Cask System Research Articles

Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part I: sensitivity studies.

The United States (US) Department Of Energy (DOE) has addressed the thermal analysis of the Spent Nuclear Fuel (SNF) stored within a dry cask system as a matter of high priority. In this regard, it is of utmost importance that simulation tools effectively reproduce the general thermal behavior of the modelled cask, including heat exchange and removal. Temperature distribution in the different components of the system is usually the focus of performed thermal analyses. In particular, attention is paid to the maximum temperature reached in the fuel cladding, namely the Peak Cladding Temperature (PCT). Within this framework, the present paper is the first of a two-paper series aimed at developing a more accurate model for the HI-STORM 100S cask. The dry cask in question is modelled and its behavior is simulated by means of the MELCOR code (version 2.2.18019). Stressing the need for a more realistic model rather than a conservative one, this paper reports the efforts undertaken to evaluate the influence of some specific modelling choices on the PCT. The study of the cask performance is therefore conducted taking into consideration three main factors: the axial power distribution in the Fuel Assembly (FA), the flow losses in the air gap between the internal canister and the external overpack, and the conductivity of the overpack concrete.

A Limited-Scope Probabilistic Risk Assessment Study to Risk-Inform the Design of a Fuel Storage System for Spent Pebble-Filled Dry Casks

This limited-scope study demonstrates the application of probabilistic risk assessment (PRA) methodologies to a spent fuel storage system for spent pebble-filled dry cask with a focus only on the necessary PRA technical elements sufficient to risk-inform the spent fuel storage system design. A dropping canister scenario in a silo of the spent fuel storage system is analyzed through an initiating event (IE) identification from the Master Logic Diagram (MLD); event sequence analysis (ES) by establishing the event tree; data analysis (DA) for event sequence quantification (ESQ) with uncertainty quantification; mechanistic source term (MST) analysis by using ORIGEN; radiological consequence analysis (RC) by deploying MicroShield, and risk integration (RI) by showing the Frequency-Consequence (F-C) target curve in the emergency area boundary (EAB). Additionally, a sensitivity study is conducted using the ordinary least square (OLS) regression method to assess the impact of variables such as failed pebble numbers, their location in the canister, and building wall thickness. Furthermore, the release categories grouped from the end states in the event tree are verified as safety cases through the F-C curve. This study highlights the implementation of PRA elements in a logical and structured manner, using appropriate methodologies and computational tools, thereby showing how to risk-inform the design of a dry cask system for storing spent pebble-filled fuel.

Modeling and results of the dry cask simulator (DCS) with STAR-CCM+

To understand the behavior of commercial nuclear fuel during its storage phase in spent fuel dry storage casks, simulations are necessary and to validate the simulation models, full-scale experiments are required. Therefore, an experimental facility called the dry cask simulator (DCS) was built in Sandia National Laboratories with the purpose of producing validation-quality data that can be used to test the accuracy of the modeling used to determine cladding temperatures in modern vertical dry casks. The DCS has been created under the framework of the Extended Storage Collaboration Programme (ESCP), coordinated by EPRI.The ultimate goal of the DCS was to study the thermal–hydraulic response of a single boiling water reactor (BWR) fuel assembly under variety of heat loads, internal vessel pressures, and external configurations. In this article, the simulations conducted by ENUSA Industrias Avanzadas S.A., S.M.E and UPM (Universidad Politécnica de Madrid, Spain) using STAR-CCM+ are validated against the experiments conducted at Sandia National Laboratories.In order to validate the simulation models several sensitivity analyses have been performed: a grid convergence index to estimate the discretization error in the simulations, a turbulence model sensitivity for the inner part of the cask, and three different approaches to model the fuel assembly: Case 1 (model with no spacers and no bottom tie plate), Case 2 (model with spacers and no bottom tie plate and Case 3 (model with spacers and bottom tie plate).Additionally, a thermal–hydraulic analysis is also presented, not only to explain the difference between the three ways to model the fuel assembly, but also to obtain a better understanding of the behavior of the dry cask according to pressure and heat output changes.Among the three cases, the simplest model (Case 1) has proven to obtain the most accurate results, being able to predict the temperature distribution in the fuel assembly and the air mass flow compared to the experiment measurements for all four cases of study. Meanwhile, the two other cases have shown that for steady state solutions the spacers and the bottom tie plate are not necessary and can introduce unwanted errors. Nonetheless, all cases have provided insight into the behavior of a dry storage cask system under different conditions showing the different heat transfer mechanisms interactions.

ITER Framework for RAMI Engineering

The systems engineering process starts with the discovery of the real issues that need to be resolved and the identification of failures that are the most probable or/and have the highest negative impact during the life cycle of a project. Systems engineering involves finding mitigations to these most critical problems. This logic is fully followed in reliability, availability, maintainability, inspectability (RAMI) engineering. Although this area is at its beginning in fusion technologies, a few years ago the ITER Organization developed an approach to assess the RAMI requirement of systems. As an example of what a RAMI analysis can bring to the maintainability and thus operational availability of a nuclear fusion facility like ITER, the availability of the cask and plug remote handling system in charge of handling of port plugs and their moving between the port cells to the hot cell facility is addressed in the case of diagnostic equatorial port plugs.

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

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

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

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

Consideration of spent fuel pool island as an interim management option of spent nuclear fuel for Kori unit 3 & 4 during decommissioning of Kori site

An effective decommissioning plan is a key to the safe management of spent nuclear fuel (SNF) and cost optimization during the decommissioning process. The multi-reactor Kori site in the Republic of Korea currently has 5 operating reactors and the oldest of these reactors, Kori 1, has been shut down and is scheduled to be decommissioned in the near future. This paper evaluates the feasibility of a spent fuel pool island (SFPI) for the Kori site as a part of spent fuel management strategy for decommissioning. The current decommissioning strategy for Kori 1 was reviewed and an overall decommissioning plan for Kori site was developed as a basis for this analysis.A cost assessment was performed to evaluate the cost effectiveness of an SFPI as an interim storage measure as compared to concrete cask, metal cask, and dual purpose transportation and storage cask (DPC) systems. Based on the decommissioning plan assumptions and economic parameters, the net present value (NPV) for these options was calculated and used as the basis for cost effectiveness comparisons. The total estimated cost of an SFPI is approximately $120 Million (M), which is significantly lower than that of a concrete cask system ($203 M), a metal cask system ($659 M) and DPC system ($874 M).

Design and implementation of cask system for EAST remote maintenance

An in-vessel maintenance robot called AIA has been developed by CEA to perform remote visual inspection inside the Tore Supra vacuum vessel during plasma. A project to upgrade the AIA into a fully operational robot has been undertaken with the collaboration between CEA-IRFM and ASIPP. Meanwhile, a new version of the robot called EAMA (EAST Articulated Maintenance Arm) is being developed in an Associated Laboratory for EAST in-vessel maintenance. For this purpose, a storage cask has been designed to provide vacuum and temperature conditioning and an interface to the EAST vacuum vessel for AIA (before) and EAMA (after).The cask system is composed of pumping system, baking system and cooling system. The control system of cask has been designed to ensure the well-prepared condition for robot deployment inside EAST vacuum vessel. The cask commissioning and the AIA deployment inside EAST have been presented to validate the feasibility of the cask system. This paper introduces the implementation of key technologies of cask system and presents the campaign activities.

A Criticality Analysis of the GBC-32 Dry Storage Cask with Hanbit Nuclear Power Plant Unit 3 Fuel Assemblies from the Viewpoint of Burnup Credit

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that k eff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

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.

Design and Implementation of Storage Cask System for EAST Articulated Inspection Arm (AIA) Robot

EAST Articulated Inspection Arm (AIA) robot is being mutually developed by ASIPP and CEA-IRFM for remote handling maintenance. It will permit remote visual inspection and to pick up small fragments inside the EAST tokamak vacuum vessel during experiments. Considering storage and support for EAST AIA, a sealed cask system has been designed and manufactured, which can be connected to EAST device through a ϕ250 mm connection port with two flashboard valves. The system consists of a 10 m long vacuum vessel with a linear guide rail for storage, guiding and conditioning, two mobile wagons for support and some auxiliary systems for keeping suitable work conditions and measurement. Besides, a stainless steel shuttle has been developed to support AIA robot and assemble with the linear guide. It can push the robot into tokamak vessel and back to the storage cask with a gear-rack driving mechanism. This paper mainly presents the overall description of the system design and some obtained implementation progress.

Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug

The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.

Study on the Flexible Support Legs for ITER Transfer Cask System

In fields of remote handing i.e. robot technology for fusion engineering reactor, such as ITER or the China fusion engineering test reactor, the flexible support legs are key components for their transfer cask system to adjust its position, joining to hot cell or tokamak ports for maintaining the fusion device. For ITER machine, each support leg should withstand maximum 50 tons load and adjust its height in 150 mm. Defect in original ITER design was presented. A new concept for the support legs was configured and its feasibility was proven. Detailed design and simulation was done for the new support leg with virtual prototype technology. Simulation results show that new support leg could not only meet all required function but also has merits of constant load during the tuning process with linear relation of control variable parameters, which is intended to be used for Tokamak reactors.

An investigation on the seismic response of free-standing dry storage cask for spent fuel

For most dry storage facilities for spent fuel worldwide, storage casks are freestanding on concrete pads. Thus, relative motion between the cask and the pad may be induced during earthquakes, leading to stability concerns. In this study, the seismic stability conditions of a free-standing cask were examined quasi-statically. Then, a finite element (FE) cask-pad model considering the frictional contact at the interface was established, and a dynamic analysis was performed to simulate a scaled cask shaking table test, which was conducted at the Central Research Institute of Electric Power Industry in Japan. Furthermore, the influence of friction coefficient of the cask/pad interface on the cask response was investigated using this FE model. Apparent rocking of the cask was induced at a higher friction coefficient, while sliding dominated the cask motion at a lower one. Since the friction coefficient is critical to the seismic stability of a free-standing cask, cyclic loading tests of a scaled pedestal model of a cask system called the High Performance System of the Institute of Nuclear Energy Research (INER-HPS), which will be used in Taiwan, were conducted to investigate the actual friction coefficient at the cask/pad interface as well as the effect of normal stress and sliding rate. Based on the test results, numerical simulations were performed to ensure the seismic stability of this INER-HPS cask under the design base earthquake.

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160VS. The plasma current will be 10MA and its major radius and minor radius is 5.7m and 1.6m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.

Theoretical Calculation and Kinematics Analysis of Double Seal Door for ITER Remote Handling Transfer Cask System

The moving direction of double seal door (DSD) of International Thermonuclear Experimental Reactor (ITER) remote handling transfer cask will change suddenly at the guide rail inflexion position (GRIP), where the force of hydraulic pole also will change significantly. The structure may damage and the system will possible be failed when the DSD moving through GRIP which is a mutant site. Based on special constitution, restriction conditions and working process of DSD, kinematics simulation and analysis were done by software ADAMS. The DSD moving equations were built based on the degree of freedom (DOF) of DSD mechanism system, and then the force of DSD moving were calculated in theoretical analyzing. By the above simulation and theoretical calculation the stress distribution of guide rail and hydraulic pole were obtained, at the same time optimization design of GRIP was confirmed according to the force analysis results. The above-mentioned analysis process and results not only provide technical data for the optimization design and the prototype manufacture of DSD, but also provide the examples and references of theoretical calculation and kinematics analysis for other important components of ITER.

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.

Seismic Response for a One-Third Scale-Down Vertical Cylindrical Cask Specimen Using Cyclic Loading Testing

Since the friction coefficient is critical to the seismic stability of a freestanding cask, cyclic loading tests of a 1/3 scale-down pedestal specimen of the INER-dry storage cask system, which will be used in Taiwan, were conducted to acquire the actual friction coefficient at the cask/pad interface as well as the effect of normal stress and sliding rate on it. Test results indicated that the cyclic frequency had few influence on the friction coefficient and the friction coefficient increased with the normal stress increased. The friction coefficient of the vertical cylindrical cask with an add-on shield, vertical cylindrical cask and add-on shield under actual loading condition was about 0.67, 0.60 and 0.73, respectively. Based on the frictional properties at the cask/pad interface obtained from the test results, it was performed to ensure the actual seismic stability of this INER-dry storage cask system under the design base earthquake.

Progress on the interface between UPP and CPRHS (Cask and Plug Remote Handling System) tractor/gripping tool for ITER

EFDA finances a training programme called Goal Oriented Training Programme for Remote Handling (GOT RH), whose goal is to train engineers in Remote Handling for ITER. As part of this training programme, the conceptual design of the mechanical interface between Upper Port Plug (UPP) and Cask and Plug Remote Handling System (CPRHS) as well as the conceptual design of the needed tools for UPP Remote Handling is carried out. The paper presents the conceptual design of the UPP/Gripping Tool Interface. This includes the conceptual design of the gripping tool for introducing/removing the UPP in/from the ITER port and the mechanical features on both sides of the UPP/Gripping Tool Interface (e.g. alignment features, mechanical connectors, fasteners). In order to develop the design of the interface between UPP and CPRHS it is necessary to first identify the functional requirements of the Transfer Cask System (TCS) and the CPRHS, such as required degrees of freedom (DoF), required performances of system, geometrical constraints, loading conditions, alignment requirements, RAMI requirements. These requirements are the input data for the design of the interface between UPP and gripping tool and some of them are also described in the paper.

Transfer cask system design activities: status and plan

The ITER Cask & Plug Remote Handling System (CPRHS), a.k.a. Transfer Cask System, is a critical element of the ITER Remote Maintenance System (IRMS) devoted to transportation of components between the Tokamak building and Hot Cell. Due to the necessary confinement of contaminated components the CPRHS is defined as Safety Importance Class 1 (SIC-1) plus the mobile nature of the CPRHS brings with it a significant number of complex interfaces with other ITER sub-systems. With a total CPRHS fleet in excess of 20 units, including seven typologies, the management of design and procurement needs to be carefully planned and implemented to ensure compliance with ITER's requirements. Fusion for Energy (F4E) and its beneficiaries/contractors are currently working under ITER Task Agreements (ITAs) on the conceptual design of the CPRHS and, following the signing of the Procurement Arrangement (PA) in mid 2012, will take responsibility for the entire CPRHS fleet. F4E must, therefore, develop a robust strategy to meet the needs of both ITER machine assembly (for which a number of CPRHS units will be utilised) and the remote maintenance of ITER. Within this context this paper will present the status of the current CPRHS design activities, highlight some of the significant issues which will be faced during procurement and present the overall strategy which is being implemented by F4E in order to meet these challenging objectives.