Radiological aspects and behavior of spent fuel considering long-term storage

For spent nuclear fuel management in Germany, the concept of dry interim storage in dual purpose casks before direct disposal is applied. Current operation licenses for storage facilities have been granted for a storage time of 40 years. Due to the current delay in site selection, an extension of the storage time seems inevitable. In consideration of this issue, GRS performed burnup calculations, thermal and mechanical analyses as well as particle transport and shielding calculations for UO₂ and MOX fuels stored in a cask to investigate long-term behavior of the spent fuel related parameters and the radiological consequences. It is shown that at the beginning of the dry storage period, cladding hoop stress levels sufficient to cause hydride reorientation could be present in fuel rods with a burnup higher than 55 GWd/tHM. The long-term behavior of the cladding temperatures indicates the possibility of reaching the ductile-to-brittle transition temperature during extended storage scenarios. Surface dose rates are 3 times higher when a cask is partially loaded with 4 MOX fuel assemblies. Due to radioactive decay, long-term storage will have a positive impact on the radiological environment around the cask.     

Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

Abstract

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel.     

Radial-hydride Embrittlement of High-burnup Zircaloy-4 Fuel Cladding

Prestorage drying operations of high-burnup fuel may make Zircaloy-4 (Zry-4) fuel cladding more susceptible to failure, especially during fuel handling, transport, and post-storage retrieval. In particular, hydride precipitates may reorient from the circumferential to the radial direction of the cladding during drying operations if a threshold level of hoop stress at or above a corresponding threshold temperature is exceeded. This study indicates that the threshold stress is approximately 75–80 MPa for both nonirradiated and high-burnup stress-relieved Zry-4 fuel cladding cooled from 400°C and, under ring compression at both room temperature and 150°C, that radial-hydride precipitation embrittles Zry-4. Specifically, the plastic tensile hoop strain needed to initiate unstable crack propagation along radial hydrides decreases dramatically from >8% to lt;1% as radial-hydride fraction increases. Lower hydrogen contents (lr;300wppm) appear to be more susceptible to radial-hydride embrittlement compared to higher contents (>600 wppm), like that found in high-burnup Zry-4.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

High-burnup fuel failure during a reactivity-initiated accident has been a subject of safety-related concern. Because of wide variations in cladding metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior observed in tests in the SPERT, NSRR and CABRI reactors. In this paper, we propose a failure model that is based on temperature-sensitive tensile properties and fracture toughness. The model assumes that dynamic fracture toughness and high-strain-rate tensile properties of high-burnup cladding are sensitive to temperature and exhibit ductile–brittle transition phenomena similar to those of bcc alloys. Significant effects of temperature and shape of the pulse are predicted when a simulated test is conducted near the cladding material's ductile–brittle transition temperature. Temperature dependence of tensile properties and fracture toughness is, in turn, sensitive to cladding microstructural characteristics such as density, distribution and orientation of hydrides; distribution of oxygen in the metallic phase; and irradiation-induced damage. Because all of these characteristics are strongly influenced by corrosion, the key parameters that influence susceptibility to failure are oxide layer thickness and hydriding behavior. Therefore, high-burnup fuel failure is predicted to be more sensitive to local cladding corrosion (e.g. grid span location) than to fuel burnup.     

Effect of hydride orientation on fracture toughness of Zircaloy-4 cladding

Hydrogen embrittlement is one of the major degradation mechanisms for high burnup fuel cladding during reactor service and spent fuel dry storage, which is related to the hydrogen concentration, morphology and orientation of zirconium hydrides. In this work, the J-integral values for X-specimens with different hydride orientations are measured to evaluate the fracture toughness of Zircaloy-4 (Zry-4) cladding. The toughness values for Zry-4 cladding with various percentages of radial hydrides are much smaller than those with circumferential hydrides only in the same hydrogen content level at 25 °C. The fractograghic features reveal that the crack path is influenced by the orientation of zirconium hydride. Moreover, the fracture toughness measurements for X-specimens at 300 °C are not sensitive to a variation in hydride orientation but to hydrogen concentration.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

Post-irradiation examinations focused on fuel integrity of spent BWR-MOX and PWR-UO2 fuels stored for 20 years

Post-irradiation examinations (PIEs) of spent BWR-MOX and PWR-UO₂ fuel rods irradiated in commercial LWRs and stored for 20 years were carried out to evaluate fuel integrity during storage. Average burn-up of five fuel rods of the BWR-MOX fuel was about 20 MWd/kgHM and that of the PWR-UO₂ fuel was 58 MWd/kgHM. The PIE items included: (a) visual inspection of the cladding surface, (b) puncture test, (c) ceramographic observation on the pellet and cladding, (d) pellet density, (e) electron probe microanalysis of the pellet, (f) cladding tensile test, (g) hydrogen content and hydride orientation in the cladding, and (h) hydrogen redistribution in the cladding under temperature gradient. The PIE results showed no marked difference in the visual inspection, fission gas release, oxide layer thickness, pellet microstructure, and cladding mechanical properties or hydride orientation after storage. The result of the hydrogen redistribution experiment showed that hydrogen migration had little effect on the fuel integrity during dry storage. Hydrogen migration on the fuel rod for 40 years of storage was evaluated using the heat of transport obtained in the hydrogen redistribution experiment and calculated result showed that hydrogen migration had little effect on the fuel integrity during dry storage.     

Rupture of spent fuel Zircaloy cladding in dry storage due to delayed hydride cracking

Delayed hydride cracking in the Zircaloy alloy has been considered as a possible degradation mechanism of spent nuclear fuel cladding in interim dry storage. Some recent in-core fuel failures indicated that a long axial crack developed in the cladding was a secondary failure by delayed hydride cracking. The aim of this study is to define the effects of hydride reorientation on the failure of Zircaloy cladding. Different hydride orientations, the amount of zirconium hydride and various cracking types, all have been considered for their effects on the crack growth and stability of the cladding, and have been thoroughly discussed in this paper. A finite element computer code, ANSYS, has been used in conjunction with the strain energy density theory. In summary, the crack propagation will be aggravated if the hydride orientation is shifted from the circumferential to the radial direction. For a larger crack length, the zirconium hydride plays an important role in affecting the crack growth because the strain energy density factor increases as the hydride approaches the crack tip. Furthermore, when thermal effects are considered, a compressive stress exists at the inner side of the cladding, while a tensile stress is found at the outer side of cladding, thus resulting in crack propagation from the outer side to the inner side of the cladding. These findings are in accordance with other experimental results in related literature.     

Spent fuel transport cask thermal evaluation under normal and accident conditions

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario.The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers.Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself.In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the ‘80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are discussed.     

Effect of irradiation on the impact and seismic response of a spent fuel storage and transport cask

The spent fuel storage and transport cask must withstand various accident conditions such as fire, free drop and puncture in accordance with the requirement of the IAEA and domestic regulations. The spent fuel storage and transport cask should maintain the structural safety not to release radioactive material in any condition. And also the effects of the irradiation should be considered because the spent fuels stored in the cask for a long time and be possible to change the mechanical properties of the cask.In this study, the changed mechanical properties of the cask after irradiation for the 30 years storage periods are assumed and applied to the impact analysis using ABAQUS/Explicit code and seismic analysis using ANSYS code. The stress intensity on each part of the cask is calculated and the effects of irradiation are studied and structural integrity of the package is evaluated.     

Buckling of fuel rods under inertia loading

Abstract

The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations for a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or that participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding, the higher the inertia loads on the cladding, and, therefore, the lower the 'g' value at which buckling occurs. However, these solutions do not consider displacement compatibility between the fuel and the cladding during the buckling process. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading.     

Influence of Hydride Re-orientation on BWR Cladding Rupture under Accidental Conditions

Hydride precipitation along the radial-axial plane increases in high burn-up boiling water reactor (BWR) fuel claddings. The radially-oriented hydrides may have an important role during fuel behavior in a reactivity-initiated accident and may reduce ductility of the cladding under pellet-cladding mechanical interaction (PCMI) conditions. In order to promote a better understanding of the influence of the radial hydrides on cladding failure behavior under the PCMI conditions, tube burst tests were conducted for unirradiated BWR claddings charged with 200 to 650 ppm of hydrogen. About 20 to 30% of hydrides were re-oriented and precipitated along the radial-axial plane. The claddings exhibited large rupture openings with an axial crack at room temperature and 373 K. The crack penetrated through cladding wall preferentially along the radial hydrides, and radial cross section showed cladding failure in a brittle manner. However, reduction in residual hoop strain by precipitation of the radial hydrides was very small. It is accordingly expected that ductility of high burn-up BWR cladding is significantly reduced not only by precipitation of radial hydrides as far as hydrogen concentration and radial hydride fraction range in the present study.     

BWR spent fuel transport and storage with TN9/4 and TN24BH casks

Abstract

In 2001 the Swiss nuclear utilities started to store spent fuel in dry metallic dual purpose casks at ZWILAG, the Swiss interim storage facility. BKW FMB Energy Ltd, as the owner of the Mühleberg nuclear power plant, is involved in this process and has selected to store the spent fuel in a new high capacity dual purpose cask, the TN24BH. For the transport Cogema Logistics has developed a new medium size cask, the TN9/4, to replace the NTL9 cask, which has performed numerous shipments of BWR spent fuel in past decades. Licensed by the IAEA 1996, the TN9/4 is a 40 t transport cask, for seven BWR high burnup spent fuel assemblies. The spent fuel assemblies can be transferred to the ZWILAG hot cell in the TN24BH cask. These casks were first used in 2003. Ten TN9/4 shipments were made, and one TN24BH was loaded. After a brief presentation of the operational aspects, the paper will focus on the TN24BH high capacity dual purpose cask and the TN9/4 transport cask and describe in detail their characteristics and possibilities.     

Topics of research and development on concrete cask storage of spent nuclear fuel

This paper addresses topics of research and development (R&D) being challenged for realization of concrete cask storage of spent nuclear fuel in Japan. Comparison between metal cask storage and concrete cask storage is addressed. Background of these R&D and current status of technology on spent fuel storage are described. Need and design concepts of concrete cask storage technology, tests and evaluation of integrity of spent fuel, materials, concrete casks under normal and accident conditions, monitoring technology, etc. are systematically arranged and introduced. Topical problems of these R&D are described.     

Beyond design basis accidents in spent-fuel pools — Generic Issue 82

Nuclear reactor plants include storage facilities for the wet storage of spent-fuel assemblies. The safety function of the spent-fuel pool (SFP) and storage racks is to cool the spent-fuel assemblies and maintain them in a subcritical array during all credible storage conditions and to provide safe means of loading the assemblies into shipping casks.Generic Issue 82 (GI-82) relates to the concern that for a postulated accident sequence that results in the loss of water from a light-water reactor (LWR) spent-fuel storage pool, a Zircaloy cladding fire could occur and propagate to older stored fuel. This issue was identified during hearings concerning SFP reracking amendments in the late 1970s when licensees were starting to use high-density storage racks. High-density racks are used to accommodate the storage of spent fuel in SFPs at reactor sites until such time as the Department of Energy (DOE) repository is available and spent fuel can be removed from the reactor sites. Maintaining a low-density storage configuration for recently discharged spent fuel would reduce the Zircaloy cladding fire probability by an order of magnitude, but at a greater cost for additional onsite storage space.The accident sequences that could result in water loss from the SFP, including beyond design basis earthquakes, various types of seal failures and dropped shipping casks, and the Zircaloy cladding fire issues have been studied by the NRC staff. The results of these studies are provided in NUREG-1353, “Regulatory Analysis for the Resolution of Generic Issue 82, Beyond Design Basis Accidents in Spent-Fuel Pools”. Although these studies conclude that most of the spent-fuel pool risk is derived from beyond design basis earthquakes, this risk is not greater than the risk from core damage accidents due to these beyond design basis earthquakes. Therefore, reducing the risk from spent-fuel pools due to events beyond the safe shutdown earthquake would still leave a comparable risk due to core damage accidents. The risk due to beyond design basis accidents in spent-fuel pools, while not negligible, is sufficiently low that the added cost involved with further risk reduction is not warranted.     

Stress and temperature-dependent hydride reorientation of Zircaloy-4 cladding and its effect on the ductility degradation

A hydride reorientation can deteriorate the mechanical ductility of spent fuel cladding and make it more susceptible to failure. Therefore, an evaluation of the reorientation under dry storage conditions and their effects on the cladding ductility are critical issues in terms of the regulation criteria. In this work, biaxial stress was applied to Zircaloy-4 cladding by pressurizing Ar gas. The study showed that the hydride reorientation can occur at around 60 and 80 MPa at 400 and 300 °C, respectively. The ring compression test at room temperature showed that the ductility decreases with an increase in radial hydride quantity: Fl(45) and radial hydride continuity factor. In addition, a significant hydride reorientation can occur at high temperature conditions even if the hoop stress is equal to or less than 90 MPa which can bring a significant ductility degradation.     

核电厂乏燃料组件厂内转运关键设备研制

乏燃料组件厂内转运是解决核电厂燃料水池贮存空间不足问题的方法之一。本文分析了乏燃料组件厂内转运的设计准则、安全风险,介绍了用于运输容器内破损组件检测和运输容器内组件冷却用设备的工作原理及其应用情况。应用结果表明:破损检测设备可以快速有效地检测乏燃料运输容器内是否存在破损组件;乏燃料组件冷却设备可以较为安全地冷却装有乏燃料组件的运输容器。      

Spent Fuel Transport Associated With Other Dangerous Goods in Regular Train Units — Assessment of Hypothetical Explosion Impacts

Abstract

Continental railway transport regulations (RID) do not exclude the transport of spent fuel casks in a regular train unit that also contains wagons with other hazardous materials. In the case of a train accident the release or reactions of those dangerous goods could potentially give significant accidental impacts on to the spent fuel casks. The assessment of fires from inflammable liquids and the explosion impacts from pressurised inflammable gases (like LPG) is well known from other studies which have usually revealed sufficient safety margins to the robust spent fuel cask designs. A new problem to be assessed is the potential impact from a detonation blast wave from explosives transported in the same train unit as a spent fuel cask. BAM is assessing this problem by developing a numerical model to calculate the effect of the dynamic pressure of a external shockwave on the cask construction. The calculation results show that the integrity of a robust monolithic cask with a screwed lid closure system is preserved after the effect of a 21 tonne (equivalent weight of TNT) explosive detonation in the regular transport configuration with a distance of 25 m between the centre of the explosion and the front of the cask.     

Finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel storage cask

A spent fuel storage cask is required to prove the safety of its canister under a hypothetical accidental drop condition which means that the canister is assumed to be free dropped on to a pad of the storage cask during the loading of the canister into a storage cask. In this paper, finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel dry storage cask were carried out to improve the structural integrity of the canister under a hypothetical accidental drop condition. The pad of the storage cask was originally designed as cylindrical steel structure filled with concrete. The pad was modified by using the structure composed of steel and polyurethane-foam instead of the quarter of the upper concrete as an impact limiter. The effects of the shape and the thickness of the steel structure and the density of the polyurethane-foam which was used in between steel structures were studied. As the optimized pad of a spent fuel dry storage cask, the quarter of the upper concrete was replaced with 12 mm thick circular steel structure and polyurethane-foam whose density was 85 kg/m³. The drop tests of a 1/3 scale model for the canister on to the original pad and the optimized pad were conducted. The effect of the pad structure was evaluated from the drop tests. The optimized pad has a greater shock-absorbing effect than the original pad. In order to verify the analysis results, strains and accelerations in the time domain by the analytical methods were compared with those by a test. The numerical method of simulating the free drop test for a dry storage cask was verified and the numerical results were found to be reliable.