Spent Fuel Transportation Cask Research Articles

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.

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.

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.

Design and Application of Impact Limiter for a Spent Fuel Transport Cask

Design and Application of Impact Limiter for a Spent Fuel Transport Cask

Design and Application of Impact Limiter for a Spent Fuel Transport Cask

Design and Application of Impact Limiter for a Spent Fuel Transport Cask

A Study on Radiation Safety Evaluation for Spent Fuel Transportation Cask

A Study on Radiation Safety Evaluation for Spent Fuel Transportation Cask

Spent fuel transportation cask under accidental fire conditions: Numerical analysis of gas transport phenomena affecting heat transfer in shielding materials

In the nuclear industry, spent fuel is usually conveyed to the reprocessing or long term storage facilities in dedicated containers: the transportation casks. These structures are designed to meet rigorous standards intended to prevent any failure under both normal transport and severe accidental conditions. The design process includes thermal analyses in order to predict the temperature response of the cask to a 30 min engulfing fire at 800 °C. Materials presenting endothermic decomposition reactions are often used as thermal and radiologic protection in the cask walls. This paper focuses on numerically investigating the secondary effects induced by the vapor transfer phenomena specific to these endothermic shielding materials, using finite element modeling of heat transfer within the wall of a typical cask. Mechanisms such as vaporization, diffusion and recondensation of the water content, resulting in a transfer of latent heat within the medium, are taken into account in the model. Three materials are considered: polyester resin compound, plaster and phenolic foam. The predictions of this detailed model are compared to those of a model assuming only conduction heat transfer. It is shown that gas transport phenomena have a significant effect on the heating kinetics. Finally, the influence of model parameters such as the material porosity and the condensation coefficient is discussed.

Study on aluminum foam as a filler material for impact limiter of spent fuel transport cask

Spent fuel transport cask is a significant carrier of spent fuel transport. The main function of impact limiters installed at both ends of the container is to absorb energy and limit overload to ensure the integrity of the structure. The quasi-static compression process of aluminum foam was simulated on the platform of ANSYS Workbench. Foam aluminum was prepared by melt foaming method and quasi static compression test was carried out. The experimental results show that the deformation process of aluminum foam is basically the same as that of experiment, and the aluminum foam has good compressive and energy absorption properties. The yield stress (σys) and plateau stress (σpl) of aluminum foam with density of 0.64 g/cm3 can reach 8.26 MPa and 11.11 MPa respectively, and the energy absorption capacity (WEA) and unit energy absorption capacity (WSEA) can reach 6.31 x 103KJ/m3 and 9.87 KJ/Kg respectively, and the difference between the foam with density of 0.61g/cm3 and its various properties is very small. It can be concluded that the aluminum foam in a certain density range has roughly the same performance, and it also reflected the stability of aluminum foam's performance. Additionally, aluminum foam is an isotropic material, which can overcome directional limitation when used as shock absorber filler material for spent fuel transport cask.

Application of Dose Evaluation of the MCNP Code for the Spent Fuel Transport Cask

Application of Dose Evaluation of the MCNP Code for the Spent Fuel Transport Cask

Monte Carlo Shielding Calculations for a Spent Fuel Transport Cask with Automated Monte Carlo Variance Reduction

For the purpose of performing reasonable shielding calculation of a spent fuel transport cask, the use of Monte Carlo methods has been proposed for solving the radiation transport problem on a detailed structure of the transport cask considering fixed neutron sources. A SMIRE (Simplified MCNP-ANISN_W Variance Reduction) system has been developed in the present study, which is possible to generate automatically the lower weight boundary of the weight window for each mesh based on the Consistent Adjoint Driven Importance Sampling (CADIS). Compared with the case of the importance based on the empirical formula, the figure of merit is increased by a factor of 25. In this system, it is possible to calculate the weight suitable for the distantly-positioned detector point from the fuel effective region to introduce the relaxation factor which relaxes the increase of the particle numbers at the boundary of weight window meshes that are generally caused by the large attenuation of adjoint flux. This system is used for a variety of the radiation transport problems as well as the transport cask.

Spent fuel transport cask thermal evaluation under normal and accident conditions

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the ‘80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are discussed.

Performance of the improved version of Monte Carlo code A3MCNP for large-scale shielding problems

A3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A3MCNP uses the three-dimensional (3-D) Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A3MCNP (referred to as A3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem.

Improvement of Monte Carlo Code A3MCNP for Large-Scale Shielding Problems

A3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic “importance” (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A3MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A3MCNP (referred to as A3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A3MCNPV for a concrete cask streaming problem and a PWR dosimetry problem.

Development of Fission Source Acceleration Method for Slow Convergence in Criticality Analyses by Using Matrix Eigenvector Applicable to Spent Fuel Transport Cask with Axial Burnup Profile

Effective fission source acceleration method is studied in criticality safety analysis for a realistic spent fuel transport cask. Various axial burnup profiles from almost symmetry to strong asymmetry based on in-core flux measurements are proposed in the OECD/NEA Phase II-C burnup credit benchmark problem. In some cases, calculations by ordinary Monte Carlo method show very slow convergence of fission source distribution, and unacceptably large skipped cycles are needed to obtain a reliable fission source distribution for statistic criticality estimation. The matrix eigenvector calculation developed and incorporated in the ordinary Monte Carlo calculation to accelerate the slow fission source convergence is applied to the benchmark. The efficiency of acceleration by the matrix eigenvector calculation depends on the precision of fission matrix elements. In a certain stage of source iteration with acceleration repetition of fission source distribution, especially for this benchmark problem of very slow convergence, more acceleration repetitions cause anomalous results because of large statistic fluctuations of the estimation of fission matrix elements for regions with very low source levels. Here, we propose a new source acceleration method to detail with the slow convergence with less calculation time by modeling the division of fissile fuel region.

Development of Fission Source Acceleration Method for Slow Convergence in Criticality Analyses by Using Matrix Ligenvector Applicable to Spent Fuel Transport Cask with Axial Burnup Profile

Development of Fission Source Acceleration Method for Slow Convergence in Criticality Analyses by Using Matrix Ligenvector Applicable to Spent Fuel Transport Cask with Axial Burnup Profile

Development of NFT Type Spent Fuel Transport Cask

NFf type casks have been developed to transport spent fuel from domestic nuclear power plants to the reprocessing plant which is under construction in Rokkasho-mura, Aomori Prefecture. The spent fuel to be transported is greater in burn up and enrichment than that for the Tokai reprocessing plant and the annual shipment will be approximately ten times greater. Therefore, in developing the cask, greatest emphasis was placed on the enhancement of transport efficiency to reduce transport frequency. To maximise. the payload, the cask weight was made as close as possible to the maximum capacity of the crane available at the reactor site. Crane capacities of the nuclear power plants were grouped into six, and six kinds of cask were developed. The design of all types of NFf casks was approved in 1994 and 40 units of the cask are under construction to prepare for shipment.

Estimate of the Crud Contribution to Shipping Cask Containment Requirements

This paper reports that a methodology is developed to relate U.S. Code of Federal Regulations, Title 10, Part 71 (10CFR71) containment requirements to leak rates for the special case in which the only radioactive species having a potential for escape form the cask is that associated with debris (crud) contained on the fuel assemblies being transported. The methodology accounts for the characteristics of the crud and for attenuation of the gas-borne crud particulates once they become suspended within the cask. Calculations are performed for typical spent-fuel transport cask geometries and the normal and accident conditions prescribed in 10CFR71. The most current published data are used for crud composition and structure, specific activity, spallation mechanics and fractions, and crud particle size. The containment criteria leak rates are calculated assuming 5-yr-old spent fuel. In each accident case, the containment leak rate criteria are well in excess of 10 cm{sup 3}/s. Under normal conditions of transport, the regulatory containment requirements are met by leak rates ranging from 1.5 {times} 10{sup {minus}3} cm{sup 3}/s to 1.5 {times} 10{sup {minus}4} cm{sup 3}/s for the transport of boiling water reactor fuel assemblies and form 1.8 {times} 10{sup {minus}2} cm{sup 3}/s to 1.3 {times} 10{sup {minus}3} cm{supmore » 3}/s for pressurized water reactor fuel assemblies. The calculated leak rates are most sensitive to the cask design, type of fuel, and particle size distribution. Conservatism of the limiting leak rates is discussed.« less

Feasibility and Incentives for Burnup Credit in Spent-Fuel Transport Casks

Feasibility and Incentives for Burnup Credit in Spent-Fuel Transport Casks