Multipurpose Canister Research Articles

Fracture mechanics analysis of multipurpose canister for spent nuclear fuels under horizontal/oblique drop accidents

In this paper, elastic-plastic fracture mechanics analysis is performed to determine the critical crack sizes of the multipurpose canister (MPC) manufactured using austenitic stainless steel under dynamic loading conditions that simulate drop accidents. Firstly, dynamic finite element (FE) analysis is performed using Abaqus v.2018 with the KORAD (Korea Radioactive Waste Agency)-21 model under two drop accident conditions. Through the FE analysis, critical locations and through-thickness stress distributions in the MPC are identified, where the maximum plastic strain occurs during impact loadings. Then, the evaluation using the failure assessment diagram (FAD) is performed by postulating an external surface crack at the critical location to determine the critical crack depth. It is found that, for the drop cases considered in this paper, the principal failure mechanism for the circumferential surface crack is found to be the plastic collapse due to dominant high bending axial stress in the thickness. For axial cracks, the plastic collapse is also the dominant failure mechanism due to high membrane hoop stress, followed by the ductile tearing analysis. When incorporating the strain rate effect on yield strength and fracture toughness, the critical crack depth increases from 10 to 20%.

CFD analysis of spent fuel dry cask storage system for High burnup nuclear fuel

Storage of nuclear spent fuel and cooling of the fuel assembly is one of the vital requirements for nuclear safety. This engineering technology involves a dynamic thermo-fluid phenomenon in the dry storage casks. Therefore, this study investigated HI-STORM 100 dry cask storage container for thermal-fluid with sensitivity analysis using 3D Computational Fluid Dynamics (CFD) simulations. It is a vertical concrete cylindrical structure (6.09 m height and 3.36 m diameter) containing a welded stainless-steel sealed barrel multi-purpose canister (MPC) for storing the fuel components replaced by the reactor. MPC-32 storage barrel (4.85 m height, 1.78 m outer diameter) and Westinghouse 17 × 17 high burnup (45GWd/MTU) Pressurized Water Reactor (PWR) fuel assembly is selected in this research. 2D and 3D models of this assembly were simulated using CFD with porous media assumption. Further, Boussinesq approximation and incompressible ideal gas models are adopted in simulation; and standard k-ε and k-ω Shear Stress Transport (SST) turbulence models were also compared under natural convection. The internal flow and temperature of the MPC by the incompressible ideal gas model were found to be in line with the actual situation. The standard k-ε model can cause abnormal results at the high-temperature position. The sensitive factors are analyzed to predict the temperature distribution and flow field characteristics of the system during 10–55 cooling years. During this period, the decay thermal power of nuclear waste, the maximum temperature of the system and shell; the air outlet temperature and the volumetric flowrate decreased. Under non-uniform thermal conditions, the heat dissipation performance was studied defining the decay heat ratio (X). The heat dissipation performance of the system was found better when X ≥ 1, and the maximum temperature of the fuel assembly rise when X < 1.

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

Abstract In this paper, shielding analysis was performed to determine neutron and gamma dose rates around the transfer cask HI-TRAC VW (Holtec International Transfer Cask Variable Weight) loaded with spent fuel assemblies (SFA) from nuclear power plant (NPP) Krsko spent fuel dry storage (SFDS) campaign one. The HI-TRAC VW is a multilayered cylindrical vessel designed to accept a multipurpose canister (MPC) during loading, unloading, and transfer to dry storage building. The MPC can contain up to 37 spent fuel assemblies. The analysis was divided into two steps. The first step was the source term generation using origen-s module of the scale code package. The source was calculated based on the operating history of spent fuel assemblies currently located in the NPP Krsko spent fuel pool. The obtained particle intensities and source spectra of the SFA were used in the second step to calculate the dose rates around the transfer cask. A comprehensive hybrid shielding analysis included the calculation of dose rates resulting from fuel neutrons and gammas, neutron-induced gammas (n-γ reaction), and hardware activation gammas under normal conditions and during accident scenario. To obtain the dose rates within the acceptable uncertainties, forward weighted consistent adjoint driven importance sampling (FW-CADIS) variance reduction scheme, as implemented in advantg code, was adopted for accelerating final mcnp6.2 calculations. The dose rates around HI-TRAC VW cask were calculated using mcnp6.2 code for all 16 cask loadings belonging to campaign one in order to illustrate the impact of fuel assembly selection schemes proposed by company responsible for project realization - Holtec International (HI).

Preventing Stress Corrosion Cracking of Spent Nuclear Fuel Dry Storage Canisters

Multi-purpose canisters (MPCs) are being employed for dry canister storage of spent fuel at nuclear plant sites as a temporary approach until an interim or permanent dry storage site(s) is available. For storage in coastal or lakeside regions and even humid environments, the corrosive nature of the moist air with entrained chlorides can make the welded regions of the canisters susceptible to pitting and chloride induced stress corrosion cracking (CISCC). In this report we evaluate CISCC lifetimes of welded 316L stainless steel canister plates showing in excess of 19 times increase of laser peened panel sections vs. those left as-welded using ASTM G36 (2013) accelerated corrosive testing. Specifically cracks never developed or propagated into the laser peening area. We also provide measurements of residual stress in test plates and related calculations of stress intensity and depth expected in the full canister geometry. We discuss the relevance of stress depth to pitting depth and crack growth rates. For this project welded 316L stainless steel panels were configured to MPC geometry and laser peened in the same manner as deployed for actual canisters. Our two-dimensional stress mapping shows that high energy laser peening provides deep (>5 mm) plastic compression. Literature information is not clear as to specific dependence of crack initiation and growth rates on variables including temperature, specifics of chlorine exposure, pit nucleation and residual stress intensity and depth. In our CISCC testing, although extensive cracking developed quickly (less than 18 hours) in the unpeened area, no cracking developed at all in the laser peened sections after 340 hour exposure. We further show that cracking that initiated and grew in unpeened regions arrested upon propagating to the laser peened boundary. We performed detailed finite element stress analysis (FEA) to evaluate the difference in expected retention of stress generated by the peening of test panels and the more constrained geometry of actual storage canisters. Our analysis shows that the 4 mm depth of stress measured in unconstrained test panels correlates to actual stress depth in the confined canister geometry of 6 mm, that is approximately 2 mm deeper thus satisfying a key stress depth requirement. The accelerated CISCC tests indicate that MPC design life can be dramatically improved by laser peening thereby making the MPCs a viable technical and economic solution to prevent CISCC. Following qualification testing and system configuration, a high energy laser peening process was deployed to peen welds of the dry fuel canisters for the San Onofre Nuclear Power Plant. The laser peening thereby helps ensure that the storage canisters will remain free from chlorine induced stress corrosion cracking.

A numerical study of spent nuclear fuel dry storage systems under extreme impact loading

The structural integrity of concrete dry storage systems under an impact loading caused by a simulated commercial airplane engine crash is evaluated in this study. A finite element model of a generic vertical dry storage system including the concrete overpack, multi-purpose canister, spent nuclear fuel (SNF) basket, and SNF assemblies was developed using a commercial finite element analysis software. The strain rate effect was considered for the steel parts of the cask via a rate dependent and nonlinear constitutive model. The modeling approach was first validated using experimental data available in literature on projectile impact of concrete and steel/concrete composite (sandwich) panels. Then, a comprehensive comparison in terms of strains, damage and deformations was performed for two configurations of reinforced and sandwich cask overpacks commonly used in the industry. Critical impact height was identified by an examination of the strains and accelerations in the SNF assemblies. Furthermore, the effect of boundary conditions for the two cases of free-standing and anchored configuration was investigated. The maximum deformation and stress as well as the peak acceleration were identified in the SNF basket to assess safety of the SNF assemblies. The minimum impact velocity causing tip-over from aircraft engine impact was determined.

Flaw Stability Considering Residual Stress for Aging Management of Spent Nuclear Fuel Multiple-Purpose Canisters

A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel (SNF) assemblies. Because heat treatment for stress relief is not required for the construction of the MPC, the canister is susceptible to stress corrosion cracking in the weld or heat affected zone (HAZ) regions under long-term storage conditions. Logic for flaw acceptance is developed should crack-like flaws be detected by Inservice Inspection. The procedure recommended by API 579-1/ASME FFS-1, Fitness-for-Service, is used to calculate the instability crack length or depth by failure assessment diagram (FAD). It is demonstrated that the welding residual stress (RS) has a strong influence on the results.

Computational fluid dynamics simulations of fuel cladding and basket surface temperatures in multipurpose canister rail cask during normal transport

AbstractA two-dimensional finite volume model of a Multi-Purpose (MPC) in a rail cask with twenty-one pressurised water reactor (PWR) assemblies inside was constructed. Steady state thermal simulations were performed for a range of fuel heat generation rates, for both nitrogen and helium cover gas, and different fuel cladding emissivities. Geometrically accurate computational fluid dynamics (CFD) simulations were employed to calculate buoyancy induced motion in, and natural convection and radiation heat transfer across, all gas filled regions. The results are compared to stagnant-gas CFD (S-CFD) simulations in the same geometrically accurate mesh, and to simulations that employ Effective Thermal Conductivities (ETC) in a mesh with homogenized fuel/cover gas regions. The cask Thermal Dissipation Capacity (QTDC) is defined as the fuel heat generation rate that causes the fuel cladding to reach its allowed temperature limit. QTDC is 27% larger when helium is the backfill gas than for nitrogen. The QTDC predi...

Demonstrative drop tests of transport and storage full-scale canisters with high corrosion-resistant material

The concrete modular dry storage technologies are becoming widely used, aiming at better economic performances. With longer perspective, there are still key issues related to safety standards in operation and maintenance during storage and unloading/loading for transportation, long-term integrity of the storage components. On the other hands, due to the lack of the experimental studies related to tipping-over or drop event scenarios, it is very important to evaluate the characteristics of the behavior of the components, such as the multi-purpose canister (MPC) subjected to impact loads. Therefore, we executed the demonstration drop test program using double-lid welded MPCs, with the aim of obtaining basic data for regulating safety. This paper introduces the summary of our drop test program.

HI-STAR 100 spent fuel transport cask analytical evaluation for drop events

The US Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (CFR), Title 10, Part 71. Both the International Atomic Energy Agency (IAEA) regulations for transporting radioactive materials (Ref. 2, paragraph 727) and 10 CFR 71·73 require casks to be evaluated for hypothetical accident conditions, which includes a 9 m (30 feet) drop impact event on a flat, essentially unyielding, horizontal surface, in the most damaging orientation. The present paper examines the behaviour of one of the NRC certified transportation casks, the HI-STAR 100, for drop impact events. The specific area examined is the behaviour of the bolted connections in the 'overpack' top flange and the closure plate, which are significantly loaded during the hypothetical drop impact event. The term 'overpack' refers to the cask that receives and contains a sealed multipurpose canister (MPC) containing spent nuclear fuel. The analytical work to evaluate the NRC certified HI-STAR 100 spent fuel transport cask for a 9 m (30 feet) drop impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS and LS-DYNA finite element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9 m (30 feet) free drop hypothetical accident conditions, as specified in 10 CFR 71·73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1 m (40 inch) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71·73, was not included in the current work. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the closure plate surfaces may occur, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event.

Transport Package Response to Severe Thermal Events, Part 1: Rail Package

The response of intact and damaged versions of a 125 ton Multi-Purpose Canister (MPC) rail package conceptual design to severe thermal events is simulated using finite element computer analysis. The ‘critical’ fire duration (the minimum duration that causes the fuel cladding temperature to reach 740°C) is calculated as a function of fire temperature and for different modelling assumptions. It is found that fire temperatures below 660°C cannot cause the cladding to reach 740°C, no matter how long they last. At higher fire temperatures, the critical duration versus fire temperature envelope is significantly affected by the presence of the external neutron shield (especially for fire temperatures > 720°C), and somewhat less affected by the assumed value for the fuel cladding temperature limit. It is insensitive to the fuel region thermal conductivity model. Due to the large mass of the MPC, significant safe margins exist between the conditions of the IAEA thermal event and all of the performance envelopes calculated in this work.