Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up

In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ∼30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap.     

Comments on the threshold porosity for fission gas release in high burn-up fuels

This letter challenges two recent papers in this journal, suggesting that the high burn-up structure of LWR-fuels would evolve towards an open pore system, facilitating gas release. In contrast, recent experimental results and supporting calculations reviewed here as well as new evidence from a 3D pore-reconstruction strongly suggest that the materials in question would show closed porosity conditions and hence reduced probability of gas release, at least up to porosity fractions of about 25%. This value is most likely conservative.     

Measurement and calculation of neutron densities in BWR high burn-up fuels

A neutron-scanning device was developed for measuring accurate neutron densities of BWR high burn-up fuels up to 65 GWd tU⁻¹. Characteristic test of this device was done with a ²⁵²Cf source and adopted to measure axial distributions of neutron densities of BWR spent fuels with various enrichments (2.0–3.4%), which had been irradiated up to 60 GWd tU⁻¹ at Fukushima Daini Nuclear Power Station Unit 2(2F-2). We found the measured neutron densities were proportional to about fourth power of the corresponding burn-up values. The neutron densities calculated by the ORIGEN2.1 code and various cross section libraries showed good agreements with the measured ones in profile and absolute value except for BWR-UE file mainly based on ENDF/B-IV. The BS240J32 library based on JENDL3.2 was the best among the investigated libraries.     

ADAGIO technique: From UO2 fuels to MOX fuels

The amount of gas at the grain boundaries plays an important role in the fuel transient behaviour during accident conditions, such as a loss-of-coolant accident (LOCA) or a reactivity-initiated accident (RIA). Direct experimental determination of the grain boundary gas inventory has been performed for MOX fuel irradiated in an EDF pressurised water reactor (PWR) using the ADAGIO technique (ADAGIO is a French acronym meaning ‘Discriminatory Analysis of Accumulated Inter-granular and Occluded Gas’). The ADAGIO protocol applied to a MOX MIMAS fuel produced inter-granular gas fraction results that were consistent with those reached with other methods of evaluation i.e. electron probe microanalysis (EPMA). Furthermore, a new methodology for the numerical treatment of ⁸⁵Kr release kinetics which was developed for UO₂ was applied to MOX fuels. The corresponding results evidenced two types of release kinetics. These kinetics were attributed to the inter-granular bubbles of the UO₂ matrix and the bubbles located in the restructured zones, i.e. Pu agglomerates.     

Hertzian indentation of advanced lmfbr fuels with simulated burn-up

Hertzian indentation fracture of advanced fast breeder reactor fuel materials [mixed carbonitrides, (U_0.8, Pu_0.2)C_0.8N_0.2, and nitrides (U_0.8Pu_0.2)N was evaluated to yield the fracture surface energy, γ, and the fracture toughness, K_Ic. Both technological grade fuels and fuels with added fission products to chemically simulate burn-up values of 3 and 10 at.% were used. As in previous self-diffusion studies on the same materials, identical behavior (identical critical loads, P_c for crack formation) was observed for 3 and 10% b.u. Simulated M(C, N), whereas the 10% b.u. Simulated MN showed a cracking behavior identical with that of the undoped MN. In contrast, the 3 at.% b.u. Simulated MN showed lower P_c values. This is compatible with differences in fission product solubilities in these materials. The effect of fission products on γ was < 20% whereas γ increased from (U, Pu)(C, N) to (U, Pu)N by up to 80 to 90%, depending on content in fission products.     

Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

We report the measurement of elastic constants of non-irradiated UO₂, SIMFUEL (simulated spent fuel: UO₂ with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young’s modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.     

Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP–ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB.     

Towards validation of criticality calculations for systems with MOX powders

A series of experiments referred to as BFS/MOX was conducted in the BFS-1 experimental facility at IPPE, Russia. The program was designed to provide a basis for validation of criticality calculations for MOX fuel manufacturing processes and particularly those with low-moderated MOX fissile material. An extensive experimental program was performed, including criticality and reactor-type parameter measurements. The experiments were evaluated, peer reviewed, and analyzed with various codes and cross section data. The criticality validation study was performed employing a sensitivity/uncertainty technique based on the generalized linear least squares method. This paper briefly describes the experimental program, shows different tools’ performance when calculating criticality for the BFS/MOX configurations, and focuses upon the validation study and results for generic applications with weapons-grade plutonium.     

Accuracy of criticality calculations made using current nuclear data libraries for MOX fuelled fast critical assemblies

Criticality calculations have been made for a set of ten mixed plutonium–uranium oxide (MOX) fuelled fast critical assemblies using the current nuclear data libraries, JEFF-3.1, JEFF-3.1.1, JENDL-3.3 and ENDF/B-VII.0. The results obtained using the different libraries are compared and conclusions drawn concerning the accuracy of criticality calculations made for MOX fuelled fast reactors.     

核燃料的燃耗测量方法综述

介绍了非破坏性分析和破坏性分析的燃耗测量现状，论述了包括使用γ谱仪和中子探测等的多探头的乏燃料测量系统是燃耗测量的发展趋势。     

Modeling cesium behavior in nuclear reactor fuels at high temperatures

Cesium has important influences on the steady state and transient behavior of nuclear reactor fuel because of its large fission yield and high volatility. Recent experiments show that the release of cesium coincides with the disruptive behavior of rapidly heated fuel. This report investigates the pressure buildup of cesium in fuel pores during fast power transients. A model was developed which estimates the pressure of cesium in the pores as a function of temperature and oxygen/metal ratio of the fuel. The results of the calculation with this model show that cesium has a higher potential for pressure buildup, at temperatures near the fuel melting point, than xenon. There are, however, open questions which concern the kinetics of the cesium release to the pores, the chemical stability of the cesium compounds, and the microscopic distribution of the cesium.     

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO₂ LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded “laser-flash” device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO₂ of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.     

Hydrogen for synthetic fuels via nuclear energy

Fluctuations in availability and recent increases in price of petroleum have had profound effects on the national economy. As synthetic fuels, in particular, hydrogen, become increasingly attractive, nuclear energy has a role in developing such fuels. It is postulated that the nuclear radiation of the fission process itself can be utilized directly in fluid fueled devices or radiation and heat can be used in special purpose solid-fuel reactors. Both fusion and fission are considered in this light.     

Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up

New thermal diffusivity data for homogeneous SBR and heterogeneous MIMAS and OCOM MOX fuels are reported. No significant difference between the thermal diffusivity of the homogeneous and heterogeneous fuels was found at the burn-up up to 44 MWd/kgHM. These measurements, combined with previously published results or correlation functions for irradiated UO₂ and MOX were compared and it was found that separate correlations for these two fuels are not justified. A correlation for the thermal conductivity of irradiated UO₂ and MOX as a function of burn-up and irradiation temperature is proposed.     

The F4E programme on nuclear data validation and nuclear instrumentation techniques for TBM in ITER

A dedicated effort on nuclear data validation and nuclear instrumentation techniques for Test Blanket Modules (TBM) in ITER is conducted as an integral part of Fusion for Energy's (F4E) programme to ensure validated nuclear analysis capabilities for fusion technology applications. It is closely linked to nuclear data development activities, which are jointly coordinated and conducted by F4E and nuclear data consortia formed by European research institutions. The current experimental activities, an integral copper validation experiment and gas production experiments in EUROFER elements, as well as activities on the development and testing of candidate nuclear detectors for TBM in ITER and the related design integration assessment are presented in this paper.     

Oxygen potential of high burn-up fast breeder oxide fuel

The oxygen potential of fast breeder mixed oxide fuel (U_0.8Pu_0.2)O_1.98 irradiated to different burn-ups up to 11 at% has been determined between 900 and 1300 K by measurements of the electromotive force in a galvanic microcell. The oxygen potential increases continuously with burn-up, due to the oxidative nature of fission and due to fission products dissolved in the fuel matrix. Lattice parameter measurements of similar fuel indicate that the fuel with the initial O/M ratio of 1.98 is still substoichiometric even at 7% burn-up. If the lattice parameter measurements are accepted, the increase in due to fission products is larger than assumed so far.     

Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

Abstract

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel.     

MOX燃料在轻水堆核电站中的应用

目前MOX燃料已成为一种可用于轻水堆核电站成熟的核燃料。简要介绍了国外该领域的发展状况以及MOX燃料对反应堆性能的主要影响和应对措施。探讨了MOX燃料在国内压水堆核电站中的应用问题。     

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of ²³⁹Pu and ²⁴¹Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.