Effect of Fuel Pellet Eccentricity with Cladding on Fuel Rod Thermal Behavior under Reactivity Initiated Accident Condition

Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g?UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum temperature difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g?UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA conditions.     

An overview of reactivity initiated accident behavior of rock-like fueled pressurized water reactors

Reactivity initiated accident (RIA) analyses of plutonium rock-like oxide (ROX) fueled PWRs have been carried out with the point kinetics calculations. As a result, the analyses have shown a very severe transient behavior of the ROX fueled PWR, which is unacceptable without any improvement. It was also found that the RIA behavior of ROX fueled PWRs can be improved by increasing the negative fuel temperature coefficient (_f). For this improvement, the additives in the ROX fuel such as UO₂ and ThO₂ were considered, as well as a ROX assembly partial loading UO₂ core. With UO₂ additive, it was successful to have satisfying _f and RIA behavior of ROX fuel core, while the partial loading core must be further improved. Besides the ROX-PWR RIA analytical study, the actual behavior of the ROX fuel pin under RIA condition has been experimentally investigated at the Nuclear Safety Research Reactor (NSRR) of JAERI. Though the ROX fuel pin failure mechanism with fuel melting seems quite different from that of UO₂ pin with cladding melting, the ROX pin failure threshold was found to be roughly the same as that of UO₂ in terms of accumulated energy per unit fuel volume.     

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.     

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.     

Characterization of High Burnup Fuel for Safety Related Fuel Testing

Fuel Assemblies designed and fabricated by Westinghouse Electric Sweden (WSE) to reach high burnup have been operated in the Leibstadt nuclear power plant (KKL) for seven cycles attaining an assembly average burnup above 60 MWd/kgU. The irradiation conditions in KKL featured linear heat generation rates ranging from 250 W/cm early in life down to 100W/cm in the last cycle and normal water chemistry with zinc injection. Selected rods have been extracted at both intermediate and final irradiation stages and hot cell examinations have been carried out. The results show that the fuel is well suited for high burnup applications and rod segments have been provided to the OECD Halden Reactor Program, the OECD Studsvik Cladding Integrity Program and the Japanese ALPS program for dedicated high burnup tests with regard to fission as release and cladding lift-off as well as behavior under power transient, RIA and LOCA.     

Experimental Study on the Fuel Behavior during Reactivity Accident at Power Operation Condition

In the Nuclear Safety Research Reactor (NSRR) program of Japan Atomic Energy Research Institute (JAERI), the fuel behavior in reactivity initiated accidents (RIAs) has been studied through irradiation tests with simulated power burst using fresh or preirradiated test fuel rods. In order to investigate possible influence of the difference of initial temperature profile in the fuel pellet on the fuel failure behavior, two tests were conducted with fresh fuel rods for RIAs at power operation using the newly developed NSRR operation mode and the results were compared with the results of previous irradiation tests which were for RIAs at zero power.

In the tests for RIAs at power, the reactivity of 2.0$ or 2.3$ was inserted rapidly after the linear heat rate of the test fuel rod was kept constant at 39kW/m for 5s. It has been shown through this study with fresh fuel rods that the fuel enthalpy of the failure threshold for RIAs at power is the same as that for RIAs at zero power and that the failure mechanism is the same as that of RIA at zero power. It has been clarified that there is no obvious influence of initial temperature distribution on the fuel behavior during RIAs in case of fresh fuels. The evaluation method of fuel enthalpy with which the fuel failure threshold is described was also studied.     

Irradiation Studies of JAERI's Fuel at Halden Reactor

Japan Atomic Energy Research Institute (JAERI) has conducted irradiation experiments utilizing extensive fuel instrumentations at HBWR in Norway since 1967. The major effort has been given to the study of pellet cladding interaction. A special emphasis was placed on the studies of inpile fuel rod deformation including cladding diameter measurements. The experimental conditions cover steady state irradiation, power ramping and power cycling. The data obtained from the experiments were used for development and verification of the computer code FEMAXI. The present paper reviews and summarizes the major results of the experiments.     

Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Agency (JAEA). The paper presents recent results obtained from the NSRR power burst experiments with high burnup fuels, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Results from the recent four experiments on high burnup (about 60 to 78 MWd/kgU) PWR UO₂ rods with advanced cladding alloys showed that the fuel rods with improved corrosion resistance have larger safety margin against the PCMI failure than conventional Zircaloy-4 rods. The tests also suggested that the smaller inventory of inter-granular gas in the pellets with the large grain could reduce the fission gas release during the RIA transient; and high burnup structure in pellet periphery (so-called rim structure) does not have strong effect on reduction of the failure threshold because the PCMI load is produced primarily by solid thermal expansion.     

Influence of Cladding-Peripheral Hydride on Mechanical Fuel Failure under Reactivity-Initiated Accident Conditions

In order to investigate the influence of hydrogen embrittlement on fuel failure under reactivity-initiated accident (RIA) conditions, pulse irradiation experiments were performed with unirradiated fuel rods at the Nuclear Safety Research Reactor (NSRR). Fresh cladding was pre-hydrided so that the other factors of cladding degradation, such as irradiation damage and oxidation, were excluded. Hydride clusters are circumferentially oriented and localized in the cladding periphery in order to simulate ‘hydride rim’ which is formed in high burnup PWR cladding. The present study demonstrated hydride-assisted pellet-cladding mechanical interaction (PCMI) failure which has been observed in high burnup fuel experiments. The fuel enthalpy at failure was lower when the cladding had a thicker hydride rim where surface cracks were easily generated. It indicates that the failure limit is highly correlated with the stress intensity factor assuming that the crack depth is equivalent to the hydride rim thickness. Hence, we conclude that hydride rim formation is the primary factor of decreasing the failure limit for high burnup fuels. Based on the experimental results together with an analysis on cladding mechanical state during PCMI, the present study suggests a process of through-wall crack generation which is originated with brittle cracking within the hydride rim.     

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.     

RIA Failure Threshold and LOCA Limit at High Burn-up

A correlation for RIA fuel failure threshold has been derived and compared with recent experimental data. The correlation can be used for UO₂ and MOX fuel and at hot zero power and cold zero power transients, account taken for the different initial enthalpy and for the lower ductility at cold conditions. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. A probabilistic approach has been attempted for the cold zero power case. For LOCA, a correlation has been derived expressing the LOCA limit as a function of the hydrogen accumulated in the cladding during base irradiation. However, LOCA experimental data exhibit a significant scatter, possibly because results depend on details in the conduct of the tests. Consequently, predictions of LOCA limits vs. burn-up are affected by appreciable uncertainty, and more data derived from testing of high burn-up cladding specimens are needed in order to reach firm conclusions.     

Behavior of Coated Fuel Particle of High-Temperature Gas-Cooled Reactor under Reactivity-Initiated Accident Conditions

In order to clarify the failure mechanism and determine the failure limit of the High-Temperature Gascooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 0.578 to 1.869 kJ/gUO₂, in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above 1.40 kJ/gUO₂, where the peak fuel temperature reached over the melting point of UO₂ fuel. It was concluded that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions.     

压水堆燃料棒锆-4包壳在大破口LOCA条件下的鼓胀爆破实验

研制成ＦＲＳ－２型压水堆锆－４包壳电加热模拟燃料棒，提供一种先进的实验方法和瞬态测量技术，目的在于研究锆－４包壳在大破口ＬＯＣＡ条件下的鼓胀爆破行为，给出秦山核电厂安全分析所需的爆破数据。报道了模拟燃料棒的结构、性能、包壳鼓胀爆破实验方法和破口检验内容。     

RANNS Code Analysis on the Local Mechanical Conditions of Cladding of High Burnup Fuel Rods under PCMI in RIA-Simulated Experiments in NSRR

The RANNS code analyzes behavior of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the RIA-simulated experiments in the Nuclear Safety Research Reactor (NSRR), OI-10 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. The pre-accident, or End-of-Life conditions of the rod were predicted by the fuel performance code FEAMXI-6. In the calculations by the two-dimensional model of RANNS, the plastic strain increases at the cladding ridges during PCMI were compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.     

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。     

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.     

Assessment of SAFER03 Code Using ROSA-III Break Area Spectrum Tests on Boiling Water Reactor Loss-of-Coolant Accidents

The SAFER03 computer code has a newly developed evaluation model for the analysis of various boiling water reactor (BWR) loss-of-coolant accidents (LOCAs). Analyses of the ROSA-III break area spectrum tests in a recirculation line were performed using the SAFER03 to assess the predictive capability of the code for a BWR LOCA. The ROSA-III test facility at the Japan Atomic Energy Research Institute (JAERI) was constructed to simulate a LOCA in a BWR/6-251 plant with 848 fuel bundles and 24 jet pumps. This paper summarizes the assessment results of SAFER03 which predicted the system responses and key phenomena well and the conservative peak cladding temperature (PCT) for recirculation line break tests with different break areas.     

Overview of SAFER03 Assessment Studies

The safety research for BWRs has been positively done by the JAERI, Japanese BWR utilities and BWR vendors in this decade and has shown the important phenomena under BWR LOCA conditions. Based on these significant results, the SAFER03 computer code was jointly developed by Toshiba, Hitachi and General Electric. SAFER03 has been qualified against the BWR simulation test data obtained from TBL, ROSA-III and FIST-ABWR test facilities. The objectives of this study are to assess the predictive capability of SAFER03 code to simulate the significant LOCA phenomena and to catch key parameters during BWR LOCA. This paper summarized the results of these SAFER03 assessments and showed that SAFER03 could predict the realistic behavior of BWR LOCA with slight conservative peak cladding temperatures.     

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

ABSTRACT

To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73–85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these fuel cladding tube specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10%–30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520–530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.     

LOCA Analysis of Super Fast Reactor

This paper describes loss of coolant accident (LOCA) analyses of the Supercritical-pressure Water-Cooled Fast Reactor (Super Fast Reactor). The features of the Super Fast Reactor are high power density and downward flow cooled fuel channels for the improvement of the economic potential of the Super Fast Reactor with high outlet steam temperature. The LOCA induces large pressure and coolant density change in the core. This change influences the flow distribution among the downward flow parallel channels. It will affect the safety of the Super Fast Reactor. LOCA analysis of Super Fast Reactor is important to understand the safety features of the Super Fast Reactor. Keeping the flow rate in the core is important for the safety of the Super Fast Reactor. In LOCA, it is difficult to maintain an adequate flow rate due to the once-through coolant cycle and the downward flow cooled fuel assemblies. Therefore, the early actuation of the Automatic Depressurization System (ADS) and reduction of the maximum linear heat generation rates of the downward flow seed fuel assemblies and Low-Pressure Core Spray (LPCS) system are necessary for the Super Fast Reactor to cool the core under LOCA. Analysis results show that the Super Fast Reactor can satisfy the safety criteria with these systems.