Analysis on split failure of cladding of high burnup BWR rods in reactivity-initiated accident conditions by RANNS code

A computer code RANNS was developed to analyze fuel rod behaviors in the reactivity-initiated accident (RIA) conditions. RANNS performs thermal and finite-element mechanical calculation for a single rod in axis-symmetric geometry, where fuel pellet consists of 36 equal-volume ring elements and cladding metallic wall consists of eight equal-thickness ring elements and one outer oxide element. The code can calculate temperature profile inside the rod, contact pressure generated by pellet–clad mechanical interaction (PCMI), stress–strain distribution and their interactions elaborately. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by the fuel performance code FEMAXI-6.In the present study, analysis was performed on the simulated RIA experiments in the “nuclear safety research reactor” (NSRR), FK-10 and FK-12, with high burnup BWR rods in a cold-start up condition, and stress–strain evolution in the PCMI process was calculated extensively. In the analysis, the pellet–clad bonding was assumed both in the heat conduction and in mechanical restraint. The calculated hoop strain increase was compared with the measured strain gauge data, and satisfactory agreement was obtained. Simulation calculations with broader power pulses anticipated in RIA of commercial BWR were carried out and the resulted cladding hoop stress was compared with the failure stress estimated by comparison of analysis with experimental data.     

Thresholds for failure of high-burnup LWR fuels by Pellet Cladding mechanical interaction under reactivity-initiated accident conditions

ABSTRACT

To contribute to the future updating on the Japanese safety criteria for pellet/cladding mechanical interaction (PCMI) failure of light water reactor fuels under reactivity-initiated accident (RIA) conditions, this paper summarizes the recent important outcomes from research programs with the Nuclear Safety Research Reactor (NSRR). Applicability of current criteria, which are defined as a function of fuel burnup and possibility of introducing another parameter for new criteria were evaluated based on the results of the RIA-simulated pulse irradiation tests, post-test examinations, and supporting analytical work, such as the reevaluation of fuel enthalpies in earlier NSRR experiments. Failure-threshold curves based on cladding hydrogen content as a primary measure of fuel degradation have been proposed as a possible alternative that can be used to judge the occurrence of PCMI failure to ensure conservativeness in a more pertinent manner.     

Stress biaxiality in high-burnup PWR fuel cladding under reactivity-initiated accident conditions

In order to promote a better understanding of failure mechanisms of high-burnup pressurized water reactor (PWR) fuels under reactivity-initiated accident (RIA) conditions, stress biaxiality in cladding has been estimated for the pellet-cladding (PC) mechanical interaction (PCMI) phase. The estimation was based on an analysis of the transient elongations of a pellet stack and a cladding tube measured in RIA-simulating experiments in the nuclear safety research reactor (NSRR) using the RANNS code. Stress biaxiality in the high-burnup PWR fuel cladding during the PCMI phase has been estimated to be 0.7–0.8, on average, at the mid-wall of the cladding. A comparison with fresh fuel test results and a sensitivity analysis showed that the effects of burnup and pulse width on cladding stress biaxiality are less than 10% for the investigated range. The present analysis also indicated that PC friction is strong, and that the cladding constraint on pellet stack elongation is significant irrespective of burnup. Therefore, it is recommended that strong PC friction be assumed, which is similar to the mechanical bonding condition, and that fuel pellets be treated as deformable materials in models of fuel behavior during the PCMI phase.     

SMART behavior under over-pressurizing accident conditions

SMART (System-integrated Modular Advanced ReacTor) is an integral reactor of 330 MW capacity with passive safety features under development in Korea. The design is developed by combining the firmly-established commercial reactor technologies with new and advanced technologies such as industry proven KOFA (Korea Optimized Fuel Assembly) based nuclear fuels, self-pressurizing pressurizer, helically coiled once-through steam generators, and new control concepts. The design of SMART focuses on enhancing the safety and reliability of the reactor by employing inherent safety features such as low core power density, elimination of large break loss of coolant accident, etc. In addition, in order to prevent the progression of emergency situations into accidents, the SMART is provided with a number of engineered safety features such as Passive Residual Heat Removal System, Passive Emergency Core Cooling System, Safeguard Vessel, and Passive Containment Over-Pressure Protection System. This paper presents an overview of the SMART design, characteristics of it’s safety systems, and results of over-pressure accident analyses. The results of the accident analyses show that the SMART provides the inherent over-pressure protection capability for design basis accidents without actuation of any protection devices such as safety valves, rupture disks, etc.     

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.     

Failure strains and proposed limit strains for an reactor pressure vessel under severe accident conditions

The local failure strains of essential design elements of a reactor vessel are investigated. The size influence of the structure is of special interest. Typical severe accident conditions including elevated temperatures and dynamic loads are considered.The main part of work consists of test families with specimens under uniaxial and biaxial load. Within one test family the specimen geometry and the load conditions are similar, but the size is varied up to reactor dimensions. Special attention is given to geometries with a hole or a notch causing non-uniform stress and strain distributions typical for the reactor vessel. A key problem is to determine the local failure strain. Here suitable methods had to be developed including the so-called “vanishing gap method”, and the “forging die method”. They are based on post-test geometrical measurements of the fracture surfaces and reconstructions of the related strain fields using finite element models.The results indicate that stresses versus dimensionless deformations are approximately size independent up to failure for specimens of similar geometry under similar load conditions. Local failure strains could be determined. The values are rather high and size dependent. Statistical evaluation allow the proposal of limit strains which are also size dependent. If these limit strains are not exceeded, the structures will not fracture.     

Determination of enrichment of recycle uranium fuels for different burnup values

Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000–50,000 MWd/tonU.     

Simulation of the fracture behavior of Zircaloy-4 cladding under reactivity-initiated accident conditions with a damage mechanics model combined with fuel performance codes FEMAXI-7 and RANNS

A continuum damage mechanics model using FEM calculations was proposed to be applied to an analysis of the fuel failure due to pellet cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions. The model expressed ductile fracture via two processes: damage nucleation related to void nucleation and damage evolution related to void growth and linkage. The boundary conditions for the simulations were input from the fuel performance codes FEMAXI-7 and RANNS. The simulation made reasonable predictions for the cladding hoop strain at failure and reproduced the typical fracture behavior of the fuel cladding under the PCMI loading, characterized by a ductile shear zone in the inner region of the cladding wall. It was shown that occurrence of a through-wall crack is determined at an early stage of crack propagation, and the rest of the through-wall penetration process is achieved with a negligible increment in strain. The effect of a local temperature rise in the cladding inner region on the failure strain was found to be less than 5% for the conditions investigated. Failure strains predicted under a plane strain loading were smaller by 20%–30% than those predicted under equibiaxial tensions between the hoop and the axial directions.     

HTGR plant dynamics analysis under accident conditions

Important features of high temperature gas-cooled reactor (HTGR) systems related to plant dynamics and accident analysis are discussed. Because of the basic simplicity of the HTGR system, it is possible to analyze the full reactor plant (core, helium circulators, steam generators and reheaters, feedwater controls, turbine controls, and plant protective action) in a single computer code. Representative dynamics analysis is presented for the Fort St. Vrain Power Station.     

Creep assessment of Zry-4 cladded high burnup fuel under dry storage

Cladding creep rupture is thought to be the most likely and limiting failure mechanism of spent fuel in dry storage. In spite of being highly unlikely, the current trend towards high burnups is drawing further attention to the potential creep effect on cladding integrity of fuels burnt over 45 GWd/tU.This paper explores the burnup influence on cladding creep during dry storage by modelling two different high burnup scenarios (51 GWd/tU and 67 GWd/tU). In addition, sensitivity of the results to the in-reactor average power and power history has been conducted. The computation tool used in this study has been an extension of FRAPCON-3.4 capable of simulating dry storage scenarios. Burnup and average linear power have been shown to make creep grow quite substantially during the first two years in dry storage, adopting a quasi-asymptotic trend from then on. However, even though this profile seems to have a generic nature, the net creep value reached depends not just on integral and average variables, but also on magnitudes describing the entire irradiation history, like linear power history. In none of the cases explored creep approaches the 1% threshold. In-reactor FGR modelling has been highlighted as a key element to get accurate estimates of creep.     

Failure assessment of RPV nozzle under loss of coolant accident

In case of a postulated loss of coolant accident (LOCA) of a reactor pressure vessel (RPV), the nozzle region experiences higher stresses and lower temperatures than the remaining part of the RPV. Thus, the nozzle is to be considered in the RPV safety assessment. For a LOCA event, three-dimensional elastic–plastic finite element calculations of stresses and strains in the intact RPV were performed. Using the substructure technique, fracture mechanics analyses were then carried out for several postulated cracks in the nozzle corner and in the circumferential weld below the nozzle. For different crack geometries and locations, the J-integral and the stress intensity factor were calculated as functions of the crack tip temperature. Based on the K_IC-reference curve and the J_R curve, both brittle and ductile instability of the postulated cracks were excluded. In order to reduce the expenses of three-dimensional finite element analyses for various crack geometries, an analytical procedure for calculating stress intensity factors of subclad cracks in cylindrical components was extended for cracks in the nozzle corner.     

Rock-like oxide fuel behavior under reactivity initiated accident conditions

Pulse irradiation tests of two types of rock-like oxide (ROX) fuel, i.e. yttria stabilized zirconia (YSZ) and YSZ/Spinel composite, were conducted in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under reactivity-initiated accident conditions. The ROX fuels failed with cladding burst at fuel volumetric enthalpies above 10 GJ m⁻³, which was comparable to that of UO₂ fuel. The failure of the ROX fuels, however, occurred with considerable fuel melting and was quite different to that of UO₂ fuel, which was caused by cladding melting and embrittlement due to heavy oxidation. Lower fuel melting temperature of the ROX fuels compared to that of UO₂ contributed to the different fuel failure modes. Certain amount of molten ROX fuel dispersed out at the failure. However, the mechanical energy generation due to the molten fuel/water interaction was negligible for the ROX fuels at peak fuel enthalpies below 12 GJ m⁻³.     

Iodine-steel reactions under severe accident conditions in light-water reactors

Owing to large surface areas, the reaction of volatile molecular iodine (I₂) with steel surfaces in the containment may play an important role in predicting the source term to the environment. Both wall retention of iodine and conversion of volatile into non-volatile iodine compounds at steel surfaces have to be considered. Two types of laboratory experiment were carried out at Siemens (KWU) in order to investigate the reaction of I₂ at steel surfaces representative for German power plants.

• 1. 

  (1) For steel coupons submerged in an I2 solution at T = 50, 90 or 140 °C the reaction rate of the I2−I− conversion was determined. No iodine loading was observed on the steel in the aqueous phase tests. I2 reacts with the steel components (Fe, Cr or Ni) to form metal iodides on the surface which are all immediately dissolved in water under dissociation into the metal and the iodide ions. From these experiments, the I2−I− conversion rate constants over the temperature range 50–140 °C as well as the activation energy were determined. The measured data are suitable to be included in severe accident iodine codes such as IMPAIR.

• 2. 

  (2) Steel tubes were exposed to a steam-I2 flow under dry air at T = 120 °C and steam-condensing conditions at T = 120 and 160 °C. In dry air, I2, was retained on the steel surface and a deposition rate constant was measured. Under steam-condensing conditions there is an effective conversion of volatile I2 to non-volatile I− which is subsequently washed off from the steel surface. The I2−T− conversion rate constants suitable for modelling this process were determined. No temperature dependence was found in the range 120–160 °C.

     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

Failure characteristics of cladding tubes under RIA conditions

In the frame of actions for improving the safety of its nuclear power plants, Electricité de France needs to build the mechanical criteria ensuring the clad integrity for several operating conditions.This paper presents analytical mechanical models used to derive failure criteria for reactivity insertion accidents (RIA) and interpretation of the CABRI REP-Na experimental tests.Building analytical criteria requires an experimental database. Mechanical tests performed on non-irradiated and irradiated cladding tubes have been provided from French and international programmes (PROMETRA, EPRI, …). These tests consist of tube burst and axial tension, and ring tension. Several strain biaxiality ratios are thus available: pure circumferential tension (from ring tension), pure axial tension (from tube axial tension), and plane strain conditions (from tube burst tests). Several strain rates, temperatures, irradiation conditions are also available.The major feature of our study has been to make it possible that these several thermomechanical conditions be representative of “standard” RIA loading conditions.To this aim, we have derived some biaxiality and strain rate corrections to be applied to the results of experimental tests, in such a way that they could be representative of RIA biaxiality conditions (which are assumed to be strain equibiaxiality), and also representative of RIA strain rate conditions (which are assumed to be 5 s⁻¹).The corrections that we derive are based on the fracture properties of hydrided zirconium alloys (especially in terms of anisotropy), and also on an assumed form of the material constitutive equations.Each test of the “homogenized” database has thus been used to calculate a strain energy density, representative of its fracture (the strain energy density is defined as the integral of the stress times strain rate states, over the duration of the mechanical test). The SED values are plotted against the sample's oxide thickness, and a lower bound limit can be established, with respect to oxide thickness.In order to address the problem of representativeness of the laboratory database, an experimental set-up has been developed that aims at characterizing the failure behavior of cladding tubes under RIA conditions. The developed experimental set-up is based on electromagnetic forming.The development of the test and in particular of the die is delicate but leads to repeatable results with a controlled strain biaxiality ratio higher than those obtained through conventional tests such as ring tests or PSU ring tests. The use of electromagnetic forming process allows testing the specimen with very high strain rates. For the next test series other zircaloy alloys at the reception state and in hydrided conditions will be tested in order to look at hydrides influence on fracture strains.A first finite element simulation of the test was engaged. The simulation and the experimental results are in quite good agreement. In future, the consistency of the previously developed analytical mechanical criterion with the electromagnetic forming experimental results will be verified.     

Comparison of UF6 packages under certification test and accident conditions

Abstract

Currently there are three packages approved by the NRC for US domestic shipments of fissile quantities of UF₆: NCI-21PF-1, UX-30, and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR part 71. The primary objective of this project was to compare conditions experienced during these tests to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR part 71 tests was achieved by means of computer modelling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from tests and other fire scenarios. In addition, the likelihood of encountering bodies of water during transport over representative truck routes was assessed. Modelled effects and their associated probabilities, accident rates, and other characteristics gathered from representative routes were combined with existing event tree data to derive generalized probabilities of encountering accident conditions comparable to or exceeding the 10 CFR part 71 test conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents.     

Inelastic analysis of cylindrical steel containment vessels under internal accident conditions

The present paper is concerned with the structural safety assessment of a proposed nuclear steel containment shell during a postulated loss-of-coolant accident scenario. The structural evaluation is performed using a computational second-order refined plastic-hinge method, which is capable of accurately predicting all possible modes of failure in an efficient and computationally less expensive way than the general FEM formulation. A tangent modulus model and a gradual reduction of the inelastic resistance surface are used to take into account directly the structural strength and stability performances in the element formulation. The implemented numerical method provides more reliable safety margins and maintainability, exhibiting a more uniform structural safety level than the linear elastic analysis. A simplified non-linear heat transfer model, developed for symmetrical cross-sections, is used to determine the steel temperature gradient and to establish a link between the thermo and the mechanical analysis. The load resulting from pressure and temperature thermodynamic calculations, obtained for the accident scenario, are considered in the structural quasi–static analysis, so that the structural response can be tracked for the entire duration of the simulated accident.     

Thermal analysis of NAC-LWT cask under normal and fire accident conditions

Abstract

Two- and three-dimensional thermal models of a Nuclear Assurance Corporation Legal Weight Truck (NAC-LWT) cask were constructed using the PATRAN commercial finite element package. The two-dimensional model included the effect of radial stiffeners in the package’s external neutron shield but the three-dimensional model did not. A normal conditions of transport (NCT) simulation using both models predicted the peak cladding temperature was roughly 210°C. The NCT package temperatures were used as initial conditions for transient fire/post-fire simulations. Different assumptions were used to determine when the neutron shield liquid drained from the tank and was replaced by air. When the liquid was assumed to remain within the tank during and after the fire, the peak cladding temperature was predicted to exhibit a temporal maximum of roughly 300°C, ～6 h after the end of the fire. If the liquid drained from the tank during the fire, the cladding temperature did not exhibit a temporal peak. Rather, it eventually reached a maximum temperature of roughly 280°C, which is the steady state NCT peak temperature when air is in the neutron shield tank. This undergraduate project will be used to lay down a foundation for further research on NAC-LWT casks. Two and three dimension package of the cask will be constructed using ANSYS, and simulations will be run for NCT and fire/post-fire conditions. The models will also be linked to Container Analysis Fire Environment (CAFE) to predict response of the package in fire.