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。     

Optimum utilization of nuclear fuel with gas and vapor core reactors

Gas and Vapor Core Reactors (G/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Furthermore, G/VCR systems feature a low inventory and fully integrated fuel cycle with exceptional sustainability and safety characteristics. With respect to fuel utilization, there is practically no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from almost solely a burner to a breeder is achievable. The continuous recycling of fuel in G/VCR systems allows for continuous burning and transmutation of actinides without removing and reprocessing of the fuel. The only waste product at the backend of the gas core reactors' fuel cycle is fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing.

G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than Light Water Reactors (LWRs). In addition, the continuous transmutation and burning of actinides further reduces the quality of the fissile plutonium inventory. The low fuel inventory (about two orders of magnitude lower than LWRs for the same power generation level) combined with continuous burning of actinides, significantly reduces the need for emergency planning and the vulnerability to external threats. Low fuel inventory, low fuel heat content, and online separation of fission fragments are among the key constituent safety features of G/VCR systems.     

Spectrometric method of evaluating spent nuclear fuel characteristics

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.     

Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

Application of adjoint sensitivity analysis to nuclear reactor fuel rod performance

Adjoint sensitivity analysis in nuclear fuel behavior modeling is extended to operate on the entire power history for both Zircaloy and stainless steel cladding via the computer codes FCODE-ALPHA/SS and SCODE/SS. The sensitivities of key variables to input parameters are found to be highly non-intuitive and strongly dependent on the fuel-clad gap status and the history of the fuel during the cycle. The sensitivities of five key variables, clad circumferential stress and strain, fission gas release, fuel centerline temperature and fuel-clad gap, to eleven input parameters are studied. The most important input parameters (yielding significances between 1 and 100) are fabricated clad inner and outer radii and fuel radius. The least important significances (less than 0.01) are the time since reactor start-up and fuel burnup densification rate. Intermediate to these are fabricated fuel porosity, linear heat generation rate, the power history scale factor, clad outer temperature, fill gas pressure and coolant pressure. Stainless steel and Zircaloy have similar sensitivities at start-up but these diverge as burnup proceeds due to the effect of the higher creep rate of Zircaloy which causes the system to be more responsive to changes in input parameters. The value of adjoint sensitivity analysis lies in its capability of uncovering dependencies of fuel variables on input parameters that cannot be determined by a sequential thought process.     

Optimal fuel utilization during decommissioning of nuclear power plants with RBMK reactors

When RBMK reactors are decommissioned successively in the same nuclear power plant, part of the fuel of the stopped reactors can be transferred to other units and additionally burned in continuing operations. The problem of minimizing the consumption of fresh fuel by optimal distribution of the additionally burned fuel over the reactors is examined. The limitations on the refueling rates, the holding time of fuel assemblies prior to transfer, the service life of fuel assemblies, and certain characteristics of reactors are taken into account. It is shown that the reuse of fuel in other units permits saving from one to almost two thousand fresh fuel assemblies and that the effect of optimizing the additional burn regime can reach several hundreds of saved fuel assemblies. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 284–290, May, 2007.     

Analytical method of calculating the nonuniform burnup of fuel in nuclear reactors

An analytical method is worked out for calculating the nonuniform burnup of fuel in nuclear reactors in which the reactivity reserve for fuel burnup is compensated by a burning-up absorber, or an absorber uniformly distributed in the active zone during the whole operating period.Translated from Atomnaya Énergiya, Vol. 16, No. 6, pp. 500–504, June, 1964     

核燃料中添加锕系元素对反应堆的影响

文章的主要目的是研究核燃料中添加MA后对不同堆型的影响以及各种堆型嬗变MA的可行性.本文采用MCNP4C程序进行模拟,结果显示核燃料中添加MA后对不同的堆型产生不同程度的影响,相对于快堆而言,热堆的反应性、中子通量以及中子能谱受MA的影响很大.研究表明快堆和高通量热中子堆在嬗变MA核素方面具有很高的研究价值.     

Determination of the optimum fuel burn-up and energy intensities of nuclear fuel by the method of cost calculations

This report gives the procedure for determining the economical efficiency of the utilization of nuclear fuel in a reactor on the basis of calculated costs. The expression obtained for the fuet constituent of the costs of production of electrical energy enables one to make deductions regarding the aggregate effect of the cost of the fuel charge on the efficiency of utilization of nuclear fuel as well as the cost of the fuel component of the electrical output. The application of this procedure to a reactor of the type used at the Beloyarsk Nuclear Power Station is demonstrated by an example in which the economicaI optimum as well as the nuclear fuel burn-up are determined.Translated from Atomnaya Énergiya, Vol. 17, No. 2, pp. 94–97, August, 1964Published series of discussions.     

核燃料循环设施物项安全分级方法研究

简单合理的物项安全分级,不仅可以提高设施的安全性,而且还可以减少审评双方的分歧,降低营运单位和设计单位的工作量。在分析国内外核动力装置采用核安全功能进行物项安全分级和乏燃料后处理设施采用剂量准则开展物项安全分级的基础上,研究提出了采用放射性物质包容量开展核燃料循环设施的物项安全分级的方法,并采用“未缓解释放”的事故分析方法,将放化安全一级(250 mSv)和放化安全二级(5 mSv)对应的剂量准则转化为放射性物质包容量限值。     

从第五不稳定核素系的特性论核燃料的有效利用

以第五不稳定核素系在生长时期的耗裂转化比不断增长的特性以及不同堆型中它的任－衍生核素的饱和含量比值并不相同的特性作为论述的依据,得出了目前从压水堆中取出的废燃料并未获得有效充分利用的论断。认为只须对废燃料经过去除裂变产物的后处理去污流程,即要重新作为动力堆的核燃料使用,避免了使铀钚分离以及＾２８５Ｕ再度浓集的流程。并对如何使用此再制的核燃料提出两种方案,分别适用于压水堆和以天然铀为燃料的坎杜重水堆     

How to cope with the hazards of nuclear fuel waste

Many countries are developing concepts for geological disposal to achieve a permanent method of managing nuclear fuel wastes. This paper reviews the hazardous characteristics of nuclear fuel waste and discusses, in a semi-quantitative manner, the containment and retardation potential of various engineered and natural barriers. A durable waste form would retain the vast majority of radionuclides for many tens of thousands of years. Of the small inventory of radionuclides released from the waste form, many with relatively short half-lives decay to non-hazardous levels before release from the engineered barriers. The geosphere also has considerable containment potential, depending on its geological and hydrogeological properties. Approaches to selecting suitable combinations of engineered and natural barriers are illustrated based on long-term performance assessment case studies carried out in the Canadian Nuclear Fuel Waste Management program.     

Application of fuzzy decision making method to the evaluation of spent fuel storage options

In this paper, a fuzzy decision making method is proposed which is based on the fuzzy set theory and the hierarchical structure analysis. The method employs two key concepts: linguistic variables and fuzzy numbers. The linguistic variables and fuzzy numbers are used to represent the decision-maker's subjective assessments for the decision criteria and the decision alternatives versus the decision criteria. The fuzzy mean operator is used to aggregate the decision-maker's subjective assessments and the total integral value method is used to rank the decision alternatives. As a case study, the method is applied to evaluating the spent fuel storage options in Korea.     

Application of candle burnup strategy for future nuclear energy utilization

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. If this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means about 40% of natural or depleted uranium can be utilized without either enrichment or reprocessing.

In the ideal nuclear energy utilization system, the radioactive toxicity in the environment should remain or decrease after the utilization. This requirement is very severe and difficult to be satisfied. It may take too much time for its realization. The CANDLE burnup may substitute this period. Though it is a once-through fuel cycle, the discharged fuel burnup is about ten times of the present value for light water reactors. The space necessary for final disposal can be drastically reduced. However, in order to realize such a high burnup of discharged fuels some innovative technologies should be developed. Either new material standing still for such a high burnup or intermediate recladding will be required. Especially new fuel development will take a lot of time. For the time being a small reactor with CANDLE burnup may be a good option for nuclear power generation. Even this kind of reactor requires some innovative technologies and a long period for their developments. For the first stage of CANDLE burnup the prismatic fuel high-temperature gas cooled reactor is preferable. Since the design of this reactor fits to the CANDLE burnup very well, only a little time is required for its research and development.