Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.     

Analysis on Lift-Off experiment in Halden reactor by FEMAXI-6 code

Analysis was conducted on the Lift-Off experiment IFA-610.1 in Halden reactor by the FEMAXI-6 code using detailed measured data in the test-irradiation. Fuel center temperature was calculated on the two assumptions, i.e. (1) an enhanced thermal conductance across the pellet-clad bonding layer is maintained during the cladding creep-out by over-pressurization, and (2) the bonding layer is broken by the cladding creep-out, and these results were compared with the measured data to analyze the effect of the creep-out by over-pressure inside the test pin. The measured center temperature rise was higher by a few tens of K than the prediction performed on the assumption (1), though this difference was much smaller than the predicted rise on the assumption (2). Therefore, it is appropriate to attribute the measured center temperature rise to the decrease of effective thermal conductance by irregular re-location of pellet fragments, etc. which was caused by cladding creep-out.     

Numerical treatment of the cladding mechanics in the fuel rod computer code IAMBUS

IAMBUS (INTERATOM Model for Burn-Up Studies on Fuel Rods) is a digital computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fuel rods. The mechanical analysis of the cladding is approximated by a state of generalized plane strain. The analysis includes routines for plasticity, creep and swelling due to void nucleation and growth. The numerical integration of these equations is described in detail and the accuracy of the results is discussed.     

On the rate determining step in fission gas release from high burn-up water reactor fuel during power transients

The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved.     

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO₂ fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail.     

Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code

Capabilities of the FEMAXI-6 code to analyze the behavior of high burnup MOX fuels in LWRs have been evaluated. Coolant conditions, detailed power histories and specifications of the MIMAS-MOX fuel rods, rod 10 and rod 11, of IFA-597.4–7 irradiated in the Halden reactor were input, and calculated rod internal pressures and pellet center temperatures were compared with the measured data for the range of 0-31 MWd/kgUO₂. Some sensitivity studies were conducted mainly with respect to pellet thermal conductivity and swelling rate to investigate the changes in thermal behavior and their effects on fission gas release.

In the irradiation period up to about 23 MWd/kgUO₂, the calculated pellet center temperatures sufficiently agreed with the measured data and also the calculated rod internal pressures reproduced the tendency of an increase in the measured rod internal pressures. These results suggest that fission gas release from MOX fuels can be reasonably predicted by a diffusion process that is modeled in UO₂ pellet grains. On the other hand, the steep increase in the measured rod internal pressures observed at the power ramp around 23 MWd/kgUO₂ cannot be reproduced by FEMAXI-6 and can be regarded as the result of a relatively large amount of gas release, which possibly caused a pellet-cladding-gap closure through pellet gas-bubble swelling.     

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.     

Consequences of a step variation in coolant temperature on a fuel rod with cladding

Time dependent temperature distributions in cylindrical fuel rods with cladding are evaluated analytically. The transient condition is due to a step variation in coolant temperature. Some numerical results are related and a short discussion is introduced.     

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.     

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.     

Experimental and analytical studies on turbulent heat transfer performance of a fuel rod with spacer ribs for high temperature gas-cooled reactors

Turbulent heat transfer performance of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was studied for various Reynolds numbers using an annular channel at the same coolant condition as the reactor operation, maximum outlet temperature of 1000 °C and pressure of 4 MPa, and analytically by a numerical simulation using the k- turbulence model. The turbulent heat transfer coefficients of the fuel rod were 18–80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the empirical correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer ribs and the axial velocity increase due to a reduction in the annular channel cross-section.     

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.     

Simulation of the thermomechanical effects originated on a fuel pin with a cracked pellet by different power ramp velocities, using a two dimensional finite element model

Thermomechanical effects originated on a fuel pin with a cracked pellet by different power ramp velocities has been studied. To such purpose, the response of a simplified two dimensional finite element model including viscoelastic properties on the pellet and elastoplastic properties on the clad was investigated, under the influence of various ramps with durations ranging from 0 s (instantaneous ramp) to 200 s (slow ramp).The importance of both, the viscoelastic relaxation effects and their interaction with the plastic behaviour of the clad was put in evidence. It has been observed, that the mechanical response of this fuel pin model strongly depends on the ramp's duration.     

Application of chemical interaction between (Fe,Cr) oxides and Mo oxide at high temperature for self-healing intelligence on nuclear fuel cladding in LWRs

A novel scheme for a bilayer coating with self-healing ability is proposed in this study. The candidate materials for the coatings and the potential self-healing reaction are assessed in high-temperature aqueous environments and high-temperature air. The pure Cr₂O₃ layer and the composite of Cr₂O₃ and MoO₃ are the candidate materials for the outer layer and inner layer, respectively, due to their compatibility under normal condition and fabricability. Fe₂O₃–MoO₃ reactions exhibit a potential ability to heal the cracks because of a high reaction rate under normal condition. The self-healing process proceeds via the following mechanism under normal condition: Fe₂O₃ (a corrosion product in the coolant) diffuses into the cracks on the coating and reacts with MoO₃ (inner layer) to produce the insoluble Fe₂(MoO₄)₃, which deposits and repairs the cracks. In the loss-of-coolant accident (LOCA) situation, Cr₂O₃–MoO₃ reaction is expected to strengthen the adhesion of the coating.     

Evaluation of a pellet-clad mechanical interaction in mixed oxide fuels during power transients by using axisymmetric finite element modeling

MOX fuel rod behavior due to PCMI during power transients was evaluated using a finite element code, ABAQUS. Clad elongation is calculated through a coupled temperature–displacement analysis where a half-pellet is axisymmetrically modeled. Parametric study for the PCMI model is preliminary performed to identify the dominant factors and examine the applicable range of the model. The comparison of the predicted results with recent MOX in-pile data shows that the centerline temperature and clad elongation are evaluated within an acceptable range.     

Numerical modeling of oxygen diffusion in the wall thickness of Low-Tin Zircaloy-4 fuel cladding tube during high temperature (1100-1250 °C) steam oxidation

A new numerical tool has been developed by coupling the diffusion code “EKINOX” (Estimation Kinetics Oxidation) and the thermodynamic database “ZIRCOBASE” using TQ (ThermoCalc program interface). The aim of this tool is to calculate the oxygen diffusion during oxidation of Zr based alloys. The simulation results of the oxide growth kinetics and the induced oxygen diffusion profiles in the [1100-1250 °C] temperature range, at different times, are presented and compared to previous experimental results. An estimation of the diffusion coefficient in the α phase is deduced from this comparison. The results are discussed in comparison to previous models based on an analytical treatment and the influence of the choice of the thermodynamic data set on the final oxygen diffusion profile is explored.Finally, it is shown that the present modeling is able to predict quite accurately the “critical” oxidation time corresponding to the overall ductile-to-brittle transition of the high temperature (HT) oxidized clad.     

Analysis of data and recommendations on heat transfer by liquid metals in rod bundles

Translated from Atomnaya Énergiya, Vol. 65, No. 6, pp. 423–426, December, 1988.     

Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance

Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally has displayed gas release values during irradiation three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.     

Effect of high temperature annealing on ferromagnetism induced by energetic ion irradiation in FeRh alloy

Effects of thermal annealing on ion-irradiation induced ferromagnetism of Fe-50at.%Rh bulk alloy and the related structural change were investigated by means of superconducting quantum interference device (SQUID) and extended X-ray absorption fine structure (EXAFS), respectively. Depending on the annealing temperature from 100 to 500 °C, the magnetization induced by 10 MeV iodine ion irradiation and the lattice structure of the alloy were remarkably changed. After 500 °C annealing, the magnetization and the lattice ordering of the alloy become similar to the states before the irradiation. The experimental result indicates that the thermal relaxation of irradiation-induced atomic disordering dominates the magnetic state of ion-irradiated Fe-50at.% Rh alloy.