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1.
Analysis was conducted on the Lift-Off experiment IFA-610.1 in Halden reactor by the FEMAXI-6 code using detailed measured data in the test-irradiation. Fuel center temperature was calculated on the two assumptions, i.e. (1) an enhanced thermal conductance across the pellet-clad bonding layer is maintained during the cladding creep-out by over-pressurization, and (2) the bonding layer is broken by the cladding creep-out, and these results were compared with the measured data to analyze the effect of the creep-out by over-pressure inside the test pin. The measured center temperature rise was higher by a few tens of K than the prediction performed on the assumption (1), though this difference was much smaller than the predicted rise on the assumption (2). Therefore, it is appropriate to attribute the measured center temperature rise to the decrease of effective thermal conductance by irregular re-location of pellet fragments, etc. which was caused by cladding creep-out. 相似文献
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Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM
To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO2 fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail. 相似文献
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The OECD Halden Reactor Project is an international network dedicated to enhanced safety and reliability of nuclear power plants. The Project operates under the auspices of the OECD Nuclear Energy Agency and aims at addressing and resolving issues relevant to safety as they emerge in the nuclear community. This paper gives a concise presentation of the Project goals and of its technical infrastructure. The paper contains also a brief overview of results from the programme carried out in the time period 1997–1999 and of the main issues contemplated for the 3-year programme period 2000–2002. 相似文献
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The observation is reported of a ‘rim-type' structure with small subgrains in an advanced plutonium–uranium carbide (U0.8Pu0.2)C fuel pin, which had been irradiated in the Dounreay Fast Reactor to a burnup of 8.3% FIMA. 相似文献
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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. 相似文献
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H.U. Zwicky J. Low C. Alejano C. Casado M. Lloret J.A. Gago 《Journal of Nuclear Materials》2010,402(1):60-73
In the framework of a high burnup fuel demonstration programme, rods with an enrichment of 4.5% 235U were operated to a rod average burnup of about 70 MWd/kgU in the Spanish Vandellós 2 pressurised water reactor. The rods were sent to hot cells and used for different research projects. This paper describes the isotopic composition measurements performed on samples of those rods, and the analysis of the measurement results based on comparison against calculated values.The fraction and composition of fission gases released to the rod free volume was determined for two of the rods. About 8% of Kr and Xe produced by fission were released. From the isotopic composition of the gases, it could be concluded that the gases were not preferentially released from the peripheral part of the fuel column.Local burnup and isotopic content of gamma emitting nuclides were determined by quantitatively evaluating axial gamma scans of the full rods. Nine samples were cut at different axial levels from three of the rods and analysed in two campaigns. More than 50 isotopes of 16 different elements were assessed, most of them by Inductively Coupled Plasma Mass Spectrometry after separation with High Performance Liquid Chromatography. In general, these over 400 data points gave a consistent picture of the isotopic content of irradiated fuel as a function of burnup. Only in a few cases, the analysis provided unexpected results that seem to be wrong, in most cases due to unidentified reasons. Sample burnup analysis was performed by comparing experimental isotopic abundances of uranium and plutonium composition as well as neodymium isotopic concentrations with corresponding CASMO based data. The results were in agreement with values derived independently from gamma scanning and from core design data and plant operating records.Measured isotope abundances were finally assessed using the industry standard SAS2H sequence of the SCALE code system. This exercise showed good agreement between measured and calculated values for most of the analysed isotopes, similar to those reported previously for lower burnup ranges. Thus, it could be concluded, that SAS2H results for high burnup samples are not subject to higher uncertainty and/or different biases than for lower burnup samples, and that the different isotopic experimental measurement methods provide accurate results with acceptable precision. 相似文献
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The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour. 相似文献
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The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage. 相似文献
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A computer code RANNS was developed to analyze fuel rod behaviors in the reactivity-initiated accident (RIA) conditions. RANNS performs thermal and finite-element mechanical calculation for a single rod in axis-symmetric geometry, where fuel pellet consists of 36 equal-volume ring elements and cladding metallic wall consists of eight equal-thickness ring elements and one outer oxide element. The code can calculate temperature profile inside the rod, contact pressure generated by pellet–clad mechanical interaction (PCMI), stress–strain distribution and their interactions elaborately. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by the fuel performance code FEMAXI-6.In the present study, analysis was performed on the simulated RIA experiments in the “nuclear safety research reactor” (NSRR), FK-10 and FK-12, with high burnup BWR rods in a cold-start up condition, and stress–strain evolution in the PCMI process was calculated extensively. In the analysis, the pellet–clad bonding was assumed both in the heat conduction and in mechanical restraint. The calculated hoop strain increase was compared with the measured strain gauge data, and satisfactory agreement was obtained. Simulation calculations with broader power pulses anticipated in RIA of commercial BWR were carried out and the resulted cladding hoop stress was compared with the failure stress estimated by comparison of analysis with experimental data. 相似文献
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M. Kinoshita K. Yasunaga A. Iwase M. Sataka S. Matsumura T. Ichinomiya Y. Kaneta T. Ohnuma J. Nakamura 《Nuclear instruments & methods in physics research. Section B, Beam interactions with materials and atoms》2009,267(6):960-963
In light water commercial reactors, extensive change of grain structure was found at high burnup ceramic fuels. The mechanism is driven by bombardment of fission energy fragments and studies were conducted by combining accelerator based experiments and computer-science. Specimen of CeO2 was used as simulation material of fuel ceramics. With swift heavy ion (Xe) irradiation on CeO2, with 210 MeV, change of valence charge and lattice deviation of cations were observed by XPS and XRD. Combined irradiations of Xe implantation and swift heavy ion irradiation successfully produced sub-micrometer sized sub-grains, similar as that observed in commercial fuels. Studying components of mechanism scenarios, with first principle calculations using the VASP code, we found stable hyper-stoichiometric defect structures of UO2+x. Molecular dynamics studies revealed stability of Xe planar defects and also found rapid transport mode of oxygen-vacancy clusters. 相似文献
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The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described. 相似文献
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介绍了在实验性PWR堆上完成的深燃耗条件下测量反应性概况。用实验结论剖析了在国外核电站堆芯上应用噪声分析法对慢化剂温度系数作全燃耗期监测研究中出现的测量结果与常规方法相差2~5倍的现象。从测量公式和堆芯扰动模型图入手所作的分析结果说明,没有消除随燃耗不断增长的强自发裂变中子源干扰是产生差异的根本原因。事实说明:在多种噪声分析技术中,只有能够清除自发裂变中子源干扰的方法才能成功地应用于燃耗后堆芯的反应性测量。 相似文献
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Ken Kurosaki Yoshiyuki Saito Masayoshi Uno Shinsuke Yamanaka 《Journal of Nuclear Materials》2006,350(3):203-207
A simulated burnup UO2 based fuel (150 GWd/t) was prepared by solid-state reactions. The phase equilibria of the simulated fuel were evaluated by XRD and SEM/EDX analysis. Nanoindentation tests were performed for the simulated fuel at room temperature in air. The modulus and hardness of the matrix phase and oxide precipitates that exit in the simulated fuel were directly evaluated by the nanoindentation. 相似文献
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The conditions for attaining an operating regime with maximum burnup of nuclear fuel are examined. This can be a regime with steady continuous reloadings with the reactor replenished with fuel whose heavy atoms contain close to 50% of the fissioning material with average content 25% and burnup 35–50%. The possibility of improving the stability of the fuel elements with increasing cladding thickness and the role of different organs for compensating reactivity in a regime with steady continuous reloadings and maximum fuel burnup are examined. An assessement is made of the possibility of obtaining a negative or zero effect of reactivity with loss of coolant and developing nonproliferation measures.__________Translated from Atomnaya Énergiya, Vol. 98, No. 2, pp. 83–91, February, 2005. 相似文献
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Mariano Vela MoraAlberto Gallardo Padilla José Luis Castro PalominoLuís Antônio Albiac Terremoto 《Progress in Nuclear Energy》2011,53(4):344-353
Gamma-ray spectroscopy is an important nondestructive method for the qualification of irradiated nuclear fuels. Regarding research reactors, the main parameter required in the scope of such qualification is the average burnup of spent fuel elements. This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using 137Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. 相似文献
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The CANDLE burnup strategy, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes, is applied to the block-type high temperature gas cooled reactor. If it is successful, a burnup control rod can be eliminated, and several merits are expected. This burnup may be realized by enriched uranium and burnable poison with large neutron absorption cross-section. With the fuel enrichment of 15%, gadolinium concentration of 3.0%, and fuel cell pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half-width of power density distribution of 1.5 m. When the concentration of natural gadolinium is higher, the burning region moving speed becomes slower and the burnup becomes higher, though the effective neutron multiplication factor becomes smaller. When U-235 enrichment is higher, the effective neutron multiplication factor becomes larger, the speed becomes slower, and the burnup becomes higher. When the pitch is wider, the effective neutron multiplication factor becomes larger, the speed becomes faster, and the burnup becomes higher. 相似文献
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Masaki Inoue Koji Maeda Kozo Katsuyama Kosuke Tanaka Kenji Mondo Masaki Hisada 《Journal of Nuclear Materials》2004,326(1):59-73
Drastic evolution of fuel-to-cladding gap is observed in high burnup JOYO Mk-II driver and MONJU type uranium-plutonium oxide fuel pins. The effect of the evolution is examined from viewpoints of fuel restructuring, gaseous FP release and retention and cesium migration behaviors. Its thermal impact on fuel pin performance is also studied by one-dimensional steady state thermal analysis. Threshold condition of the evolution depends on fuel pellet characteristics, burnup and probably temperature. The evolution directly relates to as-fabricated microstructures and to gaseous FP release and retention behavior. A comparison of fuel restructuring with predicted temperature profiles indicates that, even where large residual gaps are observed, non-gaseous filler always improves the heat transfer across the gaps. 相似文献