Behavior of Irradiated ATR/MOX Fuel under Reactivity Initiated Accident Conditions

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20MWd/kgHM were conducted at the NSRR in Japan Atomic Energy Research Institute to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Relatively large radial deformation of the fuel rods due to pellet-cladding mechanical interaction occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet.     

A model for fission gas release and gaseous swelling of the uranium dioxide fuel coupled with the FALCON code

A comprehensive model GRSW-A was developed to analyse the processes of fission gas release, gaseous swelling and microstructural evolutions in the uranium dioxide fuel during base irradiation and under transient conditions. The GRSW-A analysis incorporates a number of models published in open literature, as well as some original models that were already published by the authors elsewhere. Consequently, only the most prominent aspects of GRSW-A and its coupling with the FALCON fuel behaviour analysis and licensing code are described in this paper. The analysis of fuel behaviour in the REGATE experiment is presented, which includes the base irradiation of the fuel segment in a PWR to a burn-up of about 50 MWd/kgU, which was followed by a power ramp in the SILOE research reactor. Besides, the generalized data on fission gas release (FGR) in PWR fuel during the base irradiation up to a burn-up of about 70 MWd/kgU is interpreted using coupled FALCON and GRSW-A. Moreover, a mechanistic interpretation of the published data for pellet swelling during the base irradiation up to a burn-up of 100 MWd/kgU is put forward. In all the cases, the coupled FALCON/GRSW-A analysis has shown the improved prediction capability compared to the original FALCON MOD01, which is achieved due to the account for the mutual effect of thermal and, in particular, high-burn-up-assisted mechanisms of fission gas release and swelling under steady-state and transient conditions.     

Reduction of fission gas release at extended burnups in light water reactors by decreasing fuel temperature

The effects of fuel temperature on fission gas release in light water reactor UO₂ fuel at extended burnups of up to 56 effective full power months (EFPMs) are evaluated using a simple fission gas release mechanistic model. The model is first described and then model validation comparisons are made against experimental fission gas release date. The study shows that by decreasing the maximum operating fuel temperature to below 1200°C, it is possible to reduce the amount of released fission gas at 56 EFPMs to less than that at the current design burnup of 36 EFPMs.     

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.     

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.     

Experimental Observations on Fuel Pellet Performance at High Burnup

The Japanese and Spanish nuclear industries have conducted joint experimental programmes since early 1990's to address fuel performance issues such as fuel volume change and fission gas release. These efforts have produced large amount of valuable information on in-reactor performance of fuel materials representing current and potential future fuel designs. A large number of thoroughly characterised fuel rods composed of different materials have been irradiated in the Spanish PWR Vandellós II for up to five irradiation cycles achieving rod average burnup of up to 75 MWd/kgU.

This paper looks into the fuel pellet performance at high burnup only based on the extensive PIE programme both on-site and in hot-cells carried out over this fuel and other related data on similar fuel rods thus supporting and enriching the conclusions.     

Stress biaxiality in high-burnup PWR fuel cladding under reactivity-initiated accident conditions

In order to promote a better understanding of failure mechanisms of high-burnup pressurized water reactor (PWR) fuels under reactivity-initiated accident (RIA) conditions, stress biaxiality in cladding has been estimated for the pellet-cladding (PC) mechanical interaction (PCMI) phase. The estimation was based on an analysis of the transient elongations of a pellet stack and a cladding tube measured in RIA-simulating experiments in the nuclear safety research reactor (NSRR) using the RANNS code. Stress biaxiality in the high-burnup PWR fuel cladding during the PCMI phase has been estimated to be 0.7–0.8, on average, at the mid-wall of the cladding. A comparison with fresh fuel test results and a sensitivity analysis showed that the effects of burnup and pulse width on cladding stress biaxiality are less than 10% for the investigated range. The present analysis also indicated that PC friction is strong, and that the cladding constraint on pellet stack elongation is significant irrespective of burnup. Therefore, it is recommended that strong PC friction be assumed, which is similar to the mechanical bonding condition, and that fuel pellets be treated as deformable materials in models of fuel behavior during the PCMI phase.     

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.     

The OECD Halden reactor project fuels testing programme: methods, selected results and plans

The fuels testing programme conducted in the Halden reactor (heavy boiling water reactor (HBWR)) is aimed at providing data for a mechanistic understanding of phenomena, which may affect fuel performance and safety parameters. The investigations focus on implications of high burnup and address thermal property changes, fission gas release as influenced by power level and operation mode, fuel swelling, and pellet–clad interaction. Relevant burnup levels (>50 MWd kg⁻¹ U) are provided through long-term irradiation in the HBWR and through utilisation of re-instrumented fuel segments from commercial light water reactors (LWR). Both urania and MOX fuels are being studied regarding thermal behaviour, conductivity degradation, and aspects of fission gas release. Experiments are also conducted to assess the cladding creep behaviour at different stress levels and to establish the overpressure below which the combination of fuel swelling and cladding creep does not cause increasing fuel temperatures. Clad elongation measurements provide information on the strain during a power increase, the relaxation behaviour and the extent of a possible ratcheting effect during consecutive start-ups. Investigations foreseen in the programme period 2000–2002 include the behaviour of MOX and Gd-bearing fuel and other variants developed in conjunction with burnup extension programmes. Some LWR-irradiated fuel segments will undergo a burnup increase in the HBWR to exposures not yet achieved in LWRs, while others will be re-instrumented and tested for shorter durations.     

液态燃料氯盐快堆最小需求燃耗分析

为探究采用增殖燃烧模式运行的液态燃料氯盐快堆的平均卸料燃耗深度，基于中子平衡分析方法，选取5种常用氯盐，提出在线清除裂变气体和难溶裂变产物方案来维持增殖燃烧运行模式，主要研究分析了氯盐的重金属密度和在线处理方案对最小需求燃耗的影响以及无限栅元模型下维持增殖燃烧模式可接受的堆芯中子损失项。分析表明68NaCl-32UCl₃和20UCl₃-80UCl₄的最小需求燃耗分别是30.47%FIMA（FIMA是指已裂变原子数与初始的总装料金属原子数之比）和10.28%FIMA；清除裂变气体和难溶裂变产物后，60NaCl-40UCl₃可接受的中子损失项从3.49%提高到10.68%。结果表明氯盐的重金属密度对最小需求燃耗有明显影响，同时清除裂变气体和难溶裂变产物能够较大提高燃料盐系统的中子经济性，以及提高增殖燃烧模式运行可接受的堆芯中子损失项。      

An investigation on recycling the recovered uranium from electro-refining process in a CANDU reactor

Feasibility studies for recycling the recovered uranium from electro-refining process of pyroprocessing into a Canada Deuterium Uranium (CANDU) reactor have been carried out with a source term analysis code ORIGEN-S, a reactor lattice analysis code WIMS-AECL, and a Monte Carlo analysis code MCNPX. The uranium metal can be recovered in a solid cathode during an electro-refining process and has a form of a dendrite phase with about 99.99% expecting recovery purity. Considering some impurities of transuranic (TRU) elements and fission products in the recovered uranium, sensitivity calculations were also performed for the compositions of impurities. For a typical spent PWR fuel of 3.0 wt.% of uranium enrichment, 30 GWD/tU burnup and 10 years cooling, the recovered uranium exhibited an extended burnup up to 14 GWD/tU. And among the several safety parameters, the void reactivity at the equilibrium state was estimated 15 mk. Additionally, a simple sphere model was constructed to analyze surface dose rates with the Monte Carlo calculations. It was found that the recovered uranium from the spent PWR fuel by electro-refining process has a significant radioactivity depending on the impurities such as fission products.     

UO2燃料裂变气体渗流模型研究

为分析UO₂燃料晶界气泡连通导致裂变气体间歇性释放的动力学过程,从而解决目前扩散模型预测的沿芯块径向释放份额与实验测量不符的问题,采用二维渗流模型模拟UO₂燃料晶界气泡网络的演化及与燃料棒内自由空间连通的释放过程。研究结果表明,渗流模型预测沿芯块径向的裂变气体释放份额在芯块中间部分出现局部峰值,并随着时间向芯块外侧推进,与辐照试验观察到不同燃耗下径向裂变气体分布现象定性符合。因此,本研究建立的渗流模型能够从机理上解释此前扩散模型未能预测的UO₂燃料裂变气体释放份额沿径向非单调分布现象。      

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.     

The radial distribution of fission gases and other fission products in irradiated PWR fuel

The radial distribution of fission gas (xenon) and other fission products (cesium, ruthenium, cerium) has been measured on UO₂ fuel pellets irradiated in commercial pressurized water reactors to burnups between 13.23 and 48.26 GWd/tU. Fission gas release occurs from the pellet center, and at temperatures $\text{< 1300° C}$ is confined to the region of grain growth. The maximum fractional release measured at the center ranges from 20% to 30%. Only at high burnup (48.26 GWd/tU) an additional release of cesium has been observed. This is considered as evidence for an increase in fission product release at higher burnups. At fuel center line temperature $\text{> 1500° C}$ a high fission gas release is accompanied by a high cesium release. The local release starts at the onset of fission gas bubbles precipitating on grain boundaries and saturates in the center of the pellet at a fractional release value of about 90%.     

Calculation of radial burnup and nuclides atom density distributions in a VVER-1000 fuel rod using Monte Carlo method

A fundamental knowledge of fuel behavior in different situations is required for safe and economic nuclear power generation. Due to the importance of a fuel rod behavior modelling in high burnup, in this paper, the radial distribution of burnup, fission products, and actinides atom density and their variations by increasing burnup and other factors such as temperature, enrichment and power density are studied in a fuel pellet of a VVER-1000 reactor in an operational cycle using the MCNPX 2.7 Monte Carlo code. A benchmark including a Uranium-Gadolinium (UGD) fuel assembly is used for verification of the developed model in the MCNPX code for radial burnup calculation. A sensitivity study is carried out to investigate the effect of different parameters such as the number of particles per cycle, the number of geometrical radial nodes in the fuel pellet, the number of burnup steps and the selection of different fission-product contents (i.e. those isotopes that are used for particle transport) on the MCNPX model for speed and accuracy compromising. To calculate the radial temperature profiles and to analyze the effect of temperature on the radial burnup distribution and vice versa, the HEATING 7.2 code, which is a general-purpose conduction heat transfer program, and the MCNPX code are applied together. The results show the accuracy and capability of the proposed model in the MCNPX and HEATING codes for radial burnup calculation.     

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.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.     

高燃耗下裂变气体释放行为研究与程序校验

采用燃料棒性能分析程序COPERNIC，针对哈尔登（Halden）测试燃料组件 （IFA）519.9 DK 辐照试验燃料棒辐照试验进行了计算分析，研究了高燃耗下裂变气体释放行为，并与试验数据进行了对比验证。结果表明，在燃耗达到约100 GW?d/t(U)的辐照过程中，该程序对裂变气体释放率的预测值与试验测量结果符合较好；程序未精确预测芯块孔隙率在高燃耗“边缘结构”内的演化过程，但不影响其对燃料棒辐照综合性能分析的准确性和合理性。      

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.