Thermal Stress Analysis of High-Burnup LWR Fuel Pellet Pulse-Irradiated in Reactivity-Initiated Accident Conditions

For RIA-simulated experiments in the NSRR with high-burnup PWR fuel and BWR fuel, numerical analyses were performed to evaluate the temporal changes of profiles of temperature and thermal stress in pellet induced by pulse power, using the RANNS code. The pre-pulse states of rods were calculated using the fuel performance code FEMAXI-6 along the irradiation histories in commercial reactors and the results were fed to the RANNS analysis as initial conditions of the rod. One-dimensional FEM was applied to the mechanical analysis of the fuel rod, and the calculated cladding permanent strain was compared with the measured value to confirm the validity of the PCMI calculation. The calculated changes in the profiles of temperature and stress in the pellet during an early transient phase were compared with the measured data such as the internal gas pressure rise, cracks and grain structure in the post-test pellet, anddiscussed in terms of PCMI and grain separation. The analyses indicate that the pellet cracking appearances coincided with the calculated tensile stress state and that the compressive thermal stress suppresses the fission gas bubble expansion leading to grain separation.     

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.     

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.     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

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.     

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.     

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 on split failure of cladding of high burnup BWR rods in reactivity-initiated accident conditions by RANNS code

A computer code RANNS was developed to analyze fuel rod behaviors in the reactivity-initiated accident (RIA) conditions. RANNS performs thermal and finite-element mechanical calculation for a single rod in axis-symmetric geometry, where fuel pellet consists of 36 equal-volume ring elements and cladding metallic wall consists of eight equal-thickness ring elements and one outer oxide element. The code can calculate temperature profile inside the rod, contact pressure generated by pellet–clad mechanical interaction (PCMI), stress–strain distribution and their interactions elaborately. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by the fuel performance code FEMAXI-6.In the present study, analysis was performed on the simulated RIA experiments in the “nuclear safety research reactor” (NSRR), FK-10 and FK-12, with high burnup BWR rods in a cold-start up condition, and stress–strain evolution in the PCMI process was calculated extensively. In the analysis, the pellet–clad bonding was assumed both in the heat conduction and in mechanical restraint. The calculated hoop strain increase was compared with the measured strain gauge data, and satisfactory agreement was obtained. Simulation calculations with broader power pulses anticipated in RIA of commercial BWR were carried out and the resulted cladding hoop stress was compared with the failure stress estimated by comparison of analysis with experimental data.     

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.     

Evaluation of neutron economical effect of new cladding materials in light water reactors

The present paper aims to contribute from a neutronic aspect to activities for new cladding material development for light water reactors (LWRs) that can reduce the risk of hydrogen gas explosion. Iron (Fe), nickel (Ni), titanium (Ti), niobium (Nb) and vanadium (V) are selected as possible component elements to cover a variety of new cladding materials for LWRs. The effect of larger thermal absorption cross sections of these elements than those of zirconium (Zr), together with those of silicon carbide (SiC), on the neutron economy in LWRs is evaluated by performing pin cell burnup calculations for a conventional pressurized water reactor (PWR), a low-moderation high-burnup LWR (LM-LWR) and a high-moderation high-burnup LWR (HM-LWR). As can be anticipated from the thermal cross sections, SiC has excellent neutron economy. The materials other than SiC largely decrease discharge burnup for all three types of LWRs in comparison with Zircaloy-4. Among such elements of larger thermal absorption cross section, Nb has neutron economical advantage over the other materials except SiC in softer neutron spectrum reactors such as HM-LWR in which the atomic number ratio of hydrogen to heavy metal is 6. In conventional LWRs, stainless steel of low Ni contents has the neutron economic advantage as well as Nb for cladding material. The results of the calculations are summarized for the purpose to provide reference data for new cladding material development studies, in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding.     

Deformation behavior of recrystallized and stress-relieved Zircaloy-4 fuel cladding under biaxial stress conditions

Pellet–cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to the failure of high-burnup fuel rods. Biaxial stress states generated by PCMI in Zircaloy cladding may make the cladding more susceptible to failure. In this study, we investigated the deformation behavior of Zircaloy cladding under biaxial stress conditions based on the concept of contours of equal plastic work. The major axis angles of the initial work contours of recrystallized (RX) and stress-relieved (SR) specimens were investigated and it was found that the shapes of the initial work contours of these kinds of specimens were almost symmetric across the direction where the ratio of axial stress to circumferential stress is 1. The shapes of subsequent work contours tended to change for the RX specimen while being the same as the initial for the SR specimen, as deformation proceeded. It was suggested that the textures and slip systems in the RX and SR specimens affect their initial work contours while the slip system in the RX specimens and the residual strain in the SR specimens influence the subsequent work contours.     

Improvements to COMETHE III-J fuel rod modeling code

Current improvements to the COMETHE fuel performance code focus on pellet-clad axial interaction and Zircaloy cladding failure predictions. Slipping and sticking between pellets and clad as well as trapped stack are evaluated. The main conclusions are that slipping with friction concerns only local effects while axial PCMI is primarily dependent on pellet expansion with a strong ‘strain biaxiality’ effect dictated by the dishing. The notion of locking prior to radial PCMI is also introduced, which explains experimental features not previously understood. Benckmarking of the version of COMETHE against ramp tests has been initiated and will enable assessment of the code capability in Zircaloy clad failure predictions.     

Stress Intensity Factor at the Tip of Cladding Incipient Crack in RIA-Simulating Experiments for High-Burnup PWR Fuels

RIA-simulating experiments for high-burnup PWR fuels have been performed in the NSRR, and the stress intensity factor K _I at the tip of cladding incipient crack has been evaluated in order to investigate its validity as a PCMI failure threshold under RIA conditions. An incipient crack depth was determined by observation of metallographs. The maximum hydride-rim thickness in the cladding of the test fuel rod was regarded as the incipient crack depth in each test case. Hoop stress in the cladding periphery during the pulse power transient was calculated by the RANNS code. K _I was calculated based on crack depth and hoop stress. According to the RANNS calculation, PCMI failure cases can be divided into two groups: failure in the elastic phase and failure in the plastic phase. In the former case, elastic deformation was predominant around the incipient crack at failure time. K _I is available onlyin this case. In the latter, plastic deformation was predominant around the incipient crack at failure time. Failure in the elastic phase never occurred when K _I was less than 17 MPam1^/2. For failure in the plastic phase, the plastic hoop strain of the cladding periphery at failure time clearly showed a tendency to decrease with incipient crack depth. The combination of K _I, for failure in theelastic phase, and plastic hoop strain at failure, for failure in the plastic phase, can be an effective index of PCMI failure under RIA conditions.     

Fracture behavior of recrystallized and stress-relieved Zircaloy-4 cladding under biaxial stress conditions

Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to failure of high-burnup fuel rods. The biaxial stress states of Zircaloy cladding tubes induced by PCMI loading presumably makes the tubes more susceptible to failure. To clarify the influence of the anisotropic mechanical properties of Zircaloy cladding tubes on their fracture behavior under biaxial stress conditions, biaxial tensile tests were performed, and the measured stresses and strains were converted to reduced parameters such as equivalent strain, equivalent stress, and stress triaxiality by using the anisotropic constants of the Hill yield function derived in our previous study. The minimum fracture strains of recrystallized (RX) and stress-relieved (SR) specimens were located where the stress ratio of axial to circumferential is 0.75. The equivalent plastic fracture strains tended to decrease monotonously with increasing stress triaxiality, which is a typical trend observed in ductile fracture, in the range of 0.65–0.78 for both specimens. In the case of SR specimens, however, the analysis with stress triaxiality did not reduce the fracture strains well to a single trend curve, showing that another influential factor is required such as an anisotropic fracture behavior or a plastic behavior deviated from that predicted by the yield function.     

Cladding Deformation due to Gas Bubble Swelling of UO2 Pellet in Bump Test

Fission gas release in the bump tests was correlated to the deformations of claddings via mathematical product of the number of gas atoms and their residing time on grain boundaries. A positive correlation of the deformations with the product indicated that gas bubble swelling of pellets contributed to the pellet-cladding mechanical interaction (PCMI).

Residual gaps prior to the bump tests turned out to be filled in at the bump terminal level by differential thermal expansion of pellet and cladding. Visible macro cracks existing in the central part of the pellet virtually vanished during the tests due to bubble swelling of the pellet. Together with these observations, quantitative image analysis of pellet porosity showed that the aforementioned PCMI was brought about by combination of differential thermal expansion and bubble swelling.

A model to unify gas release, bubble growth and PCMI simulated well the observed behaviors of fission gas release from bump-tested rods. It was deduced by the model that higher burnup retains a higher potential for PCMI, while power reductions and associated gas releases reduce PCMI.

In the analysis in this paper were used the data of the Rise Transient Fission Gas Release Project.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

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.     

Determination of stress within ZIRCALOY cladding due to pellet-cladding interaction

A precise calculation of the stress distribution within the Zircaloy cladding of a water-cooled reactor fuel rod subjected to a power increase is a complex problem which, in general, requires a computer code to integrate the behaviour of both the fuel and cladding. This paper develops a simplified model which decouples the clad and fuel pellet analyses, by considering two extremes of fuel pellet mechanical behaviour, which lead to two widely different boundary conditions at the pellet-clad interface. An axisymmetric fuel rod code can be used to give the mean cladding hoop strain imposed by the thermal expansion of the pellet, and when the interfacial friction coefficient is 0.5, this information along with the frictional boundary condition can be used to determine the stress distribution within the cladding near a fuel pellet crack. Results from this simplified approach, which does not involve an integrated code, are used to study the growth of stress corrosion cracks within the cladding.     

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.     

高燃耗下UN-FeCrAl燃料元件稳态及瞬态热力学性能分析

UN-FeCrAl燃料元件作为耐事故燃料高燃耗应用的主要方案之一,需要评价其在高燃耗下的热力学性能。本研究基于FUPAC软件对UN-FeCrAl燃料元件在燃耗68000 MW·d·t^-1(U)下的稳态和瞬态热力学性能进行了预测。分析结果表明,稳态工况下UN-FeCrAl燃料元件热力学性能表现良好;瞬态下UN燃料的芯块中心温度最高仅为862℃,可满足芯块温度设计要求,但FeCrAl包壳的瞬态应力最大将达到459 MPa,且瞬态应变量相比于稳态应变量最大增加了0.23%,这可能会使FeCrAl包壳面临瞬态应力和瞬态应变准则超限的风险。因此后续研究应重点关注FeCrAl包壳的瞬态应力和瞬态应变性能。