共查询到11条相似文献,搜索用时 0 毫秒
1.
《Journal of Nuclear Science and Technology》2013,50(7):644-654
A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to develop fast breeder reactors (FBRs) is tend to delay. For this purpose, an SHB FA with discharged burnup of ?80 GWd/t was investigated by utilizing MAS positively as both burnable absorbers and fissile suppliers and loading high-content Pu. It is possible to load the SHB FAs in a current PWR together with UO2 FAs and to use 2.5 times as much amount of Pu as that in a standard 1/3 MOX core. Moreover, it is found to be possible to reduce the total number of fresh FAs further from that of a high-burnup (55 GWd/t in maximum) UO2 (4.9 wt%) core and also to reduce the accumulation of MAS in the nuclear fuel cycle significantly. 相似文献
2.
《Journal of Nuclear Science and Technology》2013,50(3):316-323
Critical experiments of UO2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were aNalyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%ΔAk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. 相似文献
3.
《Journal of Nuclear Science and Technology》2013,50(5):506-513
This paper is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. To minimize plutonium proliferation concern the adoption of long-life core with no fuel radiochemical treatment on site is suggested. Current investigation relies upon light water reactor technology and plutonium-free fresh fuel. Erbium doped to uranium oxide (enrichment 19.8%) fuel is selected as the reference. Such a high enrichment is selected in attempt to approach the longest irradiation time in one batch mode. In addition to that, uranium enriched up to 20% does not consider as a nuclear material for direct use in weapon manufacture. A sequence of two irradiation cycles for the same fuel rods in two different light water reactors is the key feature of the advocated approach. It is found that the synergism of PWR and pressure tube graphite reactor offers fuel burnup up to 140GWd/tHM without compromising safety characteristics. Being as large as 8% in the final isotopic vector, fraction of 238Pu serves as an inherent protective measure against plutonium proliferation. 相似文献
4.
《Journal of Nuclear Science and Technology》2013,50(10):1119-1129
Destructive analyses for five spent fuel samples taken from a Gd bearing fuel assembly were done. The measured amounts of actinides of 234-238U, 237Np, 238-242pu 241,242m,243Am 242,244Cm, and fission products of 134Cs and 154Eu were used for evaluating the accuracy of calculation made by CASMO-MICBURN and ORIGEN-2 codes. The effect of Gd on the neutron spectrum was taken into account in the CASMO-MICBURN calculation. The amounts of 235U, 239Pu and 241Pu calculated by CASMO-MICBURN agreed well with the observed values within about 3%. On the other hand, the amounts obtained from ORIGEN-2 calculation showed lower values than those observed, especially by —12% in average in 235U for Gd203U02 fuel. The main cause of this large difference may be attributed to the effect of Gd on the neutron spectrum. The amounts of the other actinides by both calculation codes revealed no significant difference in nearly 10% except for 242mAm, in which a large fluctuation among the samples was observed. About 10% difference between the measured values and the calculated values was also observed for 134Cs, but the calculated values for 154Eu showed a significant difference from measured values. 相似文献
5.
《Journal of Nuclear Science and Technology》2013,50(8):489-492
The amounts of 241Am, 242Cm and 244Cm were determined by means of a radiochemical technique in several specimens taken from the spent fuel of the JPDR-I. The yields of the transplutonium nuclides were examined in connection with the burnup determined by a conventional method. It was found that the burnup correlates well with the yield ratios of various transplutonium nuclides. 相似文献
6.
《Journal of Nuclear Science and Technology》2013,50(7):517-526
This paper evaluated the influence of neutron spectrum on characteristics of several equilibrium fuel cycles of pressurized water reactor (PWR). In this study, five kinds of fuel cycles were investigated. Required uranium enrichment, required natural uranium amount, and toxicity of heavy metals (HMs) in spent fuel were presented for comparison. The results showed that the enrichment and the required amount of natural uranium decrease significantly with increasing number of confined heavy nuclides when uranium is discharged from the reactor. On the other hand, when uranium is totally confined, the enrichment becomes extremely high. The confinement of plutonium and minor actinides (MA) seems effective in reducing radio-toxicity of discharged wastes. By confining all heavy nuclides except uranium those three characteristics could be reduced considerably. For this fuel cycle the toxicity of HMs in spent fuel become nearly equal to or less than that of loaded uranium. 相似文献
7.
《Journal of Nuclear Science and Technology》2013,50(10):745-761
Precise measurements of burnup and accumulated transuranium elements were carried out on specimens which were taken out from spent fuel of Japan Power Demonstration Reactor (JPDR)-I, by means of chemical analysis and γ-ray spectrometry. Results of the measurements are summarized numerically. Core configuration and operational history of the reactor are also described. 相似文献
8.
《Journal of Nuclear Science and Technology》2013,50(7):615-623
A study was performed at Los Alamos National Laboratory to explore the accuracy of several reactor analysis codes in calculating 241 Am and 243Am concentrations in light water reactor spent fuel. Calculated higher-actinide concentrations were compared to measured values from the literature for three reactor fuels. The fuel samples were taken from the Mihama Unit 3 pressurized water reactor, the Garigliano boiling water reactor, and a VVER-440. The 241Am and 243Am concentrations were calculated using the HELIOS-1.4 lattice-physics code, the ORIGEN2 burnup code, and a linked MCNP/ORIGEN2 code named Monteburns 3.01. Comparisons were made between the calculated and measured values. It was determined that all codes performed consistently well for the Mihama Unit 3 measurements (within ±5% for 241Am and ±20% for 243Am) and the Garigliano measurements (within ±12% for 241 Am and ±20% for 243Am). It was determined that the ORIGEN2 pressurized water reactor libraries are insufficient for the VVER-440 measurements. The HELIOS and MONTEBURNS codes both demonstrated good ability to calculate these isotopes for VVER-440 fuel (±10% for 241Am and ±12% for 243Am). The accuracies of these codes and the associated radiochemical measurements of these higher-actinide isotopes may be insufficient for safeguards and fuel management purposes; thus, development of new methods and modification to existing data libraries may be necessary in order to enable cost-effective safeguarding of these higher-actinide materials. 相似文献
9.
《Journal of Nuclear Science and Technology》2013,50(8):900-912
This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO2 and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management. 相似文献
10.
Melting Temperature of Irradiated UO2 and UO2-2wt%Gd2O3 Fuel Pellets up to Burnup of about 30 GWd/tU
《Journal of Nuclear Science and Technology》2013,50(6):528-533
Melting temperature of UO2 and UO2-2w/oGd2O3 fuels irradiated in a commercial LWR were determined by a thermal arrest technique in a burnup range up to approximately 30GWd/tU. No decrease in the melting temperature was observed on both UO2 and UO2-2w/oGd2O3 fuels with increment of burnup to 30GWd/tU. It was also found that the Gd2O3 addition below 2w/o has no influence on the melting temperature. 相似文献
11.
《Journal of Nuclear Science and Technology》2013,50(5):487-498
The effect of the moderator density distribution of annular flow on the fuel assembly neutronic characteristics in a boiling water nuclear reactor was investigated using the SRAC95 code system. For the investigation, a model of annular flow for fuel assembly calculation was utilized. The results of the assembly calculation with the model (Method 1) and those of the fuel assembly calculation with the uniform void fraction distribution (Method 2) were compared. It was found that Method 2 underestimates the infinite multiplication factor in the fuel assembly including the gadolinia rod (type 1 assembly). This phenomenon is explained by the fact that the capture rate in the thermal energy region in gadolinia fuel is estimated to be smaller when the liquid film of annular flow at the fuel rod surface is considered. A burnup calculation was performed under the condition of a void fraction of 65% and a volumetric fraction of the liquid film in liquid phase of 1. It is found that Method 2 underestimates the infinite multiplication factor in comparison to Method 1 in the early stage of burnup, and that Method 2 becomes to overestimate the factor after a certain degree of burnup. This is because Method 2 overestimates the depletion rate of the gadolinia. 相似文献