Failure of high burnup fuels under reactivity-initiated accident conditions

Pulse irradiation experiments of high burnup light-water-reactor fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO₂ rod at a burnup of 69 GW d/t failed due to pellet-cladding mechanical interaction (PCMI) in the test LS-1. The fuel enthalpy at which fuel failure occurred was comparable to those for PWR-UO₂ rods of 71 to 77 GW d/t with more corroded cladding. Comparison of cladding metallographs between the BWR and PWR fuel rods suggested that the morphology of hydride precipitation, which depends on the cladding texture, affects the fuel failure limit. The tests BZ-1 and BZ-2 with PWR-MOX rods of 48 and 59 GW d/t, respectively, also resulted in PCMI failure. The fuel enthalpies at failure were consistent with a tendency formed by the previous test results with UO₂ fuel rods, if the failure enthalpy is plotted as a function of the cladding outer oxide thickness. Therefore, the PCMI failure limit under RIA conditions depends on the cladding corrosion states including oxidation and hydride precipitation, and the same failure limit is applicable to UO₂ and MOX fuels below 59 GW d/t.     

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.     

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.     

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.     

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of 25–93 MWd kg⁻¹ UO₂, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.     

Analyses of used BWR fuel assay data with the integrated burnup code system SWAT4.0

Criticality safety of the fuel debris from the Fukushima Daiichi Nuclear Power Plant is one of the most important issues, and the adoption of burnup credit is desired for criticality safety evaluation. To adopt the burnup credit, validation of the burnup calculation codes is required. Assay data of the used nuclear fuel irradiated by the Fukushima Daini Nuclear Power Plant Unit 2 are evaluated to validate the SWAT4.0 code for solving the BWR fuel burnup problem. The calculation results revealed that the number densities of many heavy nuclides and fission products show good agreement with the experimental data, except for those of ²³⁷Np, ²³⁸Pu, and samarium isotopes. These differences were considered to originate from inappropriate assumption of void fraction. Our results implied overestimation of the (n, γ) cross-section of ²³⁷Np in JENDL-4.0. The Calculation/Experiment – 1 (C/E–1) value did not depend on the type of fuel rod (UO₂ or UO₂–Gd₂O₃), which was similar to the case of PWR fuel. The differences in the number densities of ²³⁵U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu, ¹⁴⁹Sm, and ¹⁵¹Sm have a large impact on k_eff. However, the reactivity uncertainty related to the burnup analysis was less than 3%. These results indicate that SWAT4.0 appropriately analyzes the isotopic composition of BWR fuel, and it has sufficient accuracy to be adopted in the burnup credit evaluation of fuel debris.     

Modelling of Zry-4 cladding oxidation by air, under severe accident conditions using the MAAP4 code

In a nuclear power plant, a potential risk in some low probability situations in severe accidents is air ingress into the vessel. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of Fission Products (FP), especially increasing that of ruthenium. This FP is of particular importance because of its high radio-toxicity and its ability to form highly volatile oxides. Oxygen affinity is decreasing between Zircaloy cladding, fuel and ruthenium inclusions in the fuel. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues.A review of existing data in the field of Zircaloy-4 oxidation in air-containing atmosphere shows that this phenomenon is quantitatively well understood. The cladding oxidation process can be divided into two kinetic regimes separated by a breakaway transition. Before transition, a protective dense zirconia scale grows following a solid state diffusion-limited regime for which experimental data are well fitted by a parabolic time dependence. For a given thickness, which depends mainly on temperature and the extent of pre-oxidation in steam, the dense scale can potentially breakdown. In case of breakaway combined with oxygen starvation, cladding oxidation can then be much faster because of the combined action of oxygen and nitrogen through a complex self sustaining nitriding-oxidation process.A review of the pre-existing correlations used to simulate zirconia scale growth under air atmospheres shows a high degree of variation from parabolic to accelerated time dependence. Variations also exist in the choice of the breakaway parameter based on zirconia phase change or oxide thickness. Several correlations and breakaway parameters found in the literature were implemented in the MAAP4.07 Severe Accident code. They were assessed by simulation of the QUENCH-10 test, which is a semi-integral test designed to study fuel bundle exposure to steam first and then to air. This paper deals with the main results obtained with MAAP4.07 when simulating QUENCH-10.     

The issue of stress state during mechanical tests to assess cladding performance during a reactivity-initiated accident (RIA)

The mechanical test procedures that address fuel cladding failure during a RIA are reviewed with an emphasis on the development of test procedures that determine the deformation and fracture behavior of cladding under conditions similar to those reached in a RIA. An analysis of cladding strain data from experimental research reactor test programs that have simulated the RIA is presented. These data show that the cladding undergoes deformation characterized by hoop extension subject to a range of multiaxial stress states and strain paths comprised between plane-strain (no axial extension of the cladding tube) and equal-biaxial tension (equal strain in both the hoop and the axial orientations). Current mechanical test procedures of cladding material are then reviewed with a focus on their ability to generate the appropriate deformation response and to induce the prototypical multiaxial stress states and failure modes activated during a RIA. Two main groups of tests currently exist. In the first group, the deformation behavior of the cladding is examined by several variations of hoop tensile tests in which an axial contraction of the specimen gage section occurs such that a near-uniaxial tension stress state results; finite element analyses are then usually employed to deduce the deformation response, often under conditions of an assumed coefficient of friction between the specimen and test fixtures. The second group includes test procedures which attempt to reproduce the deformation and failure conditions close to those seen during a RIA such that any stress-state corrections of the failure conditions are comparatively small. The advantages and disadvantages of all of these deformation/fracture tests are discussed with special reference to testing high burnup fuel cladding.     

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSTF), which simulated a full scale 30° sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved.     

Model of fracture for the Zry cladding of nuclear fuel rods included in the code DIONISIO 1.0

The DIONISIO code describes most of the main phenomena occurring in a fuel rod during normal operation of a nuclear power reactor. Starting from the irradiation history, the code predicts the temperature distribution, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the internal rod volume, gas mixing, pressure increase, irradiation growth of the cladding, development of an oxide layer on its surface and hydrogen uptake, restructuring and grain growth in the pellet.This work presents the model of Zircaloy fracture included in the code DIONISIO 1.0. The model of pellet-cladding mechanical interaction (PCMI) provides the forces caused by the solid-solid contact which add to the changing internal pressure and to the constant external pressure. Besides, the program evaluates the effects of a corrosive atmosphere (stress corrosion cracking, SCC) internal or external. With these data, the code calculates the J integral around the tip of an initiated crack, and proceeds to analyze, according to the quantity of corrosive substance dissolved and the cladding stress field, if the crack remains unchanged, if it grows due to the I-SCC mechanism, or if propagation is ductile, following the R curve of the material.Results corresponding to different PHWR and PWR reactors are presented and compared with code results. In particular, good agreement is obtained in the simulation of MOX experiments, where the cladding failed due to propagation of cracks originated in SCC.     

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.     

高燃耗下快堆燃料与包壳的化学相互作用模型建立及验证

快中子反应堆二氧化铀燃料元件在高燃耗、高中子注量率、高线功率和高温状况下运行,燃料与包壳材料会发生复杂的物理化学相互作用。燃料元件化学相互作用模型的建立对高燃耗快堆燃料元件的设计非常重要。针对快中子反应堆氧化物燃料元件与包壳材料发生的化学相互作用,采用动力学模型建立了二氧化铀与奥氏体不锈钢、铁素体-马氏体钢包壳材料的化学相互作用模型,并通过实验数据验证该模型。结果表明：建立的快堆二氧化铀燃料与奥氏体不锈钢的腐蚀模型可以成功预测最大燃耗10.8at%、辐照损伤87.5 dpa的包壳腐蚀;建立的快堆二氧化铀燃料与铁马钢的腐蚀模型可以成功预测最大燃耗9.3at%、辐照损伤76.6 dpa的包壳腐蚀。研究结果为高燃耗二氧化铀辐照元件及示范快堆燃料元件的设计和性能预测提供重要的参考价值。     

Influence of void fraction change on plutonium and minor actinides recycling in BWR with equilibrium burnup

A study on the influence of void fraction change on plutonium and minor actinides recycling in standard boiling water reactor (BWR) with equilibrium burnup model has been conducted. We considered the equilibrium burnup model since it is a simple time independent burnup method that can handle all possible produced nuclides in any nuclear system.

The uranium enrichment for the criticality of the reactor diminishes significantly for the plutonium and minor actinides recycling case compared to that of the once-through cycle of BWR case. This parameter decreases much lower with the increasing of the void fraction. A similar propensity was also shown in the required natural uranium per annum. The annual required natural uranium was calculated by assuming that the uranium concentration in the tail of the enrichment plant is 0.25 w%. The amount of loaded fuel reduces slightly with the increment of the void fraction for plutonium and minor recycling in BWR.     

Temperature increase on target due to jet impingement under BWR/PWR loss of coolant accident conditions

This report describes the temperature increase on the target plate after jet impingement on it from a ruptured pipe under BWR/PWR Loss of Coolant Accident Conditions. From test results it is shown that the temperature on the target can be conservatively estimated by taking it equal to the saturated temperature corresponding to the pressure on the target, which is given by the steam table. An experimental formula is presented to estimate the maximum temperature increase on the target.     

Determination of within-rod caesium and europium isotopic distributions in high burnup fuel rods through computerised gamma-ray emission tomography

Computerised gamma-ray emission tomography has been applied to single PWR UO₂ fuel rods, with pellet averaged burnups of 52, 71, 91 and 126 GWd/t respectively, for the determination of ¹³⁴Cs, ¹³⁷Cs and ¹⁵⁴Eu internal radial distributions. State-of-the-art image reconstruction techniques, analytical and iterative, have been applied, evaluated and compared using test phantoms first and, in a second step, the actual measured data. Further, linear attenuation maps, previously derived on the same samples by means of gamma-ray transmission tomography, have been used to correct for density inhomogeneities. The final results have indicated large central depressions in the caesium distributions, but of varying extent from sample to sample. Particularly interesting is the case of the 126 GWd/t sample, showing a very deep central depression (periphery-to-centre ratios of ∼2.5 for ¹³⁷Cs and ∼3 for ¹³⁴Cs). In addition, a difference in the relative activity distributions of ¹³⁷Cs and ¹³⁴Cs has been observed for all the samples. In contrast, the europium shows an almost flat distribution.     

Experimental study of jet discharging test results under BWR and PWR loss of coolant accident conditions

This report describes the results of the jet discharging experiments conducted at the Japan Atomic Energy Research Institute. The tests were done under BWR and PWR Loss of Coolant Accident conditions using 4 inch, 6 inch and 8 inch test pipes, and varying distance between the pipe exit and the target plate.Simple and practical experimental formulae to estimate the maximum pressure on the target plate and maximum pressure distribution are given. Further, relations between pipe reaction thrust forces and jet impingement forces are described.     

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.     

Calculations of reactivity-initiated transients in gas-cooled fast reactors using the code system fast

The FAST code system is a general tool for analyzing advanced reactors from the viewpoint of the static and dynamic behavior of the whole reactor system. It includes an integrated three-dimensional representation of the core neutronics, appropriate modeling of the core thermal-hydraulics and fuel pin behavior, coupled to models of the reactor primary and secondary systems. Use is made largely of well-established individual neutronic, thermal-hydraulic and fuel behavior modules. Clearly, it is important to verify the individual parts of the code, including the links between them. The paper is focused on this detailed verification procedure. Steady-state conditions, as well as the transient behavior of hypothetical reactivity-initiated accidents, are investigated for two specific gas-cooled fast reactors. While the first system, a CO₂-cooled CAPRA-CADRA core, is loaded with Superphénix-like MOX fuel, the second system being analyzed, a He-cooled Generation IV-like core, uses ceramic (U,Pu)C fuel dispersed in a silicon-carbide matrix. In the current study, the TRAC/PARCS elements of FAST are compared with the 3D-kinetics stand-alone ERANOS/KIN-3D code, which is considered state-of-the-art, using as far as possible equivalent options. A new methodology is proposed to improve a diffusion-theory, coarse-group PARCS-solution by scaling the original cross-section derivatives and input kinetic parameters.     

A parametric study of multichannel molten cladding motion under unprotected loss-of-flow accident conditions in LMFBR's

One of the key assumptions of the present multichannel clad motion model was that the total pressure drop over the voided channel could be supplied as a boundary condition. The incoherency effect on cladding motion can be significant for a full-scale subassembly, and therefore parametric studies of the total pressure drop and oscillatory pressure effect due to sodium chugging were examined using the multichannel model.There is an axial blanket region in demonstration plant or commercial-power-plant designs instead of a reflector in FFTF design above the top of fuel. It was shown that due to the difference in the thermal conductivities between the blanket material and reflector, significant changes in the timings of various events of the cladding relocation might occur. It is also noted that depending on the effect of the sodium voiding on the reactivity, the fuel may become molten when the molten cladding is still around. The possibility of the occurrence of this situation is studied by increasing the power in the present model.