Shipping Cask Research Articles

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

Design and application of impact limiter for a spent fuel transport cask

As a key component of the spent fuel cask, the impact limiter can avoid the leakage of radioactive substances caused by the destruction of the container during the transport process and ensure the safe transport of nuclear spent fuel. In order to design a impact limiter that meets the engineering requirements, the effect of impact limiter size on the damping effect of the transport container was studied. The dispersity of wood compressive strength was obtained by wood compressive strength test. The 9m over center of gravity angle drop analysis of the transport container was carried out using the finite element method, and the 9m over center of gravity angle drop test was also carried out. The container acceleration and impact limiter deformation obtained by the test were consistent with the simulation results. The impact limiter of the cask is reasonably designed and can provide shock absorption and cushioning to the transport container. Generally speaking, the larger the size of the impact limiter, the more energy it absorbs but the damping effect decreases after its size exceeds a certain point. The simulation results show that when the energy absorption of the impact limiter is sufficient, the cushioning effect of wood with low compressive strength is better and the effect of wood compressive strength dispersity should be fully considered in the design.

Determination of linear attenuation coefficient of aggregate serpentine concrete exposed to gamma and neutron radioactive sources

This research was designed to perform linear attenuation calculations for ordinary concrete both through simulation and experiment. The linear attenuation coefficient is an essential parameter for radiation shielding design for both source transport casks and storage bunkers. The shielding properties of concrete designed with Serpentine aggregates of different granule sizes were investigated. The simulation was performed using Radpro computer Software while the experiment involved sourcing for local Serpentine rock, crushing into four different granule sizes of 5 mm, 10 mm, 15 mm and 20 mm respectively and casting the concrete samples, exposure of the samples to gamma and neutron radioactive sources and monitoring with a radiation survey meter, sample weight measurements and concrete sample crushing using a load testing machine for determination of the concrete compressive strength. The results of the linear attenuation calculations showed that there was high consistency in the values obtained by simulation with those obtained via experiment. Very high linear attenuation property was observed when the serpentine concrete samples were exposed to a neutron source, which corroborates the fact that Serpentine rock is a Boron-rich mineral and Boron is known to have high neutron absorption property. The experimentally determined linear attenuation coefficients showed that the values at 15 mm aggregate sizes were higher than those at 5 mm, 10 mm and 15 mm sizes respectively, which demonstrates that better shielding optimization, will be obtained when the concrete cask is fabricated with 15 mm aggregate size.

The crashworthiness of Bi-directional corrugated honeycomb aluminum under oblique loading using for nuclear spent fuel transportation casks

When filling materials used for nuclear spent fuel transportation casks, they will inevitably be subjected to oblique impact. The study of material performance during oblique impact is very important for their safety. In this paper, Bi-directional corrugated honeycomb aluminum (BDCHA) is used to carry out compression-shear loading results with different oblique angles (20°, 30°, 45°) and installation directions (X-FACE, Y-HOLE) by simulation (experiments are used to verify accuracy of the simulation). Quasi-static tests perform 6 kinds of materials dates with different cell thickness/cell size ratios, and observe the deformation mode and mechanical properties. The results show that the deformation of the material is mainly the torsional buckling deformation. In addition, BDCHA mainly bears the shear and compress stress. When the oblique angle and installation directions are different, the stress is different. The material of the cell thickness/cell size ratio affects the stress of the material. Finally, comparing with wood, the energy absorption shows the feasibility of BDCHA as a shock-absorbing filling material using for nuclear spent fuel transportation casks.

Development of Model to Evaluate Thermal Fluid Flow Around a Submerged Transportation Cask of Spent Nuclear Fuel in the Deep Sea

Development of Model to Evaluate Thermal Fluid Flow Around a Submerged Transportation Cask of Spent Nuclear Fuel in the Deep Sea

Effect of material properties of impact limiter steel on drop accident of spent nuclear fuel transport cask

Effect of material properties of impact limiter steel on drop accident of spent nuclear fuel transport cask

Optimization of radiation shields made of Fe and Pb for the spent nuclear fuel transport casks

Recommendations are given to improve the efficiency of radiation protection of transport casks for SNF transportation. The attenuation of γ-quanta of long-lived isotopes 134Cs, 137mBa(137Cs), 154Eu and 60Co by optimizing the thicknesses and arrangement of layers of Fe and Pb radiation shields of transport casks is studied. The fixed radiation shielding mass (fixed mass thickness) is chosen as the main optimization criterion. The effect of the placement order of Fe and Pb layers in a combined two-layer radiation shield with an equivalent thickness of 30 cm is studied in detail. It is shown that with the same mass thicknesses of the Fe and Pb layers, the placement of Fe in the first layer, and Pb - in the second one provides more than twofold attenuation of γ-quanta compared to the reverse placement: Pb - in the first layer, Fe - in the second. The increase in the efficiency of attenuation of γ-quanta for TC with combined shielding of Fe and Pb is shown to be achieved by designing the first layer of radiation shielding around the canister with SNF from Fe of the maximum possible thickness.

Experimental Evaluation of the Thermal Integrity of a Large Capacity Pressurized Heavy Water Reactor Transport Cask

Experimental Evaluation of the Thermal Integrity of a Large Capacity Pressurized Heavy Water Reactor Transport Cask

Criticality Safety Analysis of Spent Fuel Storage Pool for NPP Mochovce using MCNP5 Code

The paper presents results of nuclear criticality safety analysis of spent fuel storage and handling for the 1st and 2nd unit of NPP Mochovce. Spent fuel storage pool (compact and reserve grid) and T-12 transport cask were modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of max eff k values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal max eff k values. The requirement of current safety regulations to ensure 5% subcriticality was met except some especially conservative cases.

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

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

Evaluation of the permissible 99Mo activity in the KL-15 cask in the design of transportation and process scheme

The demand for the use of radioactive isotopes in medicine is increasing with each coming year necessitating the increased output of radionuclide products. One of the most widely spread radionuclides used in medicine is technetium-99m (99mТс) (Feasibility of producing molybdenum-99 2015, NEA 2012, The Supply of Medical Radioisotopes 2015). The very short 99mТс life (6-hour half-life) requires its production directly on the site of medical treatment. This is achieved using molybdenum-technetium generators (Kodina and Krasikova 2014, Technical Reports No. NF-T-5.4. 2013, Technetium-99 Generator 2021) loaded with molybdenum-99 (99Мо), which uninterruptedly decays (half-life of 66 hours) yielding 99mTc. Close attention must be paid in the course of production of molybdenum-technetium generators to radiation safety during transportation of 99Мо on the territory of the manufacturing facility. The main measure for ensuring radiation safety during transportation of 99Мо is the application of special packaging kits. The Karpov Institute of Physical Chemistry JSC uses a wide range of packaging kits of types A and B for transportation of radioactive materials on the territory of the manufacturer with design features providing the required level of radiation safety. In particular, the KL-15 shipping cask loaded/unloaded from the top is used for onsite transportation of 99Мо for charging molybdenum-technetium generators. The maximum permissible activity of 99Мо is not specified in the passport of the KL-15 cask. Planned construction of a radionuclide production shop in accordance with GMP requirements will require the increase of output of target radionuclides by several times. The above considerations necessitated the evaluation of the maximum permissible activity of 99Мо planned to be transported in KL-15 casks. No other type of standard casks can be used because of their outside dimensions prohibiting the unloading of 99Мо inside the “hot” chamber. Calculation and experimental evaluation of permissible 99Мо activity during transportation inside the KL-15 cask was performed. The paper presents the calculated evaluation of the maximum permissible activity of 99Мо in a KL-15 cask to ensure the radiation exposure of personnel of group A working with the cask not exceeding the established level at the enterprise (80 μSv per shift) and not requiring the use of additional measures and means of protection. The results of the work allow us drawing the conclusion that the KL-15 cask ensures the required level of radiation safety with up to 241 Ki of 99Мо loaded in the 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

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

Developing a radiation field-based monitoring system for the transport and storage cask inventory during extended interim storage

Abstract. High-level radioactive waste must be stored safely for a very long time but a suitable site for long-term storage is yet to be found. Additionally, in Germany the licence for transport and storage of casks is limited to 40 years, beginning at the time of emplacement and begins to expire for the first containers in the 2030s. To resolve the conflict between not having a final repository in time and the licence expiring, the licence must be extended while ensuring uninterrupted safe storage and safe transport. The DCS-Monitor II research project supports this process by investigating approaches to noninvasive, radiation field-based diagnostics that enable the observation of potential geometric changes of the cask inventory. Previous feasibility studies in the predecessor project DCS-Monitor I showed that cask monitoring using gamma radiation, neutron fields and cosmic muons is promising. In the ongoing research project presented here, the investigations on the qualification of radiation field-based diagnostics are extended via simulations. For muon imaging, a suitable procedure which solves the inverse problem to monitor the cask inventory is implemented. In addition, a partially automated gamma and neutron measurement system is being constructed and a muon detector is being developed and built. Both systems will be tested in field studies on large scale geometries and real containers.

Structural integrity of KJRR-F fresh nuclear fuel under vehicle-induced vibration for normal transport condition

Nuclear fuel, including its fresh state, must be handled safely due to its critical and hazardous nature. Under normal transport conditions, several interactions take place among different components, such as transport cask used for loading the nuclear fuel and tie-down structure to attach with the vehicle. To ensure structural integrity of the nuclear fuel, vibrations and impacts transmitted from the vehicle must be sufficiently reduced. Therefore, in this study, we conducted two transportation tests from Daejeon to Kijang in Korea to verify the vehicle-induced vibrational characteristics of the KJRR-F fresh nuclear fuel when transported under normal transport conditions. The speed and location of the vehicle were obtained via GPS, and the accelerations between the vehicle and the KJRR-F fresh nuclear fuel were measured. Additionally, using the acceleration results, a structural analysis was conducted to confirm the structural integrity of the nuclear fuel under the most severe conditions during normal transport.

Large eddy fire simulation applications from nuclear industry

Abstract A computational study has been carried out for predicting the behaviour of a pool fire source using the field-model based code Fire Dynamics Simulator (FDS). Time dependent velocity and temperature fields are predicted along with the resulting changes in the plume structure and its width. Firstly, a grid study was performed to find out the best grid size for this purpose. Then calculations were done which showed a very good agreement with earlier reported experimental based correlations for the temperature of the plume region. These studies have been extended to use this field-model based tools for modelling particular separate effect phenomena like puffing frequency and to validate against experimental data. There are several applications in nuclear industry like room fires, wildland fires, smoke or ash disposal, hydrogen transport in nuclear reactor containment, natural convection in building flows etc. In this paper the use of FDS with the advanced Large Eddy Simulation (LES) based CFD turbulence model is described for various applications: Fire simulation for Alpha storage, Bhabhatran teletherapy, pool fire for transport casks, fire PSA of a representative NPP, exhaust air fan buildings of a process plant and smoke dispersion in large fires around NPPs.

Failure probability assessment of SNF cladding transverse tearing under a hypothetical transportation accident

This paper presents a structural assessment of spent nuclear fuel (SNF) rod under drop accident of a transport cask using deterministic and probabilistic approaches. The performed case study is a baseline 9-m side-drop scenario during the first 30 years of dry storage. A hypothetical database is generated for five variables related to operating conditions to predict the changes in the mechanical properties of SNF rod using Monte Carlo Simulation and material constitutive models. Finite-element analyses are conducted including a sophisticated transport cask with simplified fuel assemblies to obtain the acceleration response and a detailed full-length single fuel rod to evaluate the deformation response of fuel cladding with various representative material properties. Failure probabilities of a fuel rod subjected to a transverse inertia loading due to side-drop are estimated based on ductile tearing failure criteria. The findings from the dynamic response analysis show that the cladding might yield owing to the 9-m side-drop, but the structural integrity of SNF rod is generally maintained. Yet the probabilistic assessment highlights the importance of conducting further structural integrity evaluation for high burnup fuel under hypothetical accidents as it initially shows concerned failure probabilities comparing to low and intermediate burnup fuel. The modeling technique and analysis methodology outlined in this study have a potential advantage to consider several geometric and material characteristics of SNF rod, hence improving the structural integrity assessment of SNF further.

Energy absorption and optimization of Bi-directional corrugated honeycomb aluminum

To replace wood as the filler material for the impact limiter of the spent nuclear fuel transportation cask, a bi-directional corrugated honeycomb aluminum (BDCHA) was manufactured. First, five materials with different cell thickness/cell size ratios (t0/a) were designed, and compression tests were performed under quasi-static conditions. The results show that the initial peak stress and the plateau stress both increase with the increase of the t0/a ratio. The plateau zone determines the energy absorption performance; therefore, this study focuses on the analysis of the mechanical properties of the plateau stage. Furthermore, through the compression theory, the energy dissipation process of the unit cell was evaluated, and the mean stress equation was obtained; the theoretically obtained plateau stress was similar to the test plateau stress. Moreover, the ideal constitutive relationship of the material based on the t0/a ratio under static load was derived. Finite element analysis was used for simulation. The stress-strain results of the test and the simulation were close. Furthermore, the structural parameters were optimized via multiobjective optimization. Through the analysis of the mechanical properties of BDCHA, derived its rationality and feasibility as a shock-absorbing filling material.

Spent fuel management strategies for fluoride-cooled pebble bed reactors

Spent nuclear fuel from fluoride-salt-cooled high-temperature reactors presents different set of storage and handling challenges than fuel from light water reactors and requires the implementation of new strategies. This study explores potential strategies involving pool storage (in helium or FLiNaK) and cask storage/transportation in sequential phases and assesses the feasibility of such strategies in terms of thermal-hydraulics, criticality safety and radiation shielding. The most demanding constraints to the design of the pool storage system are set by criticality safety and require distributing the spent fuel in separate containers. Subcriticality of the fuel in the casks can be maintained by mixing boron-doped graphite spheres with the spent pebbles. High dose rates at the cask surface impede immediate loading of the pebbles in storage/ transportation casks, but the use of steel and lead based overpacks offers effective radiation shielding and good thermal properties, and allows loading spent fuel in the casks after as few as two years of pool storage.