Modeling of Jet Pump by RELAP4 Code in Loss-of-Coolant Accident Analysis

Waterlogged fuel rod experiments performed at the NSRR are analyzed using the computer code WTRLGD, which was devised for the analyses of thermo-dynamical behavior of a waterlogged fuel rod. The numerical results are compared with the data from the experiments in order to assess the validity of the computer code. Parameters in the analyses are volumetric fraction of water, reactor period, gap width, a pin hole and the end peaks. Thus the analyses cover almost all the waterlogged fuel rod experiments at the NSRR.

The comparison shows good agreement between the experimental results and numerical ones on the transient thermo-dynamical behaviors of fuel, such as, rod internal pressure, cladding surface temperature and cladding strain. The numerical results also quantitatively agree with the experimental data concerning the effects of the above parameters on failure threshold energy. From the above findings, the computer code is assessed to be valid enough for the analyses of the failure behavior of the waterlogged fuel rod under a reactivity initiated accident condition.     

Effects of Rod Pre-Pressurization on Light Water Reactor Fuel Behavior during Reactivity Initiated Accident Conditions

Previously pressurized (pre-pressurized) fuel rod tests recently performed in the Nuclear Safety Research Reactor (NSRR) investigate the effects of initial internal pressure on fuel rod behavior during reactivity initiated accident (RIA) conditions. A single PWR type fuel rod was contained within a waterfilled, ambient temperature and ambient pressure capsule. The fuel rod was then heated by the pulsing operation of the NSRR.

Results from the tests show that the effect of pre-pressurization was significant for the fuel rods with initial internal pressure of 0.8 MPa and above, and fuel rod failure occurred from rupture of the cladding with lower threshold energy deposition for failure as the initial internal pressure was increased. The cladding rupture was governed mainly by the cladding temperature rise, not by the rod internal pressure rise during the transient. The relationships between cladding burst pressure and cladding burst temperature and between cladding strain and cladding temperature at cladding rupture obtained in the present study under an RIA condition agree with the results obtained from various in- and ex-reactor experiments under a LOCA condition, although the obtained time-averaged strain rate of the Zircaloy cladding was much greater than that in a LOCA condition.     

FARST — A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:

1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.

2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.

3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.

4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.

The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR).The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant.     

Modification of thermal hydraulic code,THEATRe, with fuel rod behavior models and its application to reactivity-initiated accident (RIA)

During reactor operation, many complex changes occur in fuel rod which affects its thermal, mechanical and material properties. These changes also affect the reactor response to the transient and accident situations. Realistic simulation of fuel rod behavior under transients such as reactivity-initiated accident (RIA) is of great significance. In this study, thermal hydraulic analysis code THEATRe (Thermal Hydraulic Engineering Analysis Tool in Real-time) has been modified by addition of fuel rod behavior models for dynamic simulation of nuclear reactor. Transient changes in gas-gap parameters were taken into account by modeling the gas-gap behavior. Thermo-mechanical behavior of fuel rod is modeled to take into account the thermal, elastic and plastic deformation. To simulate RIA, point reactor kinetics model is also incorporated in the THEATRe code. To demonstrate the transient fuel rod behavior, AP1000 reactor is modeled and three hypothetical RIA cases are simulated. The RIA is considered at three different reactor power levels, i.e. 100, 50 and 1% of nominal power. The investigated parameters are fuel temperature, cladding stress and strain, fuel and cladding thermal conductivity and heat transfer coefficient in gas-gap. Modified code calculates the fuel rod temperatures according to updated fuel, clad and gas-gap parameters at the onset of steady-state operation and during the transient. The modified code provides lower steady-state fuel temperature as compared to the original code. Stress and strain analyses indicate that the hoop and radial strain is higher at high power locations of the fuel rod; therefore, gap closure process will initially occur in the central portion of the fuel rod and it should be given more emphasis in the safety analysis of the fuel rod and nuclear reactor during accidents and transients.     

基于MOOSE框架的五方程两相流分析程序开发

基于多物理场耦合框架MOOSE,采用五方程两相流模型开发了模块化程序ZEBRA,实现了高阶时间、空间离散格式两相流动传热问题的求解。采用Bartolomei开展的垂直圆管过冷沸腾实验对ZEBRA进行验证,在不同热流密度、质量流密度、压力工况下,将程序计算值与实验值进行了数值验证和计算分析。结果表明：ZEBRA中五方程模型预测值与实验值符合良好,沸腾起始点和空泡份额的预测合理,表明ZEBRA初步具备了处理两相流问题的能力。     

Principle of rationalizing the criteria for abnormal transients of the Super LWR with fuel rod analyses

A detailed fuel rod design is carried out for the first time in the development of Supercritical-pressure Light Water Reactor (Super LWR). The fuel rod design is similar to that of LWR, consisting of UO₂ pellets, a gas plenum and a Stainless Steel Cladding. The principle of rationalizing the criteria for abnormal transients of the Super LWR is developed. The fuel rod integrities can be assured by preventing plastic strains on the cladding, preventing the cladding buckling collapse, and keeping the pellet centerline temperature below its melting point. The FEMAXI-6 fuel analysis code is used to evaluate the fuel rod integrities in abnormal transient conditions. Detailed analyses have shown that allowable limits to the maximum fuel rod power and maximum cladding temperature can be determined to assure the fuel integrities. These limits may be useful in the plant safety analyses to confirm the fuel integrities during abnormal transients.     

燃料棒破损热力学模块开发及在严重事故分析程序中的应用

包壳肿胀和破损是严重事故早期阶段的重要现象。包壳形变不仅会造成局部流动堵塞,同时,水蒸气会从破裂处进入包壳气隙,增加包壳被蒸汽氧化的表面积。广泛使用的一体化严重事故分析程序不能分析早期事故过程中燃料棒的热力学行为,判断包壳破裂也只是基于简单的参数模型。本文开发了分析燃料棒热力学行为的FRTMB模块,并集成在严重事故分析程序ISAA中。使用开发的耦合系统ISAA FRTMB分析了CAP1400反应堆直接注射（DVI）管线小破口事故过程中燃料棒的热力学行为,并预计了包壳破裂时间及相应的失效温度。计算结果整体验证了ISAA FRTMB分析瞬态事故过程中燃料棒热力学行为以及判断包壳破裂的适用性和可靠性。     

Unbiasedness Evaluation of Statistical Error in Staged Monte Carlo Calculation

Fuel rod failure behavior has been studied under a reactivity initiated accident condition in Nuclear Safety Research Reactor (NSRR), JAERI. In the studies, inetallurgical observations showed that the incipient fuel rod failure mode was oxygen-induced embrittlement of the cladding independent of the test conditions such as fuel designs and cooling environments except for pressurized and waterlogged fuels. Development of the oxidation layers and embrittlement of β-Zry were quantitatively evaluated through the metallurgical examinations. A diffusion equation of oxygen was solved under a finite system with moving boundary conditions to obtain the oxygen concentration and evaluate the cladding embrittlement. The calculation showed that the wall thinning due to the cladding melt is needed for the complete embrittlement because the wall thinning enhances the oxygen concentration in the β-Zry, which well explain the experimental results. Therefore the failure threshold energy is determined by the cladding melting temperature. The failure threshold derived from this study is expected to be applicable to predicting the fuel rod failure behavior in computer analyses and also useful to evaluate the failure threshold energy for the new types of fuel rod.     

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO₂ fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail.     

Transient failure of zircaloy cladding

A state of the art study of transient failure of zircaloy reveals that a wide range of failure models exist, and their use in fuel analysis depends upon the type of application. Licensing analyses, for example, employ highly simplified criteria that depict cladding rupture in terms of a single response variable, usually the nominal circumferential stress, as function of temperature; while fuel behavior analyses use more detailed models with varying levels of sophistication. Some of these models, however, are ill-suited for detailed fuel rod analysis computer codes because of their dependence on case-specific parameters; others are limited in their utility because of inherent deficiencies in their theoretical treatment of the phenomena involved. Based on this study, a new zircaloy transient failure model is developed that takes account of the important variables involved in transient phenomena, namely, heating rate, strain rate, temperature, true stress, true strain, and prior condition of the cladding material. The model is particularly suited for detailed fuel rod analysis and thus was implemented in the FREY code.     

Transient Elongation of a Fresh Fuel Rod under Reactivity Initiated Accident Conditions

In-pile experiments of fresh fuel rods under reactivity initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor at the Japan Atomic Energy Research Institute in order to understand the basic pellet cladding mechanical interaction (PCMI) behavior. Rapid fuel pellet expansion due to a power excursion would cause radial and longitudinal deformation of the cladding. This PCMI could be one of the possible incipient failure modes of an embrittled cladding of a high burnup fuel under the RIA conditions.

Basic PCMI behavior was studied by measuring cladding deformation of a fresh fuel rod without complicated irradiation effects. The transient elongation measurements of the fuel with two kinds of gap width indicated not only PCMI-induced cladding elongation, but also reduction of the pellet stack displacement by the cladding constraint. In the tests under a high-pressure and high-temperature condition simulating an operation condition of BWRs, additional ridge-type cladding deformation was generated due to the axial collapse of the cladding. A preliminary analysis for interpretation of the tests was made using a computer code for the transient analysis of fuel rods, FRAP-T6.     

A simple forced diversion model for study of thermal-hydraulic transients in LMFBR subassembly

The generalized simple, transient, integral energy balances based on the average properties for the fuel and cladding have been used in our new multichannel thermal-hydraulic model for calculating the transient behavior of coolant in the rod bundle. This model was developed to provide a simple useful tool for analyzing the flow and thermal transients in a rod bundle with reasonable accuracy, and to understand the fundamental characteristics of flow in the rod bundle under both normal and abnormal condition of reactor-core operation.     

Evaluation of LWR fuel rod behavior under operational transient conditions

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding.The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8 × 8 RJ fuel rod temperatures under power ramp conditions.     

A model for predicting pellet-cladding interaction-induced fuel rod failure

A model for predicting pellet-cladding mechanical-interaction-induced fuel rod failure is presented. Cladding failure is predicted by explicitly modelling the formation and propagation of radial cladding cracks by the use of non-linear fracture mechanics concepts in a finite element computational framework. The failure model is intended for implementation in finite element fuel performance codes in which local pellet-clad interaction is modelled. Crack initiation is supposed to take place at pre-existing cladding flaws, the size of which is estimated by simple probabilistic concepts, and the subsequent crack propagation is assumed to be due to either iodine-induced stress corrosion cracking or ductile fracture. The novelty of the outlined approach is that the development of cladding cracks which may ultimately lead to fuel rod failure can be treated as a dynamic and time-dependent process. The influence of complex or cyclic loading, ramp rates and material creep on the failure mechanism can thereby be investigated. The presented failure model has been incorporated in the ABB Atom transient fuel performance code. Numerical results from some applications of the code are used to illustrate the usefulness of the model.     

FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.     

A methodology for the evaluation of fuel rod failures under accident conditions

Abstract

Recent studies on the long-term behaviour of high-burnup spent fuel have shown that, under normal conditions of storage, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride cracking, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar regulatory rules have not yet been developed to address failures of fuel rod cladding that could potentially lead to reconfigured fuel geometry under hypothetical transport accidents. At issue is the effect on cladding ductility of potential changes in zirconium hydride morphology during dry storage. Recent studies have shown that above a certain level of cladding hoop stress, the decaying temperature history during dry storage can cause the hydrogen in solid solution to precipitate in the form of radial hydrides, which, depending on their relative concentration, can induce brittle failures in the cladding. From a US regulatory perspective such cladding failures, if they were to cause fuel reconfiguration, could invalidate the cask's criticality and shielding licensing analyses, which are based on coherent geometry. This paper describes a methodology for high-burnup spent fuel to determine the frequency of cladding failure and failure modes under drop accidents, considering end-of-storage spent fuel conditions. The degree to which spent fuel reconfiguration could occur during handling or transport accidents would depend to a large extent on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there are no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, this paper focuses on the development of a methodology for modelling and analysis that deals with this general problem on a generic basis. First, consideration is given to defining accident loading that is equivalent to the bounding hypothetical transport accident of a 9 m drop onto an essentially unyielding surface. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A model of material behaviour, with embedded failure criteria, for cladding containing various concentrations of circumferentially and radially oriented hydrides has been developed and implemented in a finite-element code. The hydride precipitation model, which describes the hydride structure of the cladding at the end of dry storage, and the hydride-dependent properties of high-burnup fuel cladding form the main input to the constitutive model. The third element in the overall process is to utilise this material model and its host finite-element code in the structural analysis of a transport cask subjected to bounding accident loading to calculate fuel rod failures and failure mode configurations. This requires detailed modelling of the transport cask and its internal structure, which includes the canister, basket, fuel assembly grids and fuel rods. The overall methodology is described.     

Candu type fuel behavior during rapid overpower transients

Results obtained in the pulse irradiation tests performed on segmented fuel elements in the Romanian Annular Core Pulse Reactor (ACPR) are discussed below. Tests included the effects of initial element internal pressure and a wide range of energy deposition on the fuel element behavior. All tests were conducted in stagnant water at room temperature and atmospheric pressure inside the capsule. The fuel elements were instrumented with thermocouples for cladding surface temperature measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during the tests. The fuel elements were subjected to total energy depositions from 70 to 265 cal g⁻¹ UO₂. Cladding failure mechanism and the failure threshold have been established. The fuel failure mechanism is a burst type and is very similar to LOCA failure mechanism even though the rate energy deposition is higher in the ACPR tests. At higher energy deposition brittle cladding fracture near endcap weld region can be produced. The failure threshold is situated between 190 and 200 cal g⁻¹ UO₂ for standard fuel rod (0.2–0.3 MPa internal pressure) and less than 160 cal g⁻¹ UO₂ for pressurized fuel rods (internal pressure between 1 and 3.0 MPa). Pre-pressurization could be an important factor to control the failure threshold energy. The experimental program is still in progress and new experiments are foreseen to be performed in the following period.     

Analysis of mechanical load on cladding induced by fuel swelling during power ramp in high burn-up rod by fuel performance code FEMAXI-6

Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod was analyzed by a fuel performance code FEMAXI-6. The code was developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using finite element method (FEM).During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed the level that is predicted by a “steady-rate” swelling model, causing a large circumferential strain in cladding. This phenomenon was simulated by a new swelling model to take into account the fission gas bubble growth. As a result it was found that the new model can give reasonable predictions on cladding diameter expansion in comparison with PIE data. The bubble growth model assumes that the equilibrium state equation holds for a bubble under external pressure, and simultaneous solution is obtained with both bubble size determination equation and diffusion equation of fission gas atoms. In addition, a pellet-clad bonding model which has been incorporated in the code to assume solid mechanical coupling between pellet outer surface and cladding inner surface predicted the generation of bi-axial stress state in the cladding during ramp.