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.     

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.     

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.     

Fuel swelling due to retained fission gas and thermal effects during high temperature transients

Fuel swelling of previously irradiated pressurized-water-reactor-type fuel rods tested under power-cooling-mismatch conditions is due to retained fission gas and thermal effects within the film boiling region. In this paper empirical correlations for fuel swelling are presented, and mechanisms contributing to the observed swelling and the applicability of an analytical fission gas behavior computer code (GRASS-SST) to fuel swelling are evaluated. Major contributors to fuel swelling are fuel melting and expansion, expansion of solid fuel, fission gas bubble coalescence, fission gas diffusion to grain boundaries, and change in surface tension of fuel upon melting. The contributions to fuel swelling of solid fission products and the effects of cladding contraction and wall thinning on rod swelling are also included. The overall empirically-calculated fuel swelling values and the GRASS-SST code calculated values are compared with measured values. The agreement between measured and empirically calculated fuel swelling is generally close. Fuel swelling due to retained fission gas during the film boiling transient, as calculated by the GRASS-SST code, was found to be in good agreement with experimental results.     

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.     

Oxide thickness-dependent transient cladding hoop stress

A few thrice-burned Zry-4 fuel assemblies which were loaded in one of the PWRs operating in Korea were found to be failed due to PCI during a power ramp following a rector trip, while thrice-burned Zr–Nb fuel assemblies and twice-burned Zry-4 ones were intact. To investigate the effect of fuel rod oxide thickness on power ramp-induced cladding hoop stress, three intact fuel rods were selected, which include an intact twice-burned Zry-4 fuel rod, an intact thrice-burned Zr-4 fuel rod and an intact thrice-burned Zr–Nb fuel rod. With the use of a fuel performance analysis code, burnup-dependent steady-state cladding stress and ramp power-dependent cladding stresses at the power-ramped burnup were predicted for the three intact fuel rods. It was found that the cladding oxide thickness has a considerable effect on the ramp power-dependent cladding hoop stresses. In addition, the cladding maximum stress of the thrice-burned Zry-4 fuel rod with 125 μm oxide thickness exceeded an ultimate cladding tensile strength of the Zry-4 cladding when the pellet–clad friction coefficient-dependent cladding stress concentration ratio was considered. However, the thrice-burned Zr–Nb fuel rod with 50 μm oxide thickness was evaluated to have a considerable margin against the power ramp-induced PCI failure.     

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.     

Towards understanding beneficial effects of slow power ramps

The topic of pellet cladding interaction (PCI) in light water reactor (LWR) fuel rods is reviewed and further explored in this paper by considering the mitigating effect of slow ramps. To that end, a number of PCI mitigating mechanisms were considered in relation to several failure criteria, including peak stress, strain energy density and other cumulative damage formulations. Towards a tentative working hypothesis for PCI mitigation by low ramp rate, the results fuel codes simulations of some slow ramps that have been part of the 3rd SCIP MWS (Studsvik Cladding Integrity Program Modeling WorkShop) are used in the paper in order to support or question the applicability of different possible mechanisms, considered as potential mitigating factors in slow ramps. In addition, the results of out-of-pile biaxial tests on irradiated cladding are presented for a loading scheme that aimed at reproducing the stress rate conditions of a slow ramp; the relatively small contribution of stress relaxation during the power increase stage of a slow ramp that was noticed in code calculations was confirmed by the mechanical biaxial tests. Thus, the main outcome of the code calculations in the 3rd SCIP MWS and related SCIP studies coupled with the literature review was that slow ramps’ benefits cannot be explained by mechanical effects only and a hypothesis is proposed that is based on the protective role of the cladding inner oxide and the healing of its cracks during a slow ramp provided the corresponding fuel rod gap chemistry exists.     

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。     

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code.The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from the plenum - resulting in a considerable delay of the predicted moment of cladding rupture.     

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.     

Code CRATERE — calcul du comportement d'un crayon combustible en regime accidentel ou en transitoire d'exploitation dans un reacteur a eau

The CRATER code has been developed by Electricité de France, Service d'Etudes et Projets Thermiques et Nucléaires, to predict the thermal and mechanical behaviour of LWR fuel rod in transient conditions. From the thermal standpoint it is a 2-D code (r, z) using as boundary conditions the primary coolant conditions. From the mechanical standpoint it is a 1-D code (r) since the fuel length over all is much greater as compared to the clad radius. The clad ballooning and its oxydation in the high temperature range are taken into account. Nevertheless there is an axial coupling through the inner gas of the rod. The Mechanical and thermal computations are coupled particularly because of the heat transfer evaluation through the fuel-clad gap. The initial conditions of the transient are in the case of a fresh rod generated by the Crater code. However if irradiated rods are considered the input data at the beginning of the transient are provided from other codes like the COMETHE code. As an example the mechanical and thermal behaviour of the 17 × 17 fuel rod during a power transient following a control rod ejection has been evaluated.     

CUPIDON-code decrivant le comportement thermique et mecanique d'un crayon combustible de reacteur a eau sous pression au cours d'un accident de depressurisation

In the scope of the PHEBUS experimental program to be performed in Cadarache on the behaviour of PWR's fuel assemblies under loss of coolant accidental conditions, a computer code has been developed to help designing the experimental rods and to contribute to the definition of the test runs.This code, dubbed CUPIDON, deals only with the thermal and mechanical behaviour of the rods as well as the oxidation of the cladding outside surface; it does not include any thermohydraulic subroutine. Rather, it is coupled with the RELAP code for providing necessary input data such as coolant temperatures and pressures and cladding-to-coolant heat transfer coefficients. It is restricted to a single, non irradiated, rod of short length as representing the PHEBUS experimental conditions.It is a two dimensional code, using a finite difference resolving technique. It calculates the radial thermal profile across each section of the rod, the stress and creep rate to which the cladding is submitted and the rate of formation of the oxide layer on the surface of the cladding under steady state and transient conditions. As cladding plastic strain input data, it is using the EDGAR-ZY experimental results.     

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.     

Validation of frey for the safety analysis of LWR fuel using transient fuel rod experiments

The analysis and comparison of severe light water reactor transient experiments are presented from the FREY verification and validation effort. The purpose of this study was to validate the predictive capabilities of the code for severe transient analysis. The FREY code, developed under the sponsorship of the Electric Power Research Institute, uses a two-dimensional finite-element computational method for the thermomechanical analysis of LWR fuel rods under steady state and transient conditions. A total of 10 test fuel rods from experimental programs conducted in both the Power Burst Facility and the Transient Reactor Test Facility have been used in this study. The fuel rods were selected from the following test programs: Power Coolant Mismatch Tests, PCM-2 and PCM-4: Reactivity Initiated Accident Test, RIA 1–2; Loss-of-Coolant Accident Test, LOC-3; First Fuel Rod Failure Test, FRF-1; and Irradiation Effects Test, IE-3. The test programs used in this study cover a large range of code applications for severe transient analysis. The methods used to model the fuel, cladding, and coolant geometry are discussed in addition to experimental data comparisons. The results of the PCM-2, RIA 1–2, and FRF-1 analyses are presented to highlight the full two-dimensional modeling capabilities of FREY and to compare the thermal and mechanical measurements with FREY's prediction. The comparisons show good general agreement, with a tendency for FREY to overpredict the peak cladding surface temperature for a few cases where strong three-dimensional effects have been identified.     

束流瞬变下CiADS次临界反应堆燃料包壳响应特性研究

通过修改系统分析程序RELAP5 MOD4.0的点堆动力学模型与流动传热模型,使其具备了模拟液态铅铋冷却次临界反应堆动力学特性的能力;利用改进的程序模拟了加速器驱动嬗变研究装置（CiADS）的次临界反应堆燃料包壳在发生束流瞬变时的响应特性;利用ANSYS17.0程序分析了CiADS次临界反应堆燃料包壳束流瞬变下的应力变化。研究表明：失束时间越短,燃料包壳的温度回升越慢;燃料包壳不会因可能发生的束流超功率事件而发生熔毁;燃料包壳内外壁面的温差变化是影响应力变化的主要因素;CiADS次临界反应堆的燃料包壳不会因束流瞬变而发生应力破坏。     

FRAPCON-2动力堆燃料棒稳态性能程序移植

一、程序移植目的1.模拟动力堆燃料元件单棒稳态运行性能供燃料元件单棒的设计和运行参考我国反应堆燃料性能程序研究起步较晚,移植美国NRC经过多年发展和校验的标准程序有利于我国动力堆燃料元件的设计、制造和运行方面的研究,对发展我国轻水堆燃料元件稳态性能程序有一定参考价值。