad of the fuel assemblies to the exterior surface of the cask). This two-step approach shows some advantages compared to previous methodologies: the first step calculates a realistic boundary condition to apply in the second step and estimates the impact that the cask has on the near environment while the second step provides the temperature distribution for every single clad in each assembly.To ensure the validity of the methodology, several sensitivity analyses have been conducted, a grid independence study to the discretization error, a turbulence modelling has been applied to the outside and inner model and finally a study of the gaps between the canister and the inner cylinder has been carried out.Of these studies the gaps proved the most important one with changes in PCT of up to 28 °C. The changes in the size of the gaps not only change the overall temperature distribution but also the heat transfer mechanisms and the flow paths followed by the helium.

Validation of the GAMMA-PC methodology for dry cask loading optimization

This paper evaluates the performance of a new greedy multiobjective memetic algorithm, GAMMA-PC, that was developed to find dry cask loading configurations with fewer casks, lower heat loads, and earlier transportation schedules. GAMMA-PC was applied to the dry cask loading problem for Zion Nuclear Power Station. This site represents a unique opportunity to validate algorithmic performance as its fuel has been completely transferred to dry storage using a high-quality loading strategy. The optimization of Zion was performed here using the same transfer timeframe as the vendor in charge of decommissioning and using an extended ten-year transfer timeline. The results of these scenarios showed that GAMMA-PC produced comparable solutions to the real Zion strategy. In the first scenario, GAMMA-PC produced a solution that dominated the real loading configuration, achieving a lower average cask initial heat load. Under the extended timeline, the real loading configuration dominated the GAMMA-PC solution, which had a higher average initial heat load. The values for the other objectives showed no difference in quality. The analysis of the individual cask characteristics also highlighted the fact that the dry cask loading paradigm does not directly address balance among the casks. Future work might incorporate this as an objective. This validation has shown that while the algorithm and problem objectives have some room for improvement, GAMMA-PC performs well and is a promising tool for the optimization of the dry cask loading problem.

Heat transfer enhancement in dry cask storage for nuclear spent fuel using additive high density inert gas

Dry cask storage systems (DCSS) is a method of storing high-level radioactive nuclear spent fuel. Due to the decay heat from fission products, effective cooling of the spent fuel is one of the key roles of dry cask storage systems. This study proposes a method to improve the cooling performance of DCSS by adding a small amount of high-density inert gas into helium backfill gas and verifies the method by numerical analysis. First, the candidate group of additive gases that can improve heat transfer in the DCSS environment is selected through figure-of-merit (FOM) analysis based on natural convection heat transfer models. Then, the candidate gases are evaluated using detailed computational fluid dynamics (CFD) modeling without the porous media assumption. From the analysis, it is found that Xe and Kr can reduce the peak cladding temperature (PCT) inside the cask at pressure greater than 1.5 atm. The optimal composition of the additive gas (Xe and Kr) is estimated to be approximately 0.8 in mass fraction. The effect of the additive gases becomes more significant as the pressure increases due to the effect of increased density.

Thermal design optimization of passive cooling capability in a dry-storage system by adding wall undulations or semi-circular fins

Thermal design investigation of a novel liner inner wall surface geometry alteration of a ventilated concrete cask used for spent nuclear fuel storage is carried out through three-dimensional Computational Fluid Dynamics (CFD) simulations. The present study reveals that the new proposal returns values for the convective heat transfer coefficient that are almost three orders of magnitude higher than the base case. Hence, proving that even making such a rather simple alterations to the smooth walls geometry, i.e. adding undulations or semi-circular fins, can result in a significant convective heat transfer enhancement, therefore, further reducing the risks of structural damages caused by the dry-storage system overheating. Before conducting the design optimization study mentioned above, two validation test cases have been considered in order to examine the Reynolds Averaged Navier-Stokes (RANS) models’ capability to predict the flow under buoyancy driven and low turbulence levels such is the case in the passively cooled dry cask system. The test cases considered in the validation study sections are the experimental measurements of Betts and Bokhari for turbulent natural convection in an enclosed tall cavity and the temperature measurements in concrete-shielded spent fuel storage cask system reported by McKinnon. The models selected for the validation study are namely the k-ω SST model, the v2-f model, the low-Reynolds k-ε model, and the recently formulated elliptic blending k-ε model. It has been found that out of these models, the low-Reynolds k-ε model was found to provide a good combination in returning both; physically realistic and conservative solution. The present validation using well-resolved RANS simulations has shown that a more quantitative and detailed comparison of the predictions is required in order to clearly highlight the discrepancies between the models’ predictions in terms of velocity, turbulent kinetic energy and, in particular, convective heat transfer coefficient profiles. The commonly adopted methodology in limiting the validation process to temperature profiles comparison only, would cover up these differences as the temperature variation is affected by the combined effects of the convective and the radiation heat transfer modes.

Implementation of Hydrogen Solid Solubility Data and Precipitation Threshold Stresses in the Fuel Rod Code TESPA-ROD

During operation of light water reactors, the Zircaloy fuel rod cladding is susceptible for hydrogen uptake. When the local solubility limit of hydrogen in Zircaloy is reached, additional hydrogen precipitates as zirconium hydride, which affects the ductility of the fuel rod cladding. Especially, the radially aligned hydrides enhance embrittlement, while circumferential (azimuthal) hydrides have a less detrimental effect. In this work, the influence of high temperatures during the dry storage period on hydride dissolution and precipitation is demonstrated. Therefore, in a descriptive example scenario being discussed, the simulation of a limited heat removal from the cask will heat up the dry storage cask for days and causes dissolution of hydrides in the cladding. Depending on the threshold stress for reorientation, the following cooldown results on different hydride precipitation behavior. The threshold stress leads to an enhanced or delayed precipitation of radial hydrides. The GRS fuel rod code TESPA-ROD is equipped with a new model for hydrogen solubility and applied to long-term storage transients. In this article, hydride refers to zirconium hydrides formed inside the fuel rod cladding.

Development of vibrational metrics for internal damage scenarios of a scaled Transnuclear-32 dry storage cask for spent nuclear fuel

The assessment of the internal structural integrity of dry storage casks for used high burnup nuclear fuel assemblies is of critical importance before transporting these to permanent repositories. The large size and structural complexity of the Transnuclear-32 (TN-32) cask as well as the inability to access its interior make this a challenging task. To address these difficulties, we use an active acoustics approach to develop metrics that are sensitive to the internal configuration of these casks. A 6:1 scaled model of the TN-32 cask was constructed in order to study the internal configuration of the fuel assemblies including various damage scenarios. Each mock-up fuel assembly consists of bundled steel rods, and their structural failure is mimicked by steel shot of equal weight. This talk will report the amplitude- and phase-based active acoustics metrics that we developed to characterize different levels of internal damage. Our studies indicate that vibrometric signatures of various internal conditions can be measured using sources and sensors mounted on the exterior shell. Our current methodology is sensitive enough to detect structural failures at the single fuel assembly level. [Work supported by DOE NEUP Award No. DENE0008400.]

A Methodology for Optimizing the Management of Spent Fuel of Nuclear Power Plants Using Dry Storage Casks

The management of spent nuclear fuel assemblies of nuclear reactors is a priority subject among member states of the International Atomic Energy Agency. For the majority of these countries, the destination of such fuel assemblies is a decision that is yet to be made and the “wait-and-see” policy is thus adopted by them. In this case, the irradiated fuel is stored in on-site spent fuel pools until the power plant is decommissioned or, when there is no more racking space in the pool, they are stored in intermediate storage facilities, which can be another pool or dry storage systems, until the final decision is made. The objective of this study is to propose a methodology that, using optimization algorithms, determines the ideal time for removal of the fuel assemblies from the spent fuel pool and to place them into dry casks for intermediate storage. In this scenario, the methodology allows for the optimal dimensioning of the designed spent fuel pools and the casks’ characteristics, thus reducing the final costs for purchasing new Nuclear Power Plants (NPP), as the size and safety features of the pool could be reduced and dry casks, that would be needed anyway after the decommissioning of the plant, could be purchased with optimal costs. To demonstrate the steps involved in the proposed methodology, an example is given, one which uses the Monte Carlo N-Particle code (MCNP) to calculate the shielding requirements for a simplified model of a concrete dry cask. From the given example, it is possible to see that, using real-life data, the proposed methodology can become a valuable tool to help making nuclear energy a more attractive choice costwise.

Probabilistic analysis of vertical concrete dry casks subjected to tip-over and aging effects

Vertical concrete dry casks, which provide an interim solution for the storage of spent nuclear fuel, are vulnerable to tip-over due to lateral loads. This paper explores the probabilistic structural response of a representative dry cask subjected to tip-over impact loads and aging effects. Considering concrete and soil material properties, thickness of the pad, angular velocity at the onset of tip-over, and age of the cask as variables, 200 configurations of the tip-over problem are generated by Latin Hypercube Sampling. To consider aging and the associated temporal variations, material properties of the steel parts vary in the generated configurations due to temperature variations. The generated tip-over scenarios are analyzed by numerical models developed and validated in a commercial finite element analysis program. Machine learning algorithms such as polynomial response surface models, multivariate adaptive regression splines, regression trees, support vector machines, Gaussian processes and neural networks are trained on datasets provided by the finite element analysis to develop metamodels for the maximum strain of the canister and maximum acceleration of the concrete overpack. These metamodels offer statistical approximating functions that enable efficient impact fragility analysis and evaluation of the sensitivity of the fragility to age, concrete material properties of the cask and pad, and pad thickness. The effect of aging due to alkali-silica reaction on the strain fragility is also studied, and recommendations regarding the design of vertical concrete dry casks are provided. Risk analysis is performed to estimate the dry storage cask’s risk of failure (i.e., probability of limit state exceedance) and to uncover the effect of age and design parameters on the probability of strain and acceleration exceeding predefined limits. The risk analysis shows that the probability of canister failure in the tip-over event is negligible. That is, despite expected cracks in the concrete overpack, the structural integrity of the dry cask system subjected to the tip-over impact loads is retained.

Regional Cooperation for Nuclear Spent Fuel Management in East Asia: Costs, Benefits and Challenges – Part II

ABSTRACT East Asia is home to some of the most dynamic economies on earth, but also a locus of current and historical national and international conflicts. Some of the largest economies lack domestic energy resources, and nuclear power has been adopted as a perceived key to energy security. Lacking, however, is a concerted strategy for managing nuclear spent fuel at the regional or even, for the most part, national levels. Regional coopera tion on spent fuel management offers many benefits, as well as challenges. This two-part paper explores four different “scenarios” of regional spent-fuel management, ranging from a national “go-it-alone” option to regional cooperation on uranium supply, enrichment and “back-end” spent fuel management. Results for physical flows of nuclear materials and other inputs and outputs, and for the relative costs of each scenario, are presented. Results suggest that the costs of scenarios that include reprocessing are higher than those without reprocessing, and costs for increased dry-cask storage (reducing the amount of fuel stored in high-density spent fuel pools) are likely to be a tiny part of overall nuclear fuel cycle costs. As a result, radiological risk and attendant political, social and legal concerns should drive decisions regarding spent fuel management, not costs.

Investigation of Thermal Hydraulic Behavior of Dry Spent Fuel Cask Under Forced Convection

The thermal hydraulic behavior of the spent fuel in dry storage cask under forced convection mode is experimentally and numerically investigated. For this purpose a test rig is designed and constricted to simulate the cooling loop cask contains 21 spent fuel discharged from pressurized water reactor (PWR). A numerical simulation is performed by ANSYS-CFX fluid dynamic code. The effect of decay heat generation and inlet air velocity are investigated. The results show that the increase in the inlet air velocity improves the coolability of the fuel while the increase in decay heat leads to decrease the coolability of the fuel. Within the range (1.1< V < 2.8 m/s) for inlet air velocity and heaters power (630 < Q < 1260 watt), a new empirical correlation has been obtained for Nusselt number, Nu as a function in Reynolds number, Re. The comparisons between experimental and numerical results show a good agreement.

Investigations on potential methods for the long-term monitoring of the state of fuel elements in dry storage casks

Abstract Extended dry storage of spent nuclear fuel is a relevant issue in many countries operating nuclear power plants. Beside regulatory and security aspects there are questions with respect to the long-term integrity of the spent fuel as this is of relevance for final transportation and reloading to final waste repository casks. Within the frame of the BMWi project DCS-MONITOR, we investigate the potentials of different methods for non-intrusive monitoring of dry cask storage containers with spent nuclear fuel. These are thermography, radiation-based methods, and acoustic methods. For all of them we study the sensitivity and cross-sensitivity with respect to defined changes in the nuclear fuel distribution inside the containers. The analyses are mainly based on numerical simulations but also include some dedicated experimental studies.