Modelling of fission-gas release from fuel undergoing isothermal heating

Currently there is controversy surrounding the mechanism of fission-gas release from fuel subjected to transient heating. A recent model (SINGAR) suggests that the release mechanism is principally single-atom migration coupled with thermal resolution of atoms from intragranular bubbles. This contrasts markedly with the previous interpretation of the release in terms of biased gas bubble migration in the presence of a temperature gradient. Here we successfully model the extensive release observed during isothermal annealing of irradiated fuel samples. This is major evidence in favour of the SINGAR model since, in the absence of a temperature gradient, the bubble migration model will predict no release, contrary to observation.     

Evaluation of LWR fuel rod behavior under operational transient conditions

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding.The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8 × 8 RJ fuel rod temperatures under power ramp conditions.     

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.     

Fission gas release in LWR fuel rods exhibiting very high burn-up

Two UO₂---Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO₂. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15%) with the experimental results. No similar agreement could be obtained without the burn-up dependency of the release model.     

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

Studies of the rapid aqueous release of fission products from UO₂ and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of ¹³⁷Cs and ¹²⁹I for a number of fuels irradiated to burn-ups of 50–75 MWd/kgU. In addition a method for analysis of leaching solutions for ⁷⁹Se was developed. The results of the studies show that the fractional release of ¹³⁷Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional ¹²⁹I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association ¹³⁷Cs and ¹²⁹I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for ⁷⁹Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that ⁷⁹Se is not preferentially leached from spent fuel.     

Verification of a mechanistic strain rate equation for transient fuel modelling

The behaviour of CANDU-PHW fuel elements in a transient is very dependent upon the development of sheath strain during the transient. So that uncertainties in predictions that usually involve extrapolations from the data base are kept small, a sheath strain rate equation that reflects the physical processes that are involved in strain has been developed. Recently completed verification tests reveal, that the average error of predictions by this model is indeed small, but the standard deviation is very large. It is shown that variations in structure and dimensions, permitted by the manufacturing tolerances for the specimens, and uncertainties in the experimental measurements, can account for this scatter. Some identified deformation mechanisms are not yet well enough quantified to be included in the model, and their omission could be the reason for the average trend to overpredict slightly. It is concluded that the model is reliable for probabilistic predictions of fuel behaviour.     

Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO₂ temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.     

A new mechanistic code SFPR for modeling of single fuel rod performance under various regimes of LWR operation

A new mechanistic code SFPR for modeling of single fuel rod behavior under various regimes of LWR reactor operation (normal and off-normal, including severe accidents) is under development at IBRAE. The code is designed by coupling of two stand-alone mechanistic codes MFPR (for modeling of irradiated UO₂ fuel behavior and fission product release) and SVECHA/QUENCH, or S/Q (for modeling of Zr cladding thermo-mechanical and physico-chemical behavior). Both codes were initially designed for accident conditions (and for this reason, are rather mechanistic) and later extended to various normal operation conditions. On the base of thorough validation against various out-of-pile and in-pile experiments, development of an advanced fuel performance code for best estimate code calculations for both normal and off-normal LWR operation regimes is foreseen.     

LWR fuel rod behavior during severe accidents

Chemical interactions between UO₂ fuel and Zircaloy cladding up to 2350°C are described. UO₂/Zircaloy single effects tests have been performed with short LWR fuel rod segments in inert gas and under oxidizing conditions. The reaction kinetics of molten Zircaloy cladding with solid UO₂ fuel has been investigated with UO₂ crucibles containing molten Zircaloy. The UO₂/Zircaloy reactions obey parabolic rate laws. The oxygen uptake by solid Zircaloy due to chemical interaction with UO₂ occurs nearly as quickly as that from the reaction with steam or oxygen.To study the competing effects of the external and internal cladding oxidation under realistic boundary conditions and the influence of the uncontrolled temperature escalation due to the exothermic steam/Zircaloy reaction on the maximum cladding temperature, single rod and bundle experiments have been performed. Electrically heated fuel rod simulators, including absorber rod material (Ag, In, Cd alloy), guide tubes and grid spacers are used. The maximum measured cladding temperature during the temperature escalation was about 2200°C. The failure temperature of the absorber rods and the extent of bundle damage depends on the guide tube material (Zircaloy or stainless steel) and varies between 1200 and 1350°C. The molten materials and liquid reaction products can relocate and form large coherent lumps on solidification, which may result in complete blockage of the fuel rod bundle cross section. In the future, 7 × 7 bundle experiments of 2 m overall length will be performed in the new CORA facility to study, in addition, the influence of quenching on fuel rod integrity.     

Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

A model for the release of stable fission gases by diffusion from sintered LWR UO₂ fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model.     

Uranium dioxide properties for LWR fuel rods

Uranium dioxide pellets have become the most important nuclear fuel, and will remain so far a long time, with the fissile isotope ²³⁵U being replaced by PuO₂ additions. This does not significantly change the pellet properties.Uranium dioxide properties affect fuel rod performance more than previously anticipated, because UO₂ pellets show a distinct response to irradiation, and because of mechanical and chemical interaction with cladding. Here elastic and plastic behaviour, fracturing, irradiation densification, and dimensional behaviour under steady and power cycling conditions are mainly covered.     

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ∼7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.     

Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX fuel

To analyze the effect of an inhomogeneous mixture of an PuO₂ powder on fission gas release in MOX fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO₂ fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu retained in the Pu-rich particles on gas release in MOX fuel. The model shows that, for the condition of all the other remaining parameters being fixed, more gas is released in a MOX fuel for lower number density of, smaller size of, and larger fraction of Pu retained in, the Pu-rich particles. However, there exists some condition or combination of parameters for which the effect of inhomogeneity on gas release is negligible depending on the characteristics of MOX fuel. Comparison with measured data for OCOM MOX fuel shows that the present model can predict the level of gas release in MOX fuel once the release mechanism in the Pu-rich particles is known.     

改进现行轻水堆组件均匀化方法的必要性

准确获得堆内各种类型组件的均匀化参数是采用现代轻水堆堆芯分析方法进行核设计和燃料管理计算的基础,以 UO2 和 MOX燃料混合装载的 C5G7-MOX基准题的研究为例,指出对由 K.Koebke 和K.S.Smith 等人莫基、在目前商用轻水堆堆芯设计中广泛应用的等效均匀化理论加以改进是必要的,并指出采用多群模型和设法采用组件当地边界条件来产生用于空间均匀化的组件空间能谱分布是改进现行组件均匀化方法应采取的措施.     

Stress corrosion cracking of Zircaloy-4 cladding at elevated temperatures and its relevance to transient LWR fuel rod behaviour

In extensive out-of-pile experiments from 500 to 900° C it has been shown that, of all the volatile fission products in a LWR fuel rod, only iodine can cause low ductility failure of Zircaloy-4 tubing due to stress corrosion cracking up to about 800° C. The critical iodine concentration above which brittle cladding failure occurs was determined as a function of temperature in the absence and presence of UO₂ fuel. A comparison of these values with the amount expected in the fuel cladding gap during a LOCA transient shows that a clear influence of iodine on burst strain can be expected only up to 700° C. This is in agreement with the results of in-pile LOCA tests performed in the FR-2 reactor with high burnup fuel rods. Since the burst temperatures during a LOCA transient would generally be above 700° C, an influence of iodine on burst strain is not very probable in a LOCA. However, with respect to ATWS transients where the maximum cladding temperatures would be below 700° C, an influence of iodine on the mechanical properties of zircaloy can be expected.     

Calculational procedure for determining creep collapse of LWR fuel rods

Observed collapses in pressurized water reactor fuel rods have been attributed to the radiation enhanced creep of Zircaloy cladding into regions where separations in the fuel pellet stack have occurred. A computer code, COLAPX, has been written to determine the growth of ovality and the ultimate collapse of fuel rod cladding under reactor operating conditions. This paper describes the theoretical bases of this code, the finite element formulation used, the constitutive relations between the displacement fields and the element forces, and the radiation, temperature and stress dependent material model for creep of Zircaloy tubing. Comparisons of the creep rate predictions and of the ovality predictions with data from irradiated tubes and fuel cladding are presented.     

Artificial neural network modeling for fission gas release in LWR UO2 fuel under RIA conditions

A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO₂ fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.     

Evaluation of implementation of thorium fuel cycle with LWR and MSR

In order to construct a sustainable society, it is necessary to consider fairness beyond generations and between countries. It is expected that Asian countries continue growing their economy and will result consuming more energy. More CO₂ emission is not acceptable.Nuclear power has many advantages for reducing CO₂ emission. However, it still has concerns of nuclear proliferation, radioactive waste and safety. It is necessary to overcome these concerns if nuclear power is expanded to Asian countries. Thorium utilization as nuclear fuel will be an opening key of these difficulties because thorium produces less plutonium, less radioactive waste. Safety will also be enhanced. The use of molten-salt reactor (MSR) triggered by plutonium supply from ordinary light water reactor (LWR) with uranium fuel will allow implementation of thorium fuel cycle with electricity capacity of about 446 GWe around at 2050.The other important sector in a view of sustainability is transportation. Transportation is essential for economy growth. Therefore it is inevitable to reduce CO₂ emission from transportation sector. Electric vehicle (EV) will be used as a major mobility instead of gasoline engine cars. Rare-earth materials such as neodymium and dysprosium are necessary for producing EV. These materials are expected to be mined from Asian countries. It is often obtained with thorium as by-product. Thorium has not been used as nuclear fuel because it is not good for nuclear weapon and it does not have fissionable isotopes. Recent global trend of nuclear disarmament and accumulation of plutonium from uranium fuel cycle can support starting the use of thorium.Thorium utilization will help both to provide clean energy and to produce rare-earth for clean vehicle. These will create new industries in developing Asian countries. An international collaborative framework can be established by supplying resource from developing countries and supplying technology from developed countries. “THE Bank (THorium Energy Bank)” is proposed here as one part of such a framework.     

六角形轻水堆组件中子通量密度分布的计算

介绍利用穿透概率法求解二维六角形轻水堆燃料组件中子通量密度分布。子区内中子源及通量密度在空间上采用二次分布 ,子区表面通量密度在空间上采用平通量密度分布 ,在方向上采用简化 6P1近似。根据提出的模型 ,编制了TPHEX D程序 ,并对一些轻水堆六角形组件问题作了计算 ,计算结果与MC结果进行了比较 ,符合良好。本程序可用于六角形轻水堆燃料组件计算。