Fuel Cask Research Articles

Criticality safety analysis for eccentric loading of fuel assemblies in spent nuclear fuel cask

Eccentric fuel loading is one of the studies performed in a spent nuclear fuel criticality safety analysis. Several studies have been performed to investigate the effect of increasing reactivity on eccentric loading and random displacement in spent nuclear fuel pools; However, studies within spent nuclear fuel casks have not been performed. Therefore, this study investigates the effects of eccentric loading and random displacement of fuel assemblies on the reactivity in the spent nuclear fuel cask. The spent nuclear fuel cask and PLUS7 fuel are modeled using the MCNP code to investigate these effects. The result indicates that in eccentric cases, the k-effective values exceed those observed in the normal case. Also, several random displacement cases show k-effective values that exceed the normal case. These results differ from previous studies on eccentric loading in the spent fuel pool. It supports the necessity of considering both eccentric loading and random displacement in the criticality safety analysis in spent nuclear fuel cask. The increase in reactivity due to eccentric loading and random displacement of fuel assemblies within casks might be considered in criticality safety analysis and regarded as an uncertainty.

Muon Tomography for Reverification of Spent Fuel Casks (the MUTOMCA Project)

Muon Tomography for Reverification of Spent Fuel Casks (the MUTOMCA Project)

Repurposing a Spent Nuclear Fuel Cask for Disposal of Solid Intermediate Level Radioactive Waste From Decommissioning of a Nuclear Power Plant in Korea

Repurposing a Spent Nuclear Fuel Cask for Disposal of Solid Intermediate Level Radioactive Waste From Decommissioning of a Nuclear Power Plant in Korea

Neutron Tomography of Spent Fuel Casks

Dry casks for spent nuclear fuel (SNF) ensure the safe storage of SNF and provide radiation shielding. However, the presence of the thick casks encompassing several layers of steel and concrete makes inspection of the SNF a challenging task. Fast neutron interrogation is a viable method for the nondestructive assay of dry storage casks. In this study, we performed a Monte Carlo simulation-based study associated with a machine-learning-based image reconstruction method to verify the content of SNF dry storage casks. We studied the use of neutron transmission and back-scattered measurements to assess the potential damage to fuel assemblies or fuel pin diversion during transportation of dry casks. We used Geant4 to model a realistic HI-STAR 100 cask, MPC-68 canister and basket, and GE-14 fuel assembly irradiated by a D-T neutron generator. Several bundle diversion scenarios were simulated. The angular distribution of the neutrons scattered by the cask was used to identify the diversions inside the fuel cask. A fuel bundle with at least 75% of its pins removed can be identified with a drop in the back-scattered signature larger than $$2\sigma$$ compared with a fully loaded scenario. We combined an iterative reconstruction algorithm with a convolutional neural network (CNN) to obtain a cross-sectional image of the fuel inside the cask. The proposed imaging approach allows locating the position of a missing fuel bundle with at least 75% of the pins removed when performing tomographic imaging of a canister with an overall scan time of less than two hours, when using a commercial neutron generator with a source strength of $$10^{10}$$ n/s in the 4 $$\pi$$ solid angle.

Minimizing the Standard Deviation of the Thermal Load in the Spent Nuclear Fuel Cask Loading Problem

The paper assumes that, at the end of the operational period of a Spanish nuclear power plant, an Independent Spent Fuel Storage Installation will be used for long-term storage. Spent fuel assemblies are selected and transferred to casks for dry storage, with a series of imposed restrictions (e.g., limiting the thermal load). In this context, we present a variant of the problem of spent nuclear fuel cask loading in one stage (i.e., the fuel is completely transferred from the spent fuel pool to the casks at once), offering a multi-start metaheuristic of three phases. (1) A mixed integer linear programming (MILP-1) model is used to minimize the cost of the casks required. (2) A deterministic algorithm (A1) assigns the spent fuel assemblies to a specific region of a specific cask based on an MILP-1 solution. (3) Starting from the A1 solutions, a local search algorithm (A2) minimizes the standard deviation of the thermal load among casks. Instances with 1200 fuel assemblies (and six intervals for the decay heat) are optimally solved by MILP-1 plus A1 in less than one second. Additionally, A2 gets a Pearson’s coefficient of variation lower than 0.75% in less than 260s CPU (1000 iterations).

Investigation of Thermal Hydraulic Behavior of Dry Spent Fuel Cask Under Forced Convection

The thermal hydraulic behavior of the spent fuel in dry storage cask under forced convection mode is experimentally and numerically investigated. For this purpose a test rig is designed and constricted to simulate the cooling loop cask contains 21 spent fuel discharged from pressurized water reactor (PWR). A numerical simulation is performed by ANSYS-CFX fluid dynamic code. The effect of decay heat generation and inlet air velocity are investigated. The results show that the increase in the inlet air velocity improves the coolability of the fuel while the increase in decay heat leads to decrease the coolability of the fuel. Within the range (1.1< V < 2.8 m/s) for inlet air velocity and heaters power (630 < Q < 1260 watt), a new empirical correlation has been obtained for Nusselt number, Nu as a function in Reynolds number, Re. The comparisons between experimental and numerical results show a good agreement.

Design Process for Dual-Purpose Nuclear Spent Fuel Casks

Nuclear spent fuel management has become a key activity within national nuclear strategies. Spent fuel casks shall be customized, in order to release space in the pools of the nuclear power plants to continue operating. The final goal is to complete the decommissioning activities in some nuclear facilities or to load spent fuel with new enhanced irradiation parameters.Cask designers develop complex technical processes that require very skilled staff to understand and compile the requirements of the cask users. It’s necessary to adapt and validate the cask design concepts to comply with the applicable regulation. Finally, it is required to provide the detailed technical documentation as well as appropriate justifications to the nuclear regulators, for the safe management of the nuclear spent fuel.

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

A novel assessment has been completed to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing-basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed, calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014), and significant uncredited transportation dose rate margins were also observed. The results demonstrate that at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.

Development of Ductile Cast Iron for Spent Fuel Cask Applications Using Fracture Mechanics Principles

The structure–property relations of ductile cast irons (DCIs) with varying Cu content and ~1 wt% Ni has been investigated with an emphasis on examining their fracture toughness property towards the development of suitable materials for large volume containers for transport of spent fuel. The detailed microstructural characteristics, hardness, tensile and fracture toughness properties of three DCIs were assessed in as-cast and annealed conditions. Fracture toughness values were determined using both ball indentation (KBI) and J-integral (KJIc) tests. The obtained results assist to infer that: (1) the amount of pearlite and nodule count increases with increased amount of Cu, (2) the hardness and strength values increase whereas fracture toughness values marginally decrease with increased Cu content, and (3) the magnitudes of KBI estimated using a proposed analysis are in good agreement with KJIc values for the as-cast materials.

Preferential Water Ingress into a Dry Spent Fuel Cask

Water ingress into a fissile system, such as may occur during transportation of spent fuel casks, or in storage casks, may result in a significant increase in the keff of that system, which, if it were not accounted for in the abnormal operating or accident analysis of the cask, could potentially result in an inadvertent nuclear excursion with severe nuclear and radiological consequences. This paper will provide the results of an analysis of a gradual increase in water level in spent fuel casks and will discuss the way the keff of the system responds to such an increase. It will also provide the results of a preferential water ingress analysis, which sought to answer the question that if there is change in the keff of the system accompanied by a fractional increase in the amount of water in the cask, which one of the four water ingress pathways/channels would have the greatest effect on the keff. The water ingress pathways in question, and their respective unit numbers in brackets, are fuel rod (Unit 1), burnable poison rod (Unit 2), instrumentation tube (Unit 3) and cask air gap (Unit 9). This analysis will also determine which of the four channels has the highest sensitivity coefficient and will compare the effects of material composition on the keff when water is either freshwater or seawater.

Analysis on Dropping Accidents of Fresh Fuel Cask in HTGR

The fuel of HTGR is made of some spherical fuel elements with diameter of 60mm. The cask of fresh fuel is used to temporarily store and transport such a kind of new spherical fuel elements, and it may drop during transport. The dropping process can be simulated by two ways. The first is the coupled Euler-Lagrange (CEL) method, in which the new fuel elements are taking as fluid and can be considered explicitly. The second is referred to as the equivalent mass method. In the method, the cask and the new fuel are not considered individually, and the mass of the new fuel elements is averagely allocated to the bottom head of the cask. The simulating result shows that the CEL method can describe the flow and inertial effects of the new fuel elements in the dropping process, and the lateral fluid dynamic pressure generated by fuel elements to the cask. The impact effect can be fully considered in the equivalent mass method. Results show that the impact force is stronger and the contact time is shorter than that of the CEL method. Finally, parameters of equivalent fluid in the CEL method are discussed. And the results show that the fluid model is reasonable. In one word, calculation results of the two methods can be combined in structural design, which will give a more reasonable design.

Effects of Fuel Relocation for Transport Casks

Spent nuclear fuel transported in large casks must remain subcritical in all credible configurations for normal operation and hypothetical accident conditions. The effects on spent nuclear fuel reactivity from “worst-case” accident scenarios were surveyed in NUREG/CR-6835, “Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent Fuel Casks.” The survey used scenarios that were postulated to provide theoretical upper limits for reactivity effects of fuel relocation, although they were described as going “beyond credible conditions.” These scenarios involved physical changes either to fuel assembly rod arrays or to collections of fuel pellets with the fuel skeleton removed. To provide more credible estimates of the probability and maximum reactivity changes, a process is presented that deconstructs each scenario into a set of subscenarios and identifies the physical phenomena required to create the subscenario. The boundary between credible but unlikely scenarios and incredible scenarios is more easily discernible with this process.For marginally credible worst-case scenarios, it is concluded that the maximum reasonable reactivity increase either is less than the mandated administrative nuclear criticality safety margin for scenarios involving physical changes to fuel assembly rod arrays or is a substantial reactivity decrease for scenarios involving collections of fuel pellets. A cask designer could apply scenario deconstruction to evaluate the physical limits that apply to a particular transportation cask, and perform calculations specific to a particular cask design to show that criticality safety requirements are met.

ＭＣＮＰコードの金属キャスク貯蔵方式中間貯蔵施設線量評価への適用

ＭＣＮＰコードの金属キャスク貯蔵方式中間貯蔵施設線量評価への適用

Model and Method for Reasonable Shielding Calculation of Spent Fuel Casks

For the purpose of performing the reasonable shielding calculation of transport/storage spent fuel cask, the discussion of the cause of the discrepancy between the measured and calculated dose rate of a spent fuel cask is important. This paper shows the several items that may have large discrepancy between the measured and calculated dose rate of a spent fuel cask with respect to the calculation models and methods. The items include 1) the expression of source terms such as burnup profile, distribution of activated gamma-ray source, and gamma-ray source spectra, 2) the usage of a detailed geometrical model that is able to calculate the distribution of fission reaction precisely, and 3) the difference between detector response and dose conversion factor. The consideration of these effects may improve the agreement between the measurement and the calculation of a spent fuel cask.

Radiation Measurements and Shielding Calculations for Spent Fuel Casks

Radiation measurements were carried out on two kinds of newly developed NFT-type casks, the NFT-22B and the NFT-14P, in order to evaluate the cask shielding design safety margins and supply benchmarks for cask shielding calculation. The measurements involved placing about 100 sets of small dosimeters on and 1m away from the cask surface. The time integral dose equivalents for 2 to 3 weeks were measured. And gamma ray energy spectrums at some measurement points were also measured by a Nal scintillation counter. The measured dose equivalent rates were 1/2 -1/20 lower than the values estimated by the shielding design method which took the radiation source strength of the spent fuels loaded into the casks into consideration, confirming that the shielding design method used for the casks was conservative enough. The causes of the conservatism in the shielding design were also investigated. Furthermore the effects of changes in hydrogen and B4C concentrations in epoxy resin used for neutron shielding in the NFT casks were evaluated.

Numerical Assessment of Spent Fuel Casks Impacting on Real Targets

A reference container of high capacity was analysed for loads beyond those it has to withstand during a 9 m IAEA drop test onto an unyielding target. In doing this a lid-end drop with shock absorber onto a real target was simulated. This is a possible accident for the rail transport of such casks. In this case the most critical components of the containment system are the primary lid bolts. The behaviour of the lid system and its sealing function were investigated with finite element (FE) analysis. To correlate the findings with a corresponding impact velocity onto real targets an analytical method was used. Despite the conservative assumptions made in this study a two-fold safety factor compared to the 9 m drop tests onto the unyielding target could be shown. The quantification of the additional safety the cask might provide requires further basic investigations on the behaviour of the real targets considered as well as the reduction of the conservatism included in the assumptions made up to now.

Numerical Assessment of Spent Fuel Casks Impacting on Real Targets

Numerical Assessment of Spent Fuel Casks Impacting on Real Targets

Impact of an Exploding LPG Rail Tank Car Onto a Castor Spent Fuel Cask

Abstract On 27 April 1999 a fire test was performed with a 45 m3 rail tank car partially filled with 10 m3 pressurised liquid propane. A CASTOR THTR/AVR spent fuel transport cask was positioned beside the propane tank as to suffer maximum damage from any explosion. About 17 min after fire ignition the propane tank ruptured. This resulted in a BLEVE with an expanding fireball, heat radiation, explosion overpressure, and tank fragments projected towards the cask. This imposed severe mechanical and thermal impacts directly onto the CASTOR cask, moving it 7 m from its original position. This involved rotation of the cask with the lid end travelling 10 m before it crashed into the ground. Post-test investigaiions of the CASTOR cask demonstrated that no loss of leaktightness or containment and shielding integrity occurred.

Development of a Prediction System for Dose Rates Around a Fuel Cask

AbstractIt is important to determine the best way to arrange the fuel assemblies in the casks to reduce the effects of exposure on the operators that work with them. A system for arranging the fuel assemblies in TN-12/2 casks and predicting the dose rates around a cask using a personal computer is described here. For this system a new code, named SOURCE, has been developed to calculate the radiation source intensity precisely and quickly. In advance, the detailed shielding calculation for the cask is performed and the data which show the contribution from each fuel assembly location to a certain point around a cask is stored in the prediction system. To verify the accuracy of this prediction system, the measurement data of TN-12/2 casks are compared with the results of the system; fairly good agreements are obtained.

Burnup in Criticality Safety for Spent Fuel Casks

AbstractCriticality safety margins must be based upon the combination of the best available prediction of the margin and all uncertainties in the prediction. Inclusion of the effects of burnup in the evaluation of spent fuel shipping or storage casks must be based upon a thorough understanding of the prediction of the effects of burnup and the uncertainties in the measurements (or predictions) of burnup and predictions of the effects. A preliminary estimate of the effects of burnup and its uncertainties is presented. This will serve as a first step in the effort to develop acceptance criteria that assure public safety. An assembly average burnup of 20,000 MWD/MTU represents an increase in the criticality safety margin of about 20% (δk/k), and the current estimate of the uncertainty in this value is close to 4% (δk/k). The uncertainties in the components of the effects of burnup were based upon relevant literature citations and — where no other information was available — upon estimates. Consequently, the ...