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The velocity of delayed hydride cracking in irradiated É110 alloy (0.01% hydrogen) and zircalloy-2 (0.078% hydrogen) is predicted. It is shown that the velocity of cracking in VVÉR and RBMK fuel-element cladding is lower than in the stronger BWR fuel-element cladding. The maximum predicted velocity of cracking in É110 alloy is 2·10–7 m/sec and is reached at 533 K. Below this temperature, the cracking process in both materials studied follows the Arrhenius law with activation energy 53–56 kJ/mole. 相似文献
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A. D. Efanov A. P. Sorokin A. V. Zhukov G. P. Bogoslovskaya G. A. Sorokin 《Atomic Energy》2003,95(3):601-608
The results of thermomechanical and thermohydraulic studies showing the relative effect of the deformation of fuel-element claddings and lattices in fast-reactor fuel assemblies on their temperature regimes are presented. It is shown that the temperature nonuniformities in fuel assemblies largely determine the deformation of fuel assemblies and, in turn, the operating efficiency and, correspondingly, the degree of burnup of nuclear fuel in fast reactors. The increase in the efficiency of the fuel assemblies is largely due to temperature smoothing, including smoothing of local temperature nonuniformities. Various solutions to technical and structural problems can accomplish this. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(4):403-410
Abstract A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity. Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs. 相似文献
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介绍了利用乏燃料组件再次辐照和γ,谱对比法确定研究堆中235U含量及其燃耗成分的方法,描述了在俄罗斯IRT-MIFI堆上对IRT-3M燃料组件进行分析测定的条件装置和实验过程,给出了相应的实验结果和不确定度评价.结果表明,用该方法分析高浓铀核燃料组件中235U的含量可以得到小于2%的不确定度. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(12):998-1013
The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28%FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. 相似文献
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根据船用压水堆临界棒位、固体可燃毒物以及核燃料物理性能随燃耗的变化规律,分析了这些参数变化对反应堆温度系数的影响,得出船用压水堆温度系数随燃耗的变化规律,即在整个燃耗寿期内,船用压水堆具有负的温度系数,但随燃耗的加深温度系数的绝对值将逐渐减小. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(10):747-763
Temperature coefficients of reactivity have been measured up to 600°C on cluster-type UO2 fuel for three kinds of 235U enrichment and on a hollow cluster of sus-cladding tubes by using a hot He gas loop in a heavy-water-moderated, pressure-tube-type critical assembly. A new experimental method has been developed which accurately eliminates the reactivity disturbance caused by heat leakage in the measurement of an extremely small change in reactivity. The fuel (fuel pellet, cladding land pressure-tube) temperature coefficients of reactivity obtained for the temperature range below 300°C are +1.00±0.04, ?3.48±0.13 and - 6.36±0.25 in the unit of l0-5% Δk/k.°C for 0.2%, 0.7% and 1.5%235U enrichment, respectively. In the higher temperature region above 300°C, each coefficient shifts to positive side by about 2x10-5 Δk/k.°C. Temperature coefficient of reactivity for the hollow cluster of sus-cladding tubes (cladding and pressure-tube) has a large constant value with positive sign, + (6.42±0.26) x 10-5 Δk/k.°C, all through the temperature range. A calculational model to analyze a hot-loop-type measurement of temperature coefficients with use of WIMS-D code was proposed and could be successfully applied to the present measurement. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(11):1081-1089
By using the Kyoto University Critical Assembly (KUCA), a series of critical experiments was performed to measure the temperature coefficient of reactivity in a light-water-moderated and heavy-water-reflected cylindrical core loaded with highly-enriched-uranium (HEU) or medium- enriched-uranium (MEU) fuel. The measurement was performed for the approximately 20 to 70°C range to examine the effects of the size of light-water region in a heterogeneous multi- region type core, the reduced 235U enrichment, and the existence of boron burnable poison (BP) on this quantity by using six types of core configurations. In all the six types of cores, there were large light-water regions at the center of core and between the outer fuel region and the heavy-water reflector region, and it was found that these light-water regions caused a remarkably positive effect on the temperature coefficient of reactivity. In the present study, the temperature coefficients of the MEU core and the core without BP were more positive than those of the HEU core and the core with BP, respectively. The size of light-water region had a larger effect on the temperature coefficient rather than the reduced 235U enrichment and the existence of BP. The negative temperature coefficient would be realized by reducing the thickness of light-water layer existed in the core. 相似文献
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Volkovich A. G. Ivanov O. P. Potapov V. N. Simirskii Yu. N. Stepalin I. A. Stepanov A. V. 《Atomic Energy》2022,131(4):202-205
Atomic Energy - Methods of determining the mass of the burned 235U from the thermal energy released by the reactor and from indirect measurements of fission product activity are considered. A... 相似文献
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《Journal of Nuclear Science and Technology》2013,50(3):211-219
An analysis was performed for the temperature coefficient of reactivity measured in the six types of light-water-moderated and heavy-water-reflected cylindrical cores containing highly-enriched-uranium (HEU) or medium-enriched-uranium (MEU) fuel, which was constructed in the Kyoto University Critical Assembly (KUCA). The purpose of the present analysis was to reveal a mechanism why a light-water region existed in the core contributes to a large positive temperature effect on reactivity. Therefore, based on the assessment of the computational method to calculate the temperature coefficient of reactivity in a multi-region type core, studies were carried out to examine each effect of three physical processes (Doppler broadening, thermal expansion and thermal neutron spectral shift) on the temperature coefficient and to separate each contribution of the multi-regions to this physical quantity. The measured temperature coefficients were approximately simulated by the calculations using the SRAC code system. The Doppler broadening caused a slightly negative effect in the MEU cores and the thermal expansion a negative effect in all the cores, whereas the thermal neutron spectral shift caused a large positive effect in all the cores. The temperature effect on reactivity in the fuel region was negative, while that in the light-water region existed in the core was positive because of the decrease in neutron absorption due to the thermal expansion and the spectral shift effects, and it became positive in the present core where large light-water regions existed in the core. 相似文献
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《中国原子能科学研究院年报(英文版)》2017,(0)
<正>There are three different analytical methods used in spent fuel assemblies burnup measurement device.They are high resolution gamma ray spectrometry,total gamma ray counting method and total neutron counting method.The high resolution gamma ray spectrometry used one 相似文献