Distributions of density and fission products in the reaction product between irradiated MOX fuel and molten zircaloy-2

Two- and three-dimensional images were obtained by X-ray CT in the reaction product between zircaloy-2 cladding tube and MOX fuel. The gamma-ray intensity distributions in the same specimen were also obtained by gamma-ray measurements of two fission products (Cs-137 and Eu-154) and one neutron-activated nuclide (Co-60). The average values of the fuel density (about 10.5 g/cm³) and the cladding density (about 6.55 g/cm³) were obtained in the metallic phase region by evaluation of the density distributions on two-dimensional X-ray CT images. The distributions of the crushed fuel pellet and the pores were also clearly observed in the three-dimensional X-ray CT images. The following results were found from the gamma-ray measurement. First, Cs-137 was observed in the unreacted fuel region and the pore region in the metallic phase region. Second, Eu-154 was widely distributed to all regions. Finally, Co-60 was confirmed only in the metallic phase region.     

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.     

Effect of quenching on molten core-concrete interaction product

ABSTRACT

Characterization of fuel debris is required to develop fuel debris removal tools for decommissioning Fukushima Daiichi nuclear power plant (1F). Especially, knowledge about the characteristics of molten core-concrete interaction (MCCI) product is needed because of the limited information available at present. Samples from a large-scale MCCI test performed under quenching conditions, VULCANO VW-U1 were analyzed to evaluate the characteristics of the surface of MCCI product. Four samples were selected from test sections at different locations. As a result, the characteristics of the samples were found to be similar. Several corium phases, such as cubic-(U,Zr)O₂ and tetragonal ZrO₂, were detected by X-ray diffraction (XRD), but concrete-based phases, such as the crystalline SiO₂ phase, were not detected by XRD because the quantity of the SiO₂ phase was too small to be measured. The Vickers hardness of each phase in these samples was higher than that of previously analyzed samples in another VULCANO test campaign, VBS-U4. Based on a comparison between MCCI product generated under quenching condition, such as VW-U1, and gently cooled MCCI product, such as VBS-U4, the MCCI product generated under quenching condition is more homogeneous, and its hardness is higher than that of the gently cooled MCCI product.     

核电厂大破口失水事故始发严重事故的源项研究

本工作以900MW核电厂为研究对象,利用一体化安全分析程序研究大破口失水事故始发严重事故下惰性气体类、挥发类和非挥发类裂变产物释放、迁移特性及分布状况,在此基础上,计算释入环境的源项。结果表明,几乎所有的惰性气体类放射性核素均释入环境,挥发类放射性核素释入环境的份额为10-3数量级,非挥发类放射性核素释入环境的份额为10-6～10-8数量级。计算所得源项可应用于厂外后果评价。     

非能动先进压水堆核电厂严重事故下裂变产物行为研究

非能动先进压水堆核电厂在严重事故下,安全壳可能发生失效,导致大量放射性物质向环境释放。本文针对非能动先进压水堆核电厂可能发生的早期失效、中期失效、晚期失效三种释放类别,建立百万千瓦级非能动先进压水堆的事故分析模型,分别针对自动卸压系统第二级卸压阀误开启,DVI管线上发生当量直径为4英寸的破口,以及热管段发生当量直径为2英寸的破口的典型严重事故序列,在研究事故进程的基础上,分析事故下裂变产物释放和迁移的特性,重点关注惰性气体、挥发性裂变产物和非挥发性裂变产物在核电厂的分布,最终计算释入环境的裂变产物源项。本文分析结果可为严重事故管理以及厂外放射性后果评价提供支持。     

我国核电站长寿命裂变产物及超铀核素累积量预测

根据我国能源需求和核电发展状况的研究结果。对未来50年核电总体装机容量和乏燃料增长趋势进行了预测,对乏燃料中次量锕系核素、钚和几种长寿命裂变产物进行了计算,获得了累积量的预测数据。     

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.     

Effect of Pu-rich agglomerates in MOX fuel on reactivity analysis of light water reactor MOX core physics experiments

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) on the reactivity analysis of light water reactor MOX core physics experiments was studied with the continuous-energy Monte Carlo calculation code MVP II. First, the following three different models were compared in the analysis of a representative unit cell of a MOX core tested at the KRITZ reactor: a Lattice model where Pu-rich agglomerates were assumed to exist in a fixed pitch, a statistical geometry (STG) model of MVP II, and a Random model where the random distribution of Pu-rich agglomerates was directly modeled. Since the three models gave comparable results, the STG model was used in parametric calculations to systematically understand the reactivity effect depending on the characteristics of Pu-rich agglomerates. In addition, the selected unit cells composing the MOX cores and one representing MOX core tested at the EOLE criticality facility were analyzed with the measured characteristics of Pu-rich agglomerates in MOX fuel. Consequently, the reactivity differences between the calculations assuming the homogeneous Pu distributions and those considering Pu-rich agglomerates were less than 0.0005 Δk/k/k', indicating that the effect of Pu-rich agglomerates was small on the reactivity analysis of the MOX cores tested in the EOLE facility.     

秦山Ⅰ期核电厂全厂断电事故源项研究

利用MELCOR程序分析秦山Ⅰ期核电厂全厂断电事故进程中放射性裂变产物的行为，研究不同性质的裂变产物各自的释放、迁移和最终分布状况。同时计算了向环境释放的源项。这些数据可用于事故的厂外后果评价。     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.     

Design and optimization of an equilibrium reload with MOX fuel with minor actinides

In this work the design and optimization of an equilibrium core for a boiling water reactor (BWR), loaded with fuel composed of plutonium and minor actinides (Np, Am and Cm), is presented. The plutonium and minor actinides are obtained from the recycling of the spent fuel of a BWR, and are mixed with depleted uranium obtained from the enrichment tails. The design and optimization of the equilibrium reload is achieved in two steps. In the first step, the fuel assembly is adjusted and the reload pattern is designed, in order to obtain the target cycle length. In order to improve the shutdown margin, two actions were taken: to increase the boron-10 content in the control rods, and to add a burnable absorber (gadolinia) in some fuel rods. In the second step, the reload pattern, obtained in the first step, is optimized to maximize the energy, under the thermal and reactivity margins constraints; a system based on Genetic Algorithms was used in the optimization process. Results show that 5% more energy was obtained with the optimized reload.     

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.     

Estimation of the Grain Boundary Gas Inventory in MIMAS/AUC MOX Fuel and Consistency with REP-Na Test Results

In the frame of its research activities on fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” performed the REP-Na program in the CABRI reactor devoted to the study of Reactivity Initiated Accidents. Focused on high burn-up UO2 and MOX fuel behaviour, twelve tests (8 UO2 and 4 MOX) were realized from 1993 to 2000. In all these tests, the influence of grain boundary gas was evidenced and it appeared necessary to perform some estimation of its inventory in irradiated fuel. Such evaluations are presented for the MOX MIMAS/AUC fuel, based on two different approaches: “experimental” and “theoretical.” The fission gas amount located at the grain boundaries increases with burn-up in correlation with the production, but also with the initial Pu enrichment, as soon as the agglomerates have reached the full restructuring threshold for the High Burn-up Structure. The consistency with the REP-Na test results is checked, showing that a significant cladding deformation is needed, clearly higher than for UO2 fuel in order to release all the grain boundary gas in RIA. Furthermore, to the fission gas effect, adds the helium's occluded in the irradiated fuel whose amount increases with burn-up, Pu enrichment and ²⁴¹Pu and ²⁴¹Am initial content.     

Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2®) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 °C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO₂ matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO₂ grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH)_4(am) phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.     

Processing and characterization of surrogate nuclear materials with controlled radial porosity

Irradiated fuel pellets present radial gradient porosity. CeO₂ has been proven as a surrogate material to understand irradiated mixed oxide (MOX) due to its similar structural and mechanical properties. A novel compaction device was developed to produce CeO₂ cylindrical pellets with controlled radial porosity. Three blends of CeO₂ with different binder contents (0.5, 3 and 7.5 vol.% of ethylene-bis-stearamide, EBS) were prepared and used to obtain three different porosities for the core, intermediate and outer rings of pellets, respectively. Different compaction pressures were employed in each region to get the intended porosities. The whole pellet was subjected to a heating rate up to 500 °C to remove the EBS binder. Finally, a pressureless sintering step was performed at 1700 °C for 4 h. A microstructural characterization was performed in the three areas, including grain size and porosity. Mechanical properties like hardness, fracture toughness and tribo-mechanical response, as scratch resistance, were also determined. Pellets fabricated from this device have shown microstructural and mechanical properties with a good correlation to those of irradiated nuclear fuel.     

Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

In order to improve LWR source term under severe accident conditions, the first version of a fission product chemistry database named ‘ECUME’ was developed. The ECUME is intended to include several datasets of major chemical reactions and their effective kinetic constants for representative severe accident sequences. It is expected that the ECUME can serve as a fundamental basis from which fission product chemical models can be elaborated for use in the severe accident analysis codes. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system from 300 to 3000 K. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate fission product chemistry in Cs-I-B-Mo-O-H system under LWR severe accident conditions, where kinetics effects should be considered.     

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.     

Present status and practical issues on dosimetry for the lens of the eye at JAEA MOX fuel facilities

ABSTRACT

At Japan Atomic Energy Agency (JAEA) MOX fuel facilities, a worker usually wears a protective lead apron; therefore, the dose to the lens of the eye (lens dose) outside the apron is higher than that to the torso. To estimate the potential impact on the current facility operation of the International Commission of Radiological Protection (ICRP)-proposed lens dose limit reduction from 150 mSv/y to average 20 mSv/y, the authors carried out an analysis on the past dose records for the workers over the last 18 years. Of a total of 4,312 workers’ records analyzed, two workers’ annual lens doses exceeded the lowered limit of 20 mSv (23.3 mSv and 20.7 mSv), although the maximum effective dose was below 10 mSv in each case. These compiled dose data reveal that in the glovebox and related operations the lens dose will be a limiting factor in radiological control under the newly lowered dose limit. To ensure that the number of workers with an annual lens dose greater than 15 mSv (approximately 0.6% of the workers) is kept to a minimum, the implementation of an administrative control level for the lens dose is considered.     

Feasibility Study on the Analysis of MOX Particle Behavior

A plenty of plutonium is dealt in Plutonium Fuel Fabrication Facility and the facility is required to confine plutonium within a limited space such as glove box (GB) because plutonium is a-emitter and causes an internal exposure. The MOX particles entrainment occurs and some of them are transiting to the outlet of GB without deposition to floor and wall. The entraining rate and the transiting rate are reported as Airborne Release Fraction (ARF) and Respirable Fraction (RF) in the literatures. However, no activities of model development and analytical approach have been found for ARF and RF. Thus, a feasibility study is done in this paper on the behavior of MOX particles in GB such as entraining and transiting. A modeling code has been developed by improving AQUA-SF code and the RF values for abnormal occurrences, such as free-fall spill, outflow and fire, have been analyzed and compared with those reported. This paper also shows the analytical results of the improved code together with the simulated experimental results. It is found that the calculated values are almost corresponded to those reported and that the improved code can estimate MOX particle behavior in GB well.