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.     

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.     

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.     

Advanced Doped UO2 Pellets in LWR Applications

The nuclear industry strives to reduce the fuel cycle cost, enhance flexibility and improve the reliability of operation. This can be done by both increasing the fuel weight and optimizing rod internal properties that affect operational margins. Further, there is focus on reducing the consequences of fuel failures. To meet these demands Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO₂ fuel containing additions of chromium and aluminium oxides. This paper presents results from the extensive investigation program which covered examinations of doped and reference standard pellets both in the manufactured and irradiated states.

The additives facilitate pellet densification during sintering and enlarge the pellet grain size. The final manufactured doped pellets reach about 0.5% higher density within a shorter sintering time and a five fold larger grain size compared with standard UO₂ fuel pellets. The physical properties of the pellets, including heat capacity, thermal expansion coefficient, melting temperature, thermal diffusivity, have been investigated and differences between the doped and standard UO₂ pellets are small.

The in-reactor performance of the ADOPT pellets has been investigated in pool-side and hotcell Post Irradiation Examinations (PIEs), as well as in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced fission gas release, improved PCI performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. Fuel segments have been exposed to ramp tests and enhanced power steady-state operation in the Studsvik R2 reactor after base-irradiation to above 30 MWd/kgU in a commercial BWR. ADOPT reveals up to 50% lower fission gas release than standard UO₂ pellets. The fuel degradation behaviour has been studied in two oxidizing tests, a thermal-microbalance test and an erosion test under irradiation. The tests show that ADOPT pellets have a reduced rate of fuel washout, as compared to standard UO₂ pellets.

Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA-96 Optima2 reloads in 2005.     

The spacing of intergranular fission gas bubbles in irradiated UO2

A simple diffusion model is used to demonstrate that irregularly spaced intergranular fission gas bubbles tend to migrate towards regions of unoccupied grain boundary. In consequence it is concluded that coalescence of these bubbles by random migration is unimportant and that, in modelling gas release and swelling, the grain face bubbles may be assumed to remain descrete until they can no longer be accommodated without overlapping.     

A methodology to predict a fission gas release ratio of MOX fuel with heterogeneous microstructure

The objective of this study is to formulate a methodology to predict a fission gas release ratio of MIMAS MOX. An irradiated MIMAS MOX fuel with plutonium rich agglomerates was subjected to elemental analyses by electron probe micro analysis and secondary ion mass spectrometry in order to investigate xenon distribution. The results of the elemental analyses showed that the plutonium rich agglomerates at the periphery of the fuel pellet sample retained a high concentration of xenon as gas bubbles. Then, the results were used as reference data for modification of models in a fuel rod analysis code, FEMAXI-7. Using the modified FEMAXI-7, we applied an approach to prediction of fission gas release ratio of MOX fuel with plutonium rich agglomerates. In the approach, two separated analyses using FEMAXI-7 were performed for the plutonium rich agglomerates and the matrix. Fission gas release ratios obtained from the two analyses were processed through weighted-average with burnup ratios of the plutonium rich agglomerates and the matrix. Finally, the fission gas release ratios were compared with results of rod puncture tests. As a result of the comparison, it was confirmed that the proposed approach could well predict fission gas release ratio of MOX fuel with plutonium rich agglomerates.     

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.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

A mass spectrometry method for quantitative and kinetic analysis of gas release from nuclear materials and its application to helium desorption from UO2 and fission gas release from irradiated fuel

A new system has been developed to determine absolute quantities of gas (mainly noble gases) released during thermal desorption in the range from 10^?12 to 10^?5 mol with a precision of few percent. The system is actually designed for simultaneous measurement of gaseous elements like He, Xe, Kr, thermally released from nuclear fuel samples and also allows the determination of the release kinetics as a function of time. This system, called Quantitative GAs MEasurement System (Q-GAMES), is based on the principle of collecting, purifying and spiking the sample gas in a “high-pressure” chamber, and continuous sampling of the gas for mass spectrometric analysis without sample depletion during the experiment. It is equipped with its own spike generator and with different gas purification systems. It is shown that this system fulfills the requirement to work with two existing very high-temperature gas desorption facilities for nuclear materials. This paper describes the Q-GAMES principle, the spiking system, its calibration, its operative mode, the different quantification techniques, as well as its technical data, in combination with some examples of typical application.     

Irradiation swelling, creep and thermal-stress analysis of LWR fuel elements, computer code isune-2

The irradiation swelling, creep, and thermal-stress analysis of light-water reactor (LWR) oxide (UO₂) fuel elements is analysed. The analysis is based on the basic physical and mathematical assumptions and the experimental data of the fuel and cladding (or canning) materials. In the analysis, the nuclear, physical, metallurgical, and thermo-mechanical properties of the fuel and cladding materials under irradiation environment are examined carefully. The objectives of the paper are mainly (1) to formulate and carry out the irradiation swelling, irradiation creep, and thermal-stress analysis of fuel elements for LWR power reactors, and (2) to develop a computer code which will facilitate the computations for fuel element design, safety analysis, and economic optimization of the power reactors. In a general procedure of the analysis, the irradiation swelling, irradiation creep, temperature distribution, etc. in the fuel and cladding of the oxide fuel elements during the reactor in operation are studied. Some theoretical models and empirical relations (on the basis of accepted experimental data) for irradiation swelling and creep in the fuel and irradiation creep in cladding materials are postulated and developed. Some analytical and empirical relations (based on test results) for heat generation and temperature distribution in the fuel during fuel restructuring are derived. The fuel restructure is, in general, divided into the central void, columnar grain, equiaxed grain, and unaffected grain zones (or regions) after a sufficiently long period for the fuel elements to be irradiated (or operated). From these relations derived for irradiation swelling, irradiation creep, and temperature distribution in the fuel and cladding, together with the well-known strain-stress, incompressibility, compatibility, and stress equilibrium equations, the irradiation swelling, creep, and thermal-stress analysis for the LWR fuel elements can be carried out.From the analytical results obtained, a computer code, ISUNE-2 (which is in the sequence of computer code ISUNE-1 and -1A developed and used previously for liquid-metal fast breeder reactor fuel element design and safety and economic analysis), can be developed. With some reliable experimental data (measured during fuel elements in operation) as input, the computer code may predict various cases of LWR (oxide or carbide) fuel elements in operation. The general scope and resulting contribution of this paper is to provide a realistic analysis and a reliable operating LWR fuel element code for use by nuclear power utilities to predict the fuel element behavior in power reactors. The fuel element design, safety analysis, and economic optimization depend largely on the fuel element behavior in the power reactors.     

Effect of non-equilibrium fission gas and fuel creep on swelling and release in irradiated carbide fuels

The code UCSWELL was developed to simulate fission gas behavior in carbide fuels. In the present work, one of the limiting assumptions in UCSWELL - that matrix gas bubbles are in equilibrium with gas atom concentration - is removed and non-equilibrium matrix fission gas bubbles are allowed, but with relaxation to equilibrium by means of vacancy diffusion and thermal and radiation-induced creep of the fuel. For a given grain size, the difference in swelling between equilibrium and non-equilibrium with relaxation bubble fission gas treatment increases with decreasing irradiation temperature. At a given temperature, the non-equilibrium effect is more pronounced for larger grain fuel. This is to be expected because the creep rate (and hence the rate at which bubbles grow to an equilibrium size) decreases as temperature decreases and/or as grain size increases. At temperatures, where the creep rate is grain size insensitive, grain size remains important to the equilibrium process in so far as the grain boundary is a source of vacancies to the non-equilibrium bubbles. While the difference in these quantities is at the most on the order of 20% for the steady operating conditions considered, it is anticipated that the non-equilibrium effects become more pronounced during reactor overpower and undercooling transients.     

Re-solution effects and fission gas swelling in UO2

A fission gas swelling model is proposed which enables one to calculate swelling in the vicinity of grain boundary networks and in imperfection-free regions. The grain boundary swelling requires a knowledge of the gas accumulation and the reaction-rate at the boundary. The gas accumulation was calculated by deriving a modified form of Fick's second law wherein it was assumed that because of re-solution effects the in-pile diffusion coefficient can be described as a function of the gas concentration but is independent of the actual operating time. Reaction-rates for bubbles at grain boundaries were derived in the manner discussed by de Jong and Koehler in their treatment of vacancy clustering. The results indicate that there is a grain size of about 10⁻⁴ cm for which the swelling is a maximum, which increases somewhat with irradiation temperature and with depletion at a constant temperature. The results enable one to predict the swelling and the mean radii of both intergranular and intragranular bubbles. Mean bubble radii predicted using the re-solution swelling model are in reasonable agreement with radii obtained from electron micrographs of irradiated UO₂ fuel samples. It is argued that gas bubble migration is the predominant means by which gas atoms arrive at grain boundaries at irradiation temperatures above about 900°C.     

大晶粒UO2燃料芯块堆内试验组件的研制

为了开发高性能的压水堆燃料,研制了大晶粒燃料芯块。试验燃料芯块具有高的235U富集度、小直径和大晶粒尺寸的特点。通过堆内辐照试验可以对不同制造工艺的燃料芯块进行评价和筛选,以便确定燃料制造工艺。为了在中国原子能科学研究院池式研究堆中随堆考验,设计了一种试验组件,包含四根双包壳的燃料棒。双包壳燃料棒是在外包壳内装入两根单包壳燃料棒。试验组件直接由反应堆一次循环水冷却,不设专门的冷却回路。试验组件上安装了多种堆芯测量传感器,包括燃料中心温度热电偶、自给能中子探测器和冷却剂出、入口温度热电偶,可以在线监测燃料试验参数。描述了大晶粒UO2燃料芯块的研制、试验燃料组件的研制和检验。     

Chemical state of fission products in irradiated UO2

The chemical state of fission products in irradiated UO₂ fuel has been estimated for FBR as well as LWR on the basis of equilibrium calculation with the SOLGASMIX-PV code. The system considered for the calculation is composed of a gas phase, a CaF₂ type oxide phase, three grey phases, a noble metal alloy, a mixed telluride phase and several other phases each consisting of single compound. The distribution of elements into these phases and the amount of chemical species in each phase at different temperatures are obtained as a function of oxygen potential for LWR and FBR. Changes of the chemical potential of the fuel-fission products system during burnup are also evaluated with particular attention to the difference between LWR and FBR. Some informations obtained by the calculation are compared with the results of post irradiation examination of UO₂ fuels.     

Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8–54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.     

Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

The Syrian Miniature Neutron Source Reactor (MNSR), a 30 kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5 h/day, the estimated core life is 10 years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for the ORIGEN 2 computer code was developed for the Syrian MNSR research reactor. The ORIGEN 2 predicted burn-up dependent actinide compositions of MNSR spent fuel using the newly developed data base show a good agreement with the published results in the literature. In addition, the burn-up characteristics of MNSR spent fuel was analyzed with the new data base. Finally, to study the effect of burn-up on the reactivity, the microscopic cross-sections of the fission products calculated by the WlMS code (using the number densities of fission products generated by the ORIGEN 2 code as a function of burn-up time), were used as an input for the CITATION code calculations. The results contained in this paper could be used in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the MNSR core.