Effects of Coolant Flow on Light Water Reactor Fuel Behaviors during Reactivity Initiated Accident

This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.     

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.     

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.     

Experimental Results of Some Cluster Tests in NSRR

The NSRR programme is in progress in JAERI using a pulsed reactor to evaluate the behavior of reactor fuels under reactivity accident conditions. This report describes briefly the experimental results and preliminary analysis of two cluster tests.

In the cluster configuration of five fuel rods, the power distribution in outer fuel rods are not symmetric due to neutron absorption in central fuel rod. The cladding temperature on the exterior boundaries of the cluster is higher than that in interior. Good agreement was obtained between the calculated and measured cladding temperature histories. In the 3.8$ excess reactivity test, cluster averaged energy deposition of 237 cal/g-UO₂, cladding melting and deformation were limited to the portions of the fuel rods that were on the exterior boundaries of the cluster.     

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.     

Molten fuel radial motion and cladding melting during a PCM event in LWRs

The thermal behavior of the fuel and cladding during off-normal operating conditions, generally termed power-cooling-mismatch (PCM), are of great interest to light water reactor (LWR) safety analysis. During a power-cooling-mismatch event, fuel melting may begin at the center of the rods propagating radially outward. The induced pressure at the center of the rod due to fuel melting, fission gas release, and UO₂ fuel vapor would tend to force such molten fuel to flow through radially open cracks in the outer unmelted portion of the pellet and relocate in the fuel-cladding gap. The zircaloy cladding, which is at high temperature during film boiling, may undergo melting at its inside surface upon being contacted by the extruded molten fuel, eventually resulting in a thermal failure of the cladding.Three topics of interest are analyzed in this paper. First, fuel conditions during a hypothesized PCM accident are assessed with regard to pellet cracking and central fuel melting. Secondly, the transient freezing of a superheated liquid penetrating an initially empty crack, maintained at constant subfreezing temperatures, is studied analytically. The analysis is presented in a dimensionless form, illustrating the effect of the governing parameters, namely the driving pressure, crack shape (that is, a divergent, a parallel wall, or a convergent crack), density ratio, Stefan number for freezing, and steady state crust thickness. The calculational results are used to assess the radial extrusion of molten UO₂ fuel observed in some in-pile tests, in which PCM conditions in a pressurized water reactor were simulated. Thirdly, conditions for potential melting of zircaloy cladding upon being contacted by the extruded molten fuel are investigated analytically. The analytical predictions were consistent with the experimental results from PCM in-pile tests.     

Numerical Analysis of the PHEBUS Containment Thermal Hydraulics

This paper describes numerical analysis of the PHEBUS FP containment thermal-hydraulics. PHEBUS FP is an international project undertaken with the aim of evaluating the behavior of radioactive fission products released from a LWR pressure vessel into the containment vessel during a hypothetical severe accident. Six integral in-pile tests have been planned and are being carried out at Cadarache, France. The European Union, the United States, Canada, Korea and Japan are participating in this project. From Japan, the Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute are collaborating the other parties involved in the project.

Since the behavior of fission products is strongly dependent on the surrounding environmental conditions, accurate prediction of the thermal-hydraulics in the containment vessel is essential to accurately evaluate the behavior. Characteristics of condensation heat transfer in the presence of noncondensable gases play a key role in the PHEBUS thermal-hydraulics, especially under the condition of high noncondensable gas mass fraction. Many models for condensation heat transfer in the presence of noncondensable gases have been proposed. However, these models were not found suitable for PHEBUS analysis, because they were focused on the low noncondensable gas mass fraction condition.

In this study, a single-phase multi-component code, TFLOW-FP has been newly developed to predict thermal-hydraulics in the PHEBUS FP containment. Moreover, a new degradation factor correlation for the condensation heat transfer coefficient due to the presence of noncondensable gases has also been developed and incorporated into the code. This code was applied for analysis of the thermal-hydraulic benchmark tests and the first in-pile test, FPTO. The results show that the code can predict the total pressure, gas temperature distributions, the relative humidity in the containment vessel and steam condensation rate on the surface of condenser rods very well.     

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.     

Measurement and Evaluation on Pulsing Characteristics and Experimental Capability of NSRR

The NSRR programme is in progress in JAERI using a pulsed reactor to investigate fuel behaviors under the reactivity-initiated accident conditions. Pulsing characteristics and experimental capability, especially heat deposition in test fuel rods given by a single pulse are key parameters to this purpose.

In pulsing performance tests, it has been ascertained that the maximum pulsing with 4.67$ (=3.41%δk) brings peak reactor power of 21,100 MW and core energy release of 117 MW·sec. The calculated time responses of reactor power, fuel temperature and cladding surface temperature as well as these maximum values at various pulse sizes agreed well with measured data. In addition, it has been also ascertained by measurement as well as analysis that there are no essential differences in pulsing characteristics between the pulsing from critical and that from subcritical.

The heat deposition in a test fuel rod given by a single pulse is much enough as predicted, and a 2.6% enriched BWR type fuel rod gains about 230cal/g-UO₂ in the maximum pulsing. In case of irradiation of clustered five test fuel rods by a single pulse, heat deposition reduces by about 20% for a surrounding rod and about 40% for a center rod in comparison with that in a single rod irradiation.     

Pellet compliance model based on out-of-pile simulation

The thermal and mechanical behavior of fuel rods is significantly influenced by the extent of their relocation and by compliance of the cracked pellets. Movement of the cracked pellet pieces towards the cladding results in softer pellets with crack voids which accommodate some fraction of the thermoelastic pellet deformation and make the pellet more compliant under the restraint of the cladding. It is difficult to model such a pellet compliance independently of experimental observations because the cracked pellet behavior is uncertain by nature.Electrically heated simulation of pellet-cladding mechanical interaction (PCMI) facilitates much quicker and more flexible experimentation than actual in-pile tests. Testing apparatus consists of the simulated fuel rod with hollow UO₂ pellets and a tungsten rod in the center, and a diameter measuring device including three pairs of diameter sensors. Test parameters include the pellet-cladding gap and the cladding thickness. Results show that rods with a smaller gap have a larger increasing rate of cladding diameter. This suggests that a group of cracked pellet pieces induced by thermal stress has an apparent compliance which increases with pellet-cladding gap. Results also show more sensitivity to cladding thickness than those calculated assuming pellets having intrinsic stiffness. This also suggests the compliant nature of cracked pellets.Such a compliant nature can almost be described by reducing the elasticity of the pellet. A simple pellet compliance model was obtained by fitting calculations with measurements to describe a cracked pellet as a uniform axisymmetric body with apparent elasticity.     

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.     

Clad-to-Coolant Heat Transfer in NSRR Experiments

The evolution of the clad temperature during a Reactivity Initiated Accident plays a key role in the accidental sequence because it strongly influences the rod mechanical resistance against failure. The present study aimed at quantifying the heat transfer in NSRR experiments. Transient boiling curves were determined by inverse conduction calculations of NSRR experiments in which the clad outer surface temperature had been measured by spot-welded thermocouples. Critical Heat Fluxes (CHFs) as high as 13 MW/m² have been obtained, highlighting a considerable increase compared to stationary pool boiling conditions. The elevated CHFs are due to the intense transient fluid vaporization at the surface induced by a fast clad heating rate. A transient boiling model has been implemented in the SCANAIR code on the basis of the physical interpretation of the boiling curves. A good agreement between computed and experimental clad temperatures is obtained for high burnup fuel tests as for fresh fuel tests.     

An in-pile sodium loop for investigation of local cooling disturbances in LMFBR's

In the safety analysis of Liquid Metal Fast Breeder Reactors, investigations of the fuel element behavior under local off-normal cooling conditions and the possible failure propagation are of special interest. In a common program, called “Mol 7C” the Gesellschaft für Kernforschung, Karlsruhe, and the Centre d'Etude de l'Energie Nucléaire/Studiecentrum voor Kernenergie, Mol, are performing related in-pile experiments in a sodium loop in the BR 2-reactor. The test section contains a 37-rod bundle of fresh UO₂-fuel. A local blockage within the fuel bundle will initiate a certain local damage to a few rods. The experiments are expected to obtain important informations with respect to the problems of pin to pin propagation and the long term behaviour of a fuel bundle with defect pins. The in-pile part of the loop contains the fully integrated primary sodium circuit. Total heat removal capacity is about 700 kW. The equipment for the first experiment is nearly manufactured. The first experiment will start in the beginning of 1977. At first three experiments are planned.     

Behavior of Irradiated ATR/MOX Fuel under Reactivity Initiated Accident Conditions

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20MWd/kgHM were conducted at the NSRR in Japan Atomic Energy Research Institute to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Relatively large radial deformation of the fuel rods due to pellet-cladding mechanical interaction occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet.     

CFD analyses in tight-lattice subchannels and seven-rod bundle geometries of a Super Fast Reactor

This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.     

Experimental studies on the thermal and hydraulic performance of the fuel stack of the VHTR: Part II. HENDEL multi-channel test rig with twelve fuel rods

Tests on heat transfer and fluid-dynamics of the mock-up fuel stack of the Very High Temperature Gas Cooled Reactor (VHTR) were performed by the multi-channel test rig (T_{1 - M}) of the Helium Engineering Demonstration Loop (HENDEL). The T_{1 - M} simulated one column of the fuel stack in the VHTR core and contained twelve simulated fuel rods. The heat generation rate of each fuel rod was varied to simulate the power distribution of the VHTR core in the horizontal plane. In parallel with this experiment, a three-dimensional thermal analysis code was developed in order to check the experimental temperature distribution of the fuel stack.Experimental results showed that the distribution of the helium gas flow rate was influenced by the temperature distortion in the mock-up fuel stack. The maximum deviation of the helium gas flow rate from the mean value was 10% in the case of an asymmetric power distribution test at a low Reynolds number. The variation of the calculated temperature distribution in the fuel stack was about 17-35°C in each case, indicating that the temperature distortion in the fuel stack was flattened by thermal conduction in the graphite block.     

Verification of FEMAXI-7 code by using irradiation test in Halden reactor for He-pressurization effect on FGR of BWR fuels under power transient

Power ramp test for He-pressurization effect on fission gas release (FGR) of about 42GWd/tUO₂ boiling water reactor (BWR) fuel rods was analyzed by the fuel performance code FEMAXI-7. The experimental data were obtained with the two rods, which were base irradiated in the Halden reactor for 12 years (IFA-409), then subjected to the power ramp tests (IFA-535) to investigate the He-pressurization effect. The FEMAXI-7 calculations were performed by inputting rod specifications and experimental conditions in both the baseand test irradiations. The results showed that the calculations reasonably followed the trends of measured cladding elongation and FGR during the power ramp test, depending on the pellet temperature and fission gas atoms diffusion rate. Based on the calculated results, the reason that no apparent He-pressurization effect was observed in the experiment was considered to be caused by insufficient gas communication during strong pellet–clad mechanical interaction (PCMI) and enhanced gap thermal conductance by the solid–solid contact due to gap closure.     

Development of Hydride Absorber for Fast Reactor—Evaluation of Design Method for Hydride Absorber Rod—

A hydride control rod is being developed to improve the economy of fast reactor plants because it has a longer lifetime than the currently used B₄C control rod. A hydride burnable poison rod is also under development to reduce the number of control rods by decreasing core excess reactivity. Hydrogen in the hydride control rod causes neutron spectrum interference between the fuel and control rod regions. Thus, the study on core design was performed with the continuous-energy Monte Carlo code MVP using the nuclear data library JENDL-3.3 to deal with this phenomenon precisely. To evaluate the applicability of MVP to hydride absorber rod design, two benchmark calculations were carried out. One of them is a hydrogen-contained metal fuel fast core constructed in Fast Critical Assembly (FCA) and the other is the Nuclear Safety Research Reactor (NSRR) core where zirconium-hydride fuel (U-ZrH_1.6) rods are loaded. These benchmark calculations and the design study on a fast reactor core with hafnium-hydride control rods have revealed that MVP is a reliable tool for hydride absorber rod design.     

Fuel Assembly Hydraulic Test of Research Program on In-Pile Irradiation Test for Advanced UO2 Pellets

In order to ensure that the irradiation test for fuel assembly is safe, it is necessary to determine that the coolant velocity on the surface of fuel rods and hydraulic conditions. There is no flow meter on the fuel assembly, both out-of-pile and in-pile hydraulic test have been completed and the flow rate of coolant oassed through the fuel assembly is determined in terms of test results.