共查询到20条相似文献,搜索用时 15 毫秒
1.
AbstractA probabilistic risk assessment (PRA) quantifies the frequency of criticality accidents during railroad transport of spent nuclear fuel casks (SFCs) in the USA. It evaluates the likelihood that undetected errors in fuel selection and/or fuel handling could result in a misloaded SFC susceptible to a criticality event following an accident during rail transport of the cask. The PRA shows that existing fuel burnup records and formal procedures for loading a SFC make the likelihood of shipping a misloaded SFC on the order of 2·6 × 10–6 per SFC. When combined with historical evidence regarding train accidents and an estimate of the likelihood that an accident could breach and submerge a SFC, the calculated frequency of criticality is below 2 × 10–12 over the 11 000 shipments that would be required to ship the spent fuel inventory generated by the current US fleet of nuclear reactors, assuming that they each operate for 60 years. 相似文献
2.
3.
Tetsuo Matsumura Akihiro Sasahara Yasushi Nauchi Toshiari Saegusa 《Nuclear Engineering and Design》2008,238(5):1260-1263
A method is developed to monitor integrity of spent fuels stored in a canister that is sealed by weld. To achieve the monitoring, Kr-85 gas was newly adopted as a kind of probe. In the case of a fuel rod defect, Kr-85 gas of the fuel rod is leaked in the canister. By detection of gamma ray (514 keV) emitted from Kr-85 outside of the canister, defected rods can be detected without unsealing the canister. The monitoring technique was developed using small-scaled mock-up experiments and simulated calculation. The result showed that Kr-85 gas leakage of about 1011 Bq is detectable under the noise gamma rays by using the detection system with collimator, which is about 10% of Kr-85 inventory in a fuel rod. Therefore, this monitoring technique is considered as an inspection method prior to transportation of spent fuel from an interim storage facility to a reprocessing plant. 相似文献
4.
5.
G. N. Vlaskin É. V. Renard E. V. Chvankin A. V. Shumakov E. A. Parfent'ev O. G. Darenskikh E. P. Markin 《Atomic Energy》1992,72(5):399-401
Bochvar All-Union Scientific-Research Institute for Standardization in Mechanical Engineering. Mayak Industrial Association. Translated from Atomnaya Énergiya, Vol. 72, No. 5, pp. 451–453, May, 1992. 相似文献
6.
7.
8.
According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility (CISF) which is to start operation in 2016. At the start of the operation of the final repository (FR), by the year 2065, transport will then take place between the CISF and the FR. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios for a maritime transportation by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for the spent fuels in Korea. And, we estimated and compared the transportation risks for these four transportation scenarios. Also, we estimated and compared the transportation risks resulting from accidents during the transportation of PWR and PHWR spent fuels by road trailers from the CISF and the FR. From the results of this study, we found that risks resulting from accidents during the transportation of the spent fuels have a very low radiological risk activity with a manageable safety and health consequences. The results of this study can be used as basic data for the development of safe and economical logistics for a transportation of the spent fuels in Korea by considering the transportation costs for the four scenarios which will be needed in the near future. 相似文献
9.
For more than 50 years, CETAMA, the Commission for establishment of analytical methods from the French Alternative Energies and Atomic Energy Commission, has provided Certified Reference Materials and Interlaboratory Comparisons for the development and validation of analytical methods in the nuclear field. In the future, the nuclear spent fuel reprocessing industry will require new standards and methods to comply with high content plutonium fuel and new extraction solvents. These standards and methods will have to be fully validated in order to ensure the quality of the analytical results obtained by the laboratories.In this context, a new 242Pu reference material, certified for its isotopic composition, has been recently produced. A novel statistical approach for data processing has been used and has led to a certified value of 0.985459 ± 0.000052 for the n(242Pu)/n(Pu) atomic ratio. In addition, an interlaboratory comparison has also been organized for the validation of a method for the analysis of DMDOHEMA, and its degradation products. This compound is considered as a new extractant candidate in the frame of separation processes for transmutation of long-lived radionuclides. The methodology and results obtained in both cases are presented. 相似文献
10.
11.
12.
A spectrometric method of identifying spent fuel assemblies according to the type of fuel elements present in them is described.
The method is based on the results of spectrometric measurements and subsequent analysis of the radiation from fission products
and the characteristic radiation from uranium in the irradiated fuel. The fuel assemblies used in the VVR-2 and OR research
reactors contained fuel elements of a different type, differing by the initial quantity of uranium contained in them. To prepare
the spent fuel assemblies for shipment to a reprocessing facility after long-time storage in cool-down pools, the assemblies
must be sorted according to the type of fuel elements present in them. The method developed for identifying the types of fuel
elements in the irradiated fuel is based on the dependence of the intensity of the characteristic radiation from uranium on
the uranium content in a fuel element. The degree of excitation of the characteristic radiation of uranium also depends on
the intensity of the radiation from fission products, which is monitored during the spectrometric measurements performed on
the irradiated fuel; ultimately, this makes it possible to sort the spent fuel assemblies. 相似文献
13.
Thermal-fluid flow analysis and demonstration test were performed for a spent fuel storage system. The commercial computational fluid dynamics (CFD) code, FLUENT was used for the numerical analysis. Effective thermal conductivities of a spent fuel assembly and a fuel basket were derived to optimize a thermal analysis model. Also, a porous model, which can simplify a complex configuration of a fuel assembly, was used in the thermal analysis. Demonstration test were performed to verify the thermal analysis method and procedure using a half scaled-down model and an electrically heated dummy fuel. The numerical analysis results were compared with the experimental data. Thermal analyses of the storage system were carried out for normal and off-normal conditions by using the verified analysis method. 相似文献
14.
The effect of the internal elements of a tokamak with different coolants on the transmutation rate of long-lived actinides is studied. The Monte Carlo method is used to calculate nuclear reactions in a homogeneous model of a blanket and in specific designs of a blanket. The neutron-physical calculations of a homogeneous model of a blanket showed that the absorption of neutrons by the central column of the tokamak and their moderation by the beryllium coating slow the transmutation rate to 35% of the initial value with the most efficient utilization of lead coolant. For water coolant, this effect is negligible. In a heterogeneous model of a blanket where water coolant is used, plutonium must be added to the actinides (50%/50%). The use of lead as the coolant will increase the transmutation rate of the actinides without using plutonium. In this case, it will be possible to reprocess spent nuclear fuel from more than 10 VVER-1000 in the JUST-T hybrid reactor. 相似文献
15.
16.
Robert D. Mariani 《Journal of Nuclear Materials》2010,404(1):25-32
Spent metallic nuclear fuel is being treated in a pyrometallurgical process that includes electrorefining the uranium metal in molten eutectic LiCl-KCl as the supporting electrolyte. We report a model for determining the density of the molten salt. Material balances account for the net mass of salt and for the mass of actinides present. It was necessary to know the molten salt density, but difficult to measure. It was also decided to model the salt density for the initial treatment operations. The model assumes that volumes are additive for the ideal molten salt solution as a starting point; subsequently, a correction factor for the lanthanides and actinides was developed. After applying the correction factor, the percent difference between the net salt mass in the electrorefiner and the resulting modeled salt mass decreased from more than 4.0% to approximately 0.1%. As a result, there is no need to measure the salt density at 500 °C for inventory operations; the model for the salt density is found to be accurate. 相似文献
17.
18.
我国乏燃料运输现状探讨 总被引:1,自引:0,他引:1
随着我国经济的持续发展,核能作为安全、清洁能源在我国能源战略中地位日益突出。在保证安全的前提下,我国核电机组按照国家规划合理增加,乏燃料的产量也将逐步增加。根据我国核电站乏燃料贮存及外运规则,以及我国核电站主要位于东部沿海,而乏燃料后处理厂处在西北腹地这一国情,必将面临乏燃料的大量、长距离及安全运输的问题。乏燃料运输作为联接核电站与后处理厂或最终处置场的纽带,在维持核燃料循环体系的正常运行上发挥至关重要的作用。对国内外乏燃料运输涉及的运输方式、运输容器、运输安全监管及事故应急体系等问题进行了分析和讨论,对我国乏燃料运输中存在问题的解决提出了建议。 相似文献
19.
20.
Christofer Willman Ane HåkanssonOtasowie Osifo Anders BäcklinStaffan Jacobsson Svärd 《Annals of Nuclear Energy》2006
An important issue in nuclear safeguards is to verify operator-declared data of spent nuclear fuel. Various techniques have therefore been assigned for this purpose. A nondestructive approach is to measure the gamma radiation from spent nuclear fuel assemblies. Using this technique, parameters such as burnup and cooling time can be calculated or verified. 相似文献