Shipping Cask Research Articles

The study on the wood candidates for the impact limiter of a spent fuel cask

The nuclear fuel transportation package is meticulously designed to enclose radioactive materials, guaranteeing containment even after undergoing a free-drop impact test and considering potential accident situations to prevent radioactive leakage. The impact limiters attached to the transport cask are engineered to absorb impact energy. Wood, renowned for its strong energy absorption capacity, is often specified as the core material within impact limiters. It is crucial to accurately determine the wood material employed in the impact limiter. This study carried out crushing experiments on various common wood materials to obtain stress versus strain data for the selection of the appropriate wood material. Indicators such as density, crushing strength, and densification strain are utilized to assess the suitability of wood as the filler material of the impact limiter. Paulownia wood is chosen based on density, the appropriate energy absorption value per unit volume, and response acceleration. The dispersity of the strengths of Paulownia wood is taken into account with the weakest-link models. The finite element method is adopted to precisely design and verify the impact limiter. The arrangement of the glulam wood is designed to balance the energy absorption in hypothetical accident conditions. Finally, tests on the application of the impact limiter in the transportation cask were conducted to validate the proper design of the impact limiter. This study can offer reference and data support for the research on wood filler materials.

Design and dynamic analysis of transport cask for SMR fresh fuel assembly

Transportation of fresh fuel assemblies requires the utilization of transport casks to ensure the internal radioactive material does not affect the environment. In this work, we introduce an innovative design of a transport cask for the fresh fuel assembly of a Small Module Reactor (SMR). The cask is constituted by two containers, supports and various energy-absorbing structures welded at the outer container. The structural integrity of the transport cask under variant scenarios determined by the International Atomic Energy Agency (IAEA) and domestic regulations was numerically evaluated. The detailed 3D model of the cask was created, and the simulations were performed via the Finite Element Method (FEM) using commercial software ABAQUS. Simulation results obtained from static and dynamic analyses indicated that the integrity of the designed transport cask can be assured during normal and accident conditions. The most severe damage during the 9 m drop tests was caused by a horizontal oblique condition and this configuration is recommended to be tested in future experiments.

Concept of operation of the Agile Robot Transporter for ITER in-vessel maintenance

ITER machine availability is crucial for fusion reactors in order to be the main source of electricity, which is affected by many aspects. This paper focuses on maintenance aspects of the machine. As the plasma facing components inside the fusion reactors become highly activated caused by the neutrons generated by fusion reactions, maintenance of the fusion reactors needs to be done by remote means. In order to increase machine availability, it is required to have efficient maintenance scenario in terms of time and effort, which can be achieved through effective in-vessel remote handling (RH) operations, simplified maintenance of the RH system itself, well-coordinated logistics, and trained personnel, etc. All these aspects need to be investigated during the design phases, which requires a well-defined concept of operation of the RH systems. This paper describes the concept of operation of the Agile Robot Transporter (ART), an ITER in-vessel RH system, that is under development to carry out neutron and optical diagnostic calibrations, and removal/replacement of the blanket first wall samples. Based on the functional analysis of the required in-vessel tasks, system requirements are defined. Among them, key requirements to improve RH operation efficiency are identified and the ART is designed accordingly. Efficient deployment and removal operations of the ART are crucial so that it requires less vacuum vessel (VV) port opening and less cask transportation, which reduces needs for hands-on operations resulting in less human radiation exposure. Various and flexible tool exchange and storage concepts are proposed that allow parallel operations between the in-vessel and ex-vessel RH systems. Modular and re-configurable design is proposed for easy RH system maintenance, and effective rescue operations to rescue the RH system when it fails.

Development of simplified fuel assembly model for nuclear fuel transportation

To evaluate the integrity of spent nuclear fuel (SNF) during transportation, the vibration and drop impact analysis requires a transport cask model loaded with a significant amount of SNF and performing a dynamic analysis in the time domain. A cask model with a three-dimensional fuel assembly (FA) model requires substantial time. Therefore, simplified hexahedral-shaped FA models with equal weights have been used. However, FAs interact with the cask and account for a significant portion, affecting the mechanical response behavior of the transport cask and FA. Therefore, a dependable FA model must be used to obtain accurate cask and FA mechanical responses. Herein, we developed a simplified equivalent beam FA model to address the above issues. The guide tubes and fuel rods were modeled as a single equivalent beam, and the spacer grids were simplified to a shell shape by removing springs and dimples supporting the fuel rods. Rotational springs were implemented at the locations of spacer grids and nozzles to simulate the static bending and dynamic vibration of FA. The simplified FA model was compared with mechanical test results, and the results correlated well. Therefore, this model can be used for FA transportation and will be crucial in predicting the reliable mechanical responses of FA and transport cask.

Experimental and numerical investigation of scale model used in the development of a spent nuclear fuel transport cask

To investigate the dynamic response and validate the structural integrity of a spent fuel transportation cask under the 9 m drop condition, a 1/3 scale model was utilized in a 9 m side drop test, considering the constraints of development cost and test facilities. The relationship between the response variables of the scaled model and the prototype model is obtained by Π theorem under impact drop conditions. The influencing parameters for the scale model of spent fuel transport cask drop were systematically discussed. Several numerical models of varying scales are used for verification. Finally, the 9 m drop test of the 1/3 scale model was conducted to compare the acceleration response and deformation of the 1/3 scale numerical model. The study concludes that the acceleration and deformation of the numerical model closely match the test results, and the numerical model of the prototype model represents the dynamic response of the prototype model. Furthermore, the study examined the reflections of the scale model in comparison to the real model, taking into account the dispersity of wood material and strain rate. The analysis and conclusions can be applied in engineering practice and can inspire subsequent cask designs, especially for a scaled model.

Long-Service-Life Rigid Polyurethane Foam Fillings for Spent Fuel Transportation Casks.

Soft materials bearing rigid, lightweight, and vibration-dampening properties offer distinct advantages over traditional wooden and metal-based fillings for spent fuel transport casks, due to their low density, tunable structure, excellent mechanical properties, and ease of processing. In this study, a novel type of rigid polyurethane foam is prepared using a conventional polycondensation reaction between isocyanate and hydroxy groups. Moreover, the density and size of the pores in these foams are precisely controlled through simultaneous gas generation. The as-prepared polyurethane exhibits high thermal stability exceeding 185 °C. Lifetime predictions based on thermal testing indicate that these polyurethane foams could last up to over 60 years, which is double the lifetime of conventional materials of about 30 years. Due to their occlusive structure, the mechanical properties of these polymeric materials meet the design standards for spent fuel transport casks, with maximum compression and tensile stresses of 6.89 and 1.37 MPa, respectively, at a testing temperature of -40 °C. In addition, these polymers exhibit effective flame retardancy; combustion ceased within 2 s after removal of the ignition source. All in all, this study provides a simple strategy for preparing rigid polymeric foams, presenting them as promising prospects for application in spent fuel transport casks.

Radiation shielding analysis and criticality safety calculation for spent fuel transportation cask

China anticipates a shortage in storage capacity for spent fuels at reactor sites for the AP1000 NPPs within the next 10 to 20 years. We proposed a conceptual design for a transportation cask capable of loading 21 spent fuel assemblies in AP1000 reactors and studied its criticality safety and radiation safety characteristics. Throughout these operations, three key concepts were taken into account: source term, radiation shielding, and criticality calculations. The sources of the 4.95% enriched spent assemblies following a cooling period of 21 years were calculated by the cosNAB code. The results of the source term computation were utilized as input data for the Monte Carlo code cosRMC, which was employed for shielding calculation and criticality safety analysis. The calculation results have verified the feasibility of the design, which can provide a reference for the development of spent fuel transportation casks for the AP1000 reactor.

Efficiency of variance-reduction techniques using TRIPOLI-4®: Application to equivalent dose rate calculations in a spent-fuel cask

This paper presents the comparison of different variance-reduction methods implemented in the Monte Carlo code TRIPOLI-4® in the case of a deep penetration problem. In this complex radiation transport situation, where the shielding is very thick to ensure protection, variance-reduction methods are necessary to obtain results with a good accuracy in a reasonable calculation time. The configuration investigated in this work is a spent-fuel transport cask. The various variance-reduction methods used here consist either of Importance Sampling with the Exponential Transform method, or of sophisticated population-control methods. These methods are detailed, then the results are presented and compared and the efficiency of these methods is investigated. While all methods implemented in TRIPOLI-4® are efficient versus analog simulations, the pre-calculation of importance maps with deterministic methods enables a faster convergence of the Monte Carlo simulations, especially when combined with the Exponential Transform.

Radiation effect assessment on spent fuel transport CASK by aircraft sabotage attack

The purpose of this study is to evaluate the radiation effects depending on the size of the damage of cask when the outer shield of a spent nuclear fuel transportation cask is damaged by impact due to sabotage. Neutron and gamma-ray intensities and spectra were calculated using the ORIGEN module of the SCALE 6.2.4 code package (W. A. Wieselquist et al., 2020), and the equivalent dose rates were calculated using the MCNP6.2 code (C. J Werner, et al., 2017).The ABAQUS code was used to assess the degree of the opening of the cask lid due to external impact, and in the case of radiation source leakage, the dose assessment was performed by assuming a maximum of 6% damage to the nuclear fuel assemblies based on overseas experimental cases.

Correlation between charpy-V notch impact energy and KJIC of ferrite ductile iron in the transition-temperature region

Fracture toughness is an important index for the structural integrity assessment of ductile iron spent fuel transportation casks. Fracture toughness tests require a long test period, high cost, and difficult operation. Therefore, the Charpy V-notch impact energy value (KV) is considered to predict fracture toughness (KJIC).In this study, Charpy V-notch impact tests(−80 °C, −60 °C, −40 °C, −20 °C, 0 °C, 20 °C, 40 °C, 60 °C, 80 °C) and fracture toughness tests(−60 °C, −40 °C, −20 °C, 0 °C, 20 °C) of ferrite ductile iron(FDI) were carried out. A KJIC-KV model for FDI in the transition temperature region was proposed, taking into consideration the influence of test temperature(T) and yield strength(σ YS), then the proposed model was compared to the KIC-KV empirical equations of steel in the transition-temperature region, using the test data of FDI. The results indicate that the KJIC-KV model proposed in this work is more accurate and applicable for predicting fracture toughness values in the transition-temperature region than others, the residual between the estimated value of the KJIC-KV correlation model (KJIC-E) and the experimental value of KJIC(KJIC-T) was less than 4.1 MPa·m0.5, and the error was less than 5%.

The study on the main factors affecting the response analysis of nuclear fuel transport cask drop

The nuclear fuel transport cask is an important transport equipment in the nuclear power industry. Rational design requires the structure to maintain integrity and tightness after experiencing severe accident conditions such as a 9 m drop. Therefore, the design and analysis are the key links of container forensics. The calculation of drop analysis involves many factors such as nonlinear material constitutive, complex contact, high strain rate, and large deformation which brings great challenges to the reliability of simulation analysis. Summarizing and sorting out the key factors in the design and analysis of transport casks can solve the practical needs of engineering and guide the analysis and calculation of transport cask drop. This study considered the perspectives of impact limiter design, assembly clearance, analysis model, environmental factors, dropping angle, and the calculation and comparison of buffer materials' dispersion. A two-degree-of-freedom vibration system was used to simulate the impact of clearance collision, and a beam model was applied for the investigation of the response of assemblies. The considerations of the analysis model and the impact of the environment on buffer materials were expounded. Finally, the main conclusions are as follows: the dispersion of buffer materials should be considered in the design. The wood with low compressive strength will cause greater compression deformation, and the wood with high compressive strength will cause greater peak acceleration. The frequency, mass, and clearance of the internal components and the container all affect the acceleration response of the internal structure. The acceleration response is sensitive to the angle under the side and end drop conditions.

Study on application of wood and orthogonal trapezoidal honeycomb aluminum as filling material for nuclear spent-fuel transport casks

The filling material is an important part for nuclear spent-fuel transport casks, it can protect the inner structure of nuclear spent-fuel transport casks. Balsa wood and paulownia wood are common materials used as filling materials for nuclear equipment. However, wood has the disadvantages of grain and intolerance to humidity and heat. In this paper, the mechanical property of traditional filling materials -paulownia wood and balsa wood were studied. In addition, two new types of porous materials of 90° orthogonal trapezoidal honeycomb aluminum (90-OTHA) and 45° orthogonal trapezoidal honeycomb aluminum (45-OTHA) were designed. Quasi-static compression experiments (Y-direction) were used to test the mechanical property of these four materials. The experimental results showed that paulownia wood and balsa wood were better in the axial direction than in the radial direction. The mechanical properties of 45-OTHA were better than 90-OTHA. The peak stress of 45-OTHA was 1.25 times and 1.21 times and the average stress was 1.62 times and 1.32 times than that of axial paulownia wood and axial balsa wood, respectively. The total energy absorption of 45-OTHA was 1.9 times and 1.75 times than that of axial paulownia wood and axial balsa wood. It was concluded that OTHA has better mechanical properties than wood. Finally, by comparing energy absorption indicates that OTHA could be invoked as a filling material for nuclear-spent fuel transport casks to replace wood.

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.

Calculation and experimental studies for the spent nuclear fuel shipping cask sealing assembly

One of the safety requirements regarding the shipping cask for spent nuclear fuel is that its leak-tightness should be maintained by preserving the cask body structural integrity and the sealing system tightness under normal and accident transportation conditions. The cask under design has a cylindrical process penetration (port) in its bottomб which is sealed using a plug with a radial seal composed of two rubber O-rings. The cask sealing assembly design was justified by the ANSYS LS-DYNA code calculation results. In particular, the strains of the cask components were calculated when dropped from a height of 1 m with the sealing assembly hitting a vertical bar. The cask was concluded to be leak-tight or leaky based on the strain nature and amount. To verify the adequacy of the results, computer-aided and realistic simulations were undertaken with a 1/2.5 scale mockup cask dropped on a bar from a height of 1 m. The computational and experimental results show a good agreement in terms of the impact response accelerations (overloads) for the mockup cask and bar collision and in terms of the plastic strains for the key components of the mockup bottom port sealing assembly. This proves the adequacy of the numerical cask model that has been developed and the efficiency of the LS-DYNA simulations. The inner rubber O ring compression is reduced by the plastic strains in the cask’s bottom port area, leading to a loose inner radial seal, as shown by the calculations. But the outer seal remains leak-tight, ensuring so the mockup cask tightness. The physical test results have also confirmed that the mockup cask remains leak-tight.

Investigations of aged metal seals for transport package safety assessment

Abstract. Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM – the German competent authority for mechanical, thermal and containment assessment of packages liable for approval – verifies the activity release compliance with the regulatory limits. One of the fundamental aspects in assessment is the specification of conservative package design leakage rates. To ensure the required package tightness for both storage and transport of the cask before and after storage, usually metal seals of the HELICOFLEX® type are used. Due to long-term use, the seal behaviour is influenced by temperature and time. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non-aged HELICOFLEX® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged HELICOFLEX® seals. BAM performed experiments to learn more about the sealing efficiency of aged HELICOFLEX® seals with aluminium and silver outer jackets. The seals were compressed in test flanges, and for artificial ageing the complete flange systems were stored in an oven for several months at a high temperature. During the compression and decompression tests after the ageing, load-deformation characteristics of the seals and leakage rates were measured. With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: under high-impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal so that the contact area can change before compressing again. The poster presentation will show details about test conditions and first results.

Simulation of Spent Fuel Assembly Transport Vertical Drop Based on LS-DYNA

The finite element technology is used to carry out the drop simulation and safety assessment of the spent fuel assembly and its transport cask. In view of the problem of how to reasonably simplify the finite element model, a more reasonable and accurate multistage simplified model is proposed for the spent fuel assemblies packed densely in the transport cask, which needs to be analyzed synchronously when dropping. First, the “key” spent fuel assembly is retained, and the “unimportant” assembly is simplified as a cylindrical entity. Based on this, the structure will continue to be simplified and a 1/4 near-earth model containing all characteristics will be established. Then, the LS-DYNA software is used for explicit dynamic analysis of vertical drop conditions. It is concluded that the failure risk of the spacers is the greatest. Therefore, the purpose of this paper is to propose a multistage simplified method for spent fuel assembly and its transport cask, which can provide a reference for the structural integrity analysis of the spent fuel assembly transport drop conditions.

Innovation in lifetime prediction of rigid polyurethane foam through random vibration and comparison with conventional methods

As the filling material for the new nuclear fuel transport cask, the rigid polyurethane foam plays an important role in the safe transportation of the nuclear fuel assemblies, and therefore its reliability deserves special attention. Thermal analysis and long-term ageing test are commonly used to predict the service lifetime of rigid polyurethane foams, but they cannot fully consider the influences of mechanical loads like vibration, collision, and impact etc., which is not satisfactory for the transportation application. Herein, in addition to the conventional thermal analysis and long-term ageing methods, the random vibration testing was particularly conducted on the rigid polyurethane foam that has been used in the new nuclear fuel transport cask for the nuclear power plants in China, and its service lifetime was innovatively predicted by the intrinsic property of material, the resonance frequency. The result which can be easily obtained from the acceleration data through fast Fourier transform without additional test and treatment was in the same order of magnitude with that from the conventional methods, and is more rational and convenient, deserving recommended for the rigid polyurethane foams that are used in other applications involving mechanical loads.

Decay Dose Shielding Analysis with Hybrid Unstructured Mesh/Constructive Solid Geometry Monte Carlo Calculation and ADVANTG Acceleration

The target segments of the Oak Ridge National Laboratory Second Target Station (STS) neutron production facility become highly activated due to spallation reactions or nuclei transmutation by primary protons and emitted neutrons. Once the target segments are removed from their location within the core vessel, decay dose rates must be accurately quantified to determine the shielding configurations of remote-handling tools and transport casks and to aid in planning maintenance activities. For this analysis, we utilized a hybrid unstructured mesh (UM)/constructive solid geometry approach for calculating spallation products and neutron fluxes, activation calculations using the AARE package that includes the CINDER2008 activation code to calculate the decay photon source at different cooling times, and the ADVANTG code to accelerate the final decay photon transport calculation. Both Type 316 stainless steel (SS-316) and lead were investigated as candidates for shielding materials. The decay photon transport calculation through the thick SS-316 or lead shields exhibited between 25 and 30 orders-of-magnitude attenuations in the radial direction, depending on the shield. Such a difficult shielding calculation required advanced variance reduction. ADVANTG has some missing features, which limits its usability in spallation neutron source applications. It does not support volumetric sources created for MCNP6.2 UM capability. An approximate source was created for this problem. Not only was this approximate source needed for running the ADVANTG calculation to generate the weight windows, but also it was essential to develop source biasing (SB) parameters that were crucial for dramatically accelerating the decay photon transport in this problem. With this approximate source, the analysis was completed in a very reasonable computational time, and the design of the STS remote-handling equipment was finalized. This paper compares the efficiency of Monte Carlo simulations with different weight window and SB parameters calculated using different approximate ADVANTG calculations.

Friction coefficients for wood-wood and wood-steel high pressure contact under temperatures between -40 °C and 90 °C

Wood is widely used in impact limiters of transport casks for radioactive material. Encapsulated by an outer and inner steel structure, spruce wood is often applied in layers of alternating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard component. In order to get detailed information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40 °C and 90 °C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40 °C to 0.22 for 90 °C for wood-steel combinations and from 0.3 at -40 °C to 0.24 at 90 °C to for a wood-wood combination.

Using modern virtual reality techniques to perform analysis of ITER ECH EL port cell maintenance

In order to maintain an ITER Port Plug after failure, the Port Cell needs to be cleared to allow for a transport cask to access the Port Plug and transfer it for maintenance to the Hot Cell Complex. The Port Cell clearing process is a complex, manual and machine assisted procedure, taking out the equipment blocking cask access in several stages.For the Equatorial Launcher (EL) of the Electron Cyclotron Heating (ECH) system, which is located in Equatorial Port Cell #14 of the ITER tokamak, the Port Cell clearing operation includes the removal of 24 Ex-vessel Waveguides, a complex set of CCWS (component cooling) and PHTS (primary cooling) lines, several manifolds, and other ancillaries, like the Service Vacuum System (SVS) to monitor the torus vacuum tightness.This paper shows how modern Virtual Reality (VR) simulation techniques, in combination with a state of the art, high-end game-PC environment and Head Mounted Display, can be used as a tool to visualize and analyse the maintenance procedures, to predict the time to repair, and to assist in minimizing radiation exposure to take in to consideration the ALARA principles.Special emphasis is put on the verification and validation of manual and machine-assisted maintenance procedures and the identification of tooling requirements. The possibility to simulate logistics and repair actions in an early stage of the design process allows for the identification of those maintenance actions that require dedicated tests or the development of dedicated tools.The added value of using modern VR techniques to analyse the maintenance scenarios are:•Immersive & intuitive – very little training required to experience the process first-hand•Quick and accurate assessment of maintenance scenarios, based on the actual CATIA models•The ability to assess in detail geometrically very complex maintenance sequencing•Less need for expensive hardware mock ups