Shipping Cask Research Articles

Effect of γ irradiation on the impact response of rigid polyurethane foam

Abstract An experimental study to observe the change in impact response of rigid polyurethane foam due to γ irradiation is presented. The foam samples were irradiated using a 60Co based γ irradiator, γ Chamber 5000. Dynamic tests were performed on unirradiated and irradiated cubical foam specimens in a drop weight testing machine to characterize the mechanical properties. Structural characterization was performed using phase-contrast X-ray imaging and scanning electron microscopy (SEM). The results reveal a significant change in the dynamic response of polyurethane foam after irradiation due to change in structure. The response of foam gets stiffer in the initial linear phase, and densification (lockup) occurs well below the strains at which lockup occurs for an unirradiated foam. The ramifications of this study towards the application of polyurethane foam as an impact limiter in nuclear shipping casks are discussed.

Human machine interface to manually drive rhombic like vehicles such as transport casks in ITER

In nuclear facilities, such as the experimental fusion reactor ITER, the cargo transfer operations can be performed by autonomous guided vehicles under remote supervision. In ITER, these vehicles can reach up to 100 tons and, with a rhombic-like configuration, have to move in cluttered scenarios. In case of failure, the vehicles have to be manually guided. This paper presents three solutions for the teleoperation of rhombic-like vehicles. Two set of devices were used to test each solution: one is based on a gamepad and the other is based on a joystick with a rotational disc specially designed for this purpose. The solutions were experimented by the developer and by 12 users without prior experience on rhombic-like vehicles. The experiments were performed in a software simulator that provides 2D maps of the test facility and simulates the kinematics of the vehicles in real time. The main conclusions are reported.

Health physics operating experience in transporting cleanup spent resin and verification of methodology for estimating the shielding thickness of cask

Cleanup (CU) demineralizer of TAPS 1 and 2, containing mixed bed resins, is used to maintain reactor water chemistry. Once the bed is saturated, spent resin needs to be transferred to resin fixation facility for fixation and subsequent disposal in near-surface disposal facility. This is done by transporting the spent resin in a shielded resin transportation cask (SRTC) from radwaste building. The SRTC was manufactured with shielding thickness, based on methodology proposed by Rautela et al., “Shielding adequacy of proposed cask for transporting TAPS 1 and 2 CU spent resins.” Extensive radiation monitoring was carried out during the entire transfer and transportation process. Observed radiation levels on cask surface were within the accepted and calculated radiation levels. This paper describes verification methodology proposed by Rautela et al. and health physics operating experience during first CU spent resin transfer process at TAPS 1 and 2.

Tyrosine deposits in brines of salted natural sausage casings of bovine origin

Abstract Tyrosine was identified as the major component of water insoluble white particles recovered from salted bovine small intestines intended for sausage casings. The salt sludge taken from the bottom of the transport casks contained more than 2.5% (w/w) of these particles and the wash-off from the salted intestines yielded 0.04% (w/w). Qualitative analysis by confocal Raman microscopy proved tyrosine as the major constituent with an admixture of phenylalanine. The quantitative amino acid analysis revealed 85.5 ± 2.3% tyrosine and 5.0 ± 0.2% phenylalanine. The amino acid analysis of the brine samples showed a typical protein profile with the highest concentrations being observed for glutamate, leucine, alanine, and lysine. Endogenous proteolytic enzymes of the tissue rather than microbial activities are thought to have been responsible for the tyrosine precipitations.

The radiation characteristics of the transport packages with vitrified high-level waste

The calculation method of neutron yield in the (α, n) reaction for a homogeneous material of arbitrary composition is represented. It is shown that the use of the ORIGEN 2 code excluding the real elemental composition of vitrified high-level waste leads to significant underestimation of the neutron yield in the (α, n) reaction. For vitrified high-level waste and spent nuclear fuel from VVER, the neutron fluxes are analyzed. The thickness of the protective materials for a transfer cask and a shipping cask with vitrified highlevel waste are estimated.

Burnup Credit Implementation for PWR UOX Used Fuel Assemblies in France: From Study to Practical Experience

Burnup credit for used fuel assemblies, in particular, pressurized water reactor uranium oxide, has been a major focus of research in France for more than 30 years. As a result, a wealth of knowledge and experience has been gained. The first implementation of burnup credit in France used the “50-leastirradiated-cm” method approved by the French safety authority in the early 1980s. However, because of the continuous increase in fuel enrichment, the industry is interested in reducing the conservatisms of this method by taking into consideration more realistic hypotheses, such as the introduction of fission products and a nonuniform axial burnup profile. To address this concern, a working group, bringing together several French nuclear companies and institutions [AREVA, CEA (Commissariat à l’Energie Atomique et aux Energies Alternatives), EDF (Electricité de France), and IRSN (Institut de Radioprotection et de Sûreté Nucléaire)], was created in 1997. This paper presents the results of the working group’s discussions and studies on all the issues pertaining to the use of burnup credit. In addition, the practical experience of AREVA TN (a division of AREVA dealing with radioactive materials transport and storage throughout the entire nuclear fuel cycle) with transport casks, using input from these results, is described.

Multi-purpose deployer for ITER in-vessel maintenance

The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

Thermal Evaluation of Alternate Shipping Cask for Irradiated Experiments

Results of a thermal evaluation are provided for a new shipping cask under consideration for transporting irradiated experiments between the test reactor and postirradiation examination (PIE) facilities. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory and then later shipped to the Hot Fuel Examination Facility located at the Materials and Fuels Complex for PIE. To date, the General Electric (GE)-2000 cask has been used to transport experiment payloads between these facilities. However, the availability of the GE-2000 cask to support future experiment shipping is uncertain. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger payloads. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and nonfueled payloads. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask for shipping irradiated experiment payloads.

First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C−E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA–AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TN's application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies.

Effect of beam oscillation on borated stainless steel electron beam welds

Abstract Borated stainless steels are used in nuclear power plants to control neutron criticality in reactors as control rods, shielding material, spent fuel storage racks and transportation casks. In this study, bead on plate welds were made using gas tungsten arc welding (GTAW) and electron beam welding (EBW) processes. Electron beam welds made using beam oscillation technique exhibited higher tensile strength values compared to that of GTA welds. Electron beam welds were found to show fine dendritic microstructure while GTA welds exhibited larger dendrites. While both processes produced defect free welds, GTA welds are marked by partially melted zone (PMZ) where the hardness is low. EBW obviate the PMZ failure due to low heat input and in case of high heat input GTA welding process failure occurs in the PMZ.

Releasable activity and maximum permissible leakage rate within a transport cask of Tehran Research Reactor fuel samples

Containment of a transport cask during both normal and accident conditions is important to the health and safety of the public and of the operators. Based on IAEA regulations, releasable activity and maximum permissible volumetric leakage rate within the cask containing fuel samples of Tehran Research Reactor enclosed in an irradiated capsule are calculated. The contributions to the total activity from the four sources of gas, volatile, fines, and corrosion products are treated separately. These calculations are necessary to identify an appropriate leak test that must be performed on the cask and the results can be utilized as the source term for dose evaluation in the safety assessment of the cask.

발사 충격을 받는 방사성 물질 운반용기의 건전성 평가

발사 충격을 받는 방사성 물질 운반용기의 건전성 평가

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

Modelling a Liquid Material in Drop Test Simulations of a Cask for Liquid Radioactive Waste

The aim of this paper is the comparison of different material models for the simulation of fluid impact on a cask for transport of radioactive liquid material. Simulated is a 9 m drop test performed according to International Atomic Energy Agency regulations. In order to reduce computational time, proposed is to model the transported fluid as a hypothetic linearly elastic material. This enables us to use less time demanding FE explicit dynamic analysis based on Lagrange formulation only instead of combination of Lagrange and Euler formulations. Compared are the results obtained for three elastic material and three fluid models based on different equations of state.

German approach and feedback on experience of transportability of SNF packages after interim storage

AbstractIn Germany, the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised ‘Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks’ by the German Waste management Commission. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties, which satisfy the proofs for the compliance of the safety objectives at that time. In recent years, the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case depen...

Optimization of cask for transport of radioactive material under impact loading

Casks used for transporting radioactive material need to be certified fit by subjecting them to a specific set of tests (IAEA, 2012). The high cost of these casks gives rise to the need for optimizing them. Conducting actual experiments for the process of design iterations is very costly. This work outlines a procedure for optimizing Type B(U) casks through simulations of the 9m drop test conducted in ABAQUS®. Standard designs and material properties were chosen, thus making the process as realistic as reasonable even at the cost of reducing the options (design variables) available for optimization. The results, repeated for different source cavity sizes, show a scope for 6.9% reduction in cost and 4.6% reduction in weight over currently used casks.

Finite element modeling of Balsa wood structures under severe loadings

In order to compute, in various situations, the requirements for transporting packages using Balsa wood as an energy absorber, a constitutive model is needed that takes into account all of the specific characteristics of the wood, such as its anisotropy, compressibility, softening, densification, and strain rate dependence. Such a model must also include the treatment of rupture of the wood when it is in traction. The complete description of wood behavior is not sufficient: robustness is also necessary because this model has to work in presence of large deformations and of many other external nonlinear phenomena in the surrounding structures. We propose such a constitutive model that we have developed using the commercial finite element package ABAQUS. The necessary data were acquired through an extensive compilation of the existing literature with the augmentation of personal measurements. Numerous validation tests are presented that represent different impact situations that a transportation cask might endure.

Research on MEMS Based Real-Time Measurement System for the Nuclear Fuel Transport Cask

A new type of measurement and recording device for the nuclear fuel transport cask was presented on this paper. With the view to the characteristics of specific application object, the system has been designed with the integration of the MEMS sensors. Several kinds of MEMS sensors include MEMS gyroscopes, MEMS accelerometers, MEMS magnetometers and MEMS pressure sensors are integrated together with embedded CPU, micro GPS receiver, and communication unit. With the conception of the modularization, the multi-subsystem (including: stainless steel sleeve structure, shock absorbers, sealed loader, MEMS sensor components, FLASH memory, power supply components, wireless communication components, etc.) was combined to a new transport cask of nuclear fuel. The low power, low cost, high integrate MEMS sensor system can achieve date of the whole process of nuclear fuel conveying condition, and alarm at the licensing of circumstances beyond. The new nuclear fuel transport cask system include: stainless steel sleeve structure, shock absorbers, sealed loader, MEMS sensor components, FLASH memory, power supply components, wireless communication components. The self-programmed software provides a means to analyze all the data during the whole transport process. The sensors’ output, attitude variety, temperature, acceleration, GPS trace can be analyzed simultaneity or independently. Using the data recorded, System Identification is performed to find the final health status of nuclear fuel.

Additional studies of criticality safety of failed used nuclear fuel

Commercial used nuclear fuel (UNF) in the USA is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high burnup values (>45 GWd t−1) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF are not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on the criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories, and specific configurations were ev...