Prospective Power-Reactor Fuel Cycles Based on Water-Free Reprocessing of Spent Fuel

A pyroelectrochemical process for reprocessing spent fuel and fabricating granular oxides UO₂, PuO₂ or (U, Pu)O₂ from chloride melts has been developed at the Scientific-Research Institute of Nuclear Reactors for a prospective nuclear fuel cycle. The basic equipment has been developed. The basic results of a comprehensive study of fuel elements with vibrationally compacted (U, Pu)O₂ fuel for fast reactors are presented. The performance of the reactors remains high up to 30% burnup in standard BOR-60 reactor fuel assemblies and 32% burnup in experimental fuel elements. An assessment is made of the effectiveness of the pyroelectrochemical methods and vibrational compaction technology for plutonium utilization.     

Thermochemical Stability, Radiation Testing, Fabrication, and Reprocessing of Mononitride Fuel

A thermodynamic analysis and experimental investigations have shown that mononitride fuel is thermochemically stable up to 1973–2073 K, at which temperature the equilibrium vapor pressure of nitrogen does not exceed 4.5·10^–7–2.1·10^–6 MPa. It is concluded on the basis of a generalization of the data from radiation testing of mononitride fuel with burnup up to 9–10% in fast and 16.8% in thermal reactors with lineal power density from 400 to 1300 W/cm that it should operate reliably in fuel elements with helium and liquid-metal sublayers. The requirement for the impurity (oxygen and carbon) content in it is formulated. When both oxygen and carbon impurities are present simultaneously in mononitride, the mass fraction of each should not exceed 0.15%. The methods for fabricating mononitride fuel are determined by the final product of the reprocessing of irradiated fuel. Consequently, methods for fabricating mixed nitride fuel from oxides and metals are now being developed.     

Structural properties of HTGR spherical microfuel

Conclusion Table 2 and Fig. 13 give the principal characteristics, averaged over several batches, for fuel elements fabricated using the two spheroidization methods. It follows from these that there are certain differences between elements of different origin: the sol-gel particles have a smaller spread of sizes relative to the mean and are closer to a spherical shape.The difference in structural characteristics is connected with the more uniform porosity distribution in slip elements and the existence of a packed surface zone in sol-gel elements. It may be assumed that due to their higher connected open porosity, the fission products yield from slip elements will be higher for as long as the initial structure is maintained during irradiation.For identical average densities, there is a wider porosity spread in a batch of solgel particles than in slip ones. A significant correlation has been established between the size and density of slip elements: with increasing particle size, their density decreases. In other respects, the differences between elements fabricated using both methods are negligible.As statistical analysis shows, the distributions of size, porosity, and element strength closely approximate a normal distribution law, which facilitates drawing statistical conclusions in determining tolerance limits for estimating spherical microfuel quality. The nonsphericality coefficient distribution can be satisfactorily described as a Weibull distribution. As an example, in Fig. 13, along with the experimental points, curves corresponding to the theoretical distributions are given.Thus the results of comparing the quality of fuel elements obtained by the slip-mass spheroidization method and the sol-gel method confirm their usability in fabricating spherical microfuel for HTGR microfuel elements and spherical fuel elements.Translated from Atomnaya Énergiya, Vol. 67, No. 6, pp. 381–389, December, 1989.     

硅化铀的肿胀特性和发展前景综述

本文介绍了辐照引起Ｕ３Ｓｉ破裂性肿胀和Ｕ３Ｓｉ２稳定性肿胀的规律及其与显微结构变化的关系。实验证实了Ｕ３Ｓｉ２－Ａｌ板型燃料在超高通量堆中使用的可行性与制造当量铀密度７ｇ·ｃｍ＾－３燃料板的可能性。     

Radiation Conditions During the Fabrication of Experimental Fuel Elements from Reprocessed Fuel

The results of an investigation of the safety of the fabrication of experimental fuel elements with a vibrationally compacted kernels consisting of reprocessed mixed fuel are presented. Data indicating the interrelationship of the granulometric composition of the product used in the technology and the intensity of the generation of the dispersion-distributed aerosol particles are presented. The postoperation structure of the surface contamination of the equipment in the setup used for fabricating fuel elements is studied. An expression is obtained for estimating the expected flow of radioactive substances into the exhaust systems of the ventillation center. The absolute values of the flow of radioactive aerosols into the environment are determined. The results of measurements of the exposure dose rate of an experimental fuel element and data on the material balance of the fuel composition and its main components are presented. The irradiation of workers participating at all stages of the process is estimated.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 351–360, May 2005.     

Analysis of the Possibility of Using Fuel Based on Reclaimed Uranium in VVÉR Reactors

A current problem is to show that reclaimed uranium can be, in principle, brought into the VVÉR fuel cycle. The possibility of using fuel based on reprocessed uranium in VVÉR is analyzed. The requirements for the initial isotopic composition of test batches of fuel pellets with 4% effective enrichment are determined, the compensation coefficient is calculated, the requirements for monitoring the isotopic composition are determined, and the technlogy for fabricating fuel pellets from relaimed fuel is determined. It is shown that the basic neutron-physical characteristics satisfy the restrictions adopted in the VVÉR-440 and -1000 designs. The effect of radiation on the public and the environment as a result of switching to fuel fabricated from reclaimed uranium is the same as for the standard oxide fuel.     

基于中子IP板的核燃料元件间接中子CT实验方法研究

中子照相作为一种无损检测技术是分析和确定核燃料元件缺陷的重要手段。中国原子能科学研究院中子照相团队依托中国先进研究堆（CARR）中子照相测试平台,搭建了核燃料元件间接中子CT装置,并开展核燃料元件模拟件的间接三维中子成像技术研究。本文首先采用蒙特卡罗模拟方法优化确定了样品环境转移屏蔽容器的关键参数并研制出屏蔽容器,并基于该装置开展了核燃料元件模拟件的间接中子CT照相实验,从获得的三维实验数据可观测到尺寸约0.35 mm模拟芯块缺陷。实验结果表明,该装置可满足核燃料元件的间接中子CT实验检测。同时初步研究了基于IP板的间接中子成像数据处理的制约因素和方法,为后续进一步利用金属转换屏替代中子IP板等技术,真正实现乏燃料元件无损检测应用提供实验指导。     

New Generation of Microfuel-Based Fuel Elements for VVÉR Reactors

The results of development work on a new generation of fuel elements based on microfuel for VVÉR reactors using the basic data from post-reactor investigations and bench tests in experiments simulating LOCA for existing fuel elements with ceramic fuel are presented. It is shown that cermet fuel elements will make it possible to realize most fully the advantages of such fuel, specifically, to develop a sealed first loop and to simplify and reduce the cost of safety, automatic control, radiation protection, coolant puification, and other systems. For example, cermet fuel elements in VVÉR-1500 reactors can improve safety under various operating conditions, maneuverability, vibrational strength, fuel assembly lifetime, and geometric stability of fuel elements.     

Effect of Neutron Irradiation on Mechanical Properties of Niobium and Nb-Y Alloy

In a high-temperature gas-cooled reactor core, which consists of prismatic graphite fuel elements, leakage flows of coolant gas occur through gaps between blocks. Since the effects of these leakage flows on the total flow distribution are significant, their flow features must be clarified. In this paper, the leakage flows (crossflow through the interface gap between contacting fuel elements and the permeation flow through the fuel elements) in the normally stacked fuel elements were studied. In the basic experiments, leakage flow rates were measured using small-scale graphite blocks to determine the equivalent interface gap width and the permeability. The experiments using the full-scale fuel element were also carried out and the results agreed well with those of the basic experiments. Furthermore, a simple flow model was devised to predict the leakage flow in the fuel element.     

新燃料元件运输容器热工安全分析及试验验证

新燃料元件运输容器是为运输493反应堆燃料元件设计的专用设备。为保证燃料元件在运输过程中的安全性,使运输容器及燃料元件的各项性能指标符合标准GB 11806-2004的要求,对运输容器进行了热工设计计算和验证试验。通过计算与相应热工试验结果的比较,验证了运输容器热工设计的准确性。采用该运输容器运输新燃料元件,在正常运输工况和事故运输工况下可保证燃料元件的安全,完全满足GB 11806-2004的规定。     

Development and Testing of Fuel Elements with Vibrationally Compacted Oxide Fuel for Fast Reactors

The main results of a series of scientific-research and technological studies performed at the State Science Center of the Russian Federation – Scientific-Research Institute of Nuclear Reactors to substantiate the use of fuel elements with vibrationally compacted oxide fuel in fast reactors are presented. In the course of this work, the physical-mechanical and technological characteristics of granular UO₂ and UPuO₂ fuel were studied; radiation tests and materials-engineering investigations of experimental and test fuel elements were performed in BOR-60, BN-350, and -600 reactors. More than 30,000 fuel elements were fabricated. Maximum burnup 30% heavy atoms was attained in BOR-60 using fuel assemblies with the standard construction and 32.3% heavy atoms was obtained using experimental fuel elements with a collapsible fuel assembly. In testing fuel elements with vibrationally compacted UPuO₂ in BN-600, maximum burnup of 9.6% (10.8% heavy atoms for individual fuel elements) was achieved. Postreactor investigations showed that the service life of the fuel elements is determined only by the choice of the cladding material. In accordance with the concept developed at the Ministry of Atomic Energy of Russia for the utilization of weapons plutonium, the Institute set about to implement in practice a technology for converting the metallic weapons-grade plutonium into mixed uranium–plutonium oxide fuel on the basis of pyroelectrochemistry and vibrational compaction.     

3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results.     

Computational method for and analysis of the application of granular absorbers in VVÉR reactors

Questions concerning the compensation of excess reactivity in pressurized-water reactors by using consumable granular absorbers are examined. A method of computing the spatial-energy distribution of the neutrons in cells with a granular absorber is presented. The neutron-physical and thermophysical characteristics of fuel assemblies with fuel elements based on homogenized and heterogeneous arrangements of gadolinium in them are compared. It is shown that granular absorbers have certain advantages, specifically, they decrease the gadolinium content in the fuel elements and at the same time increase the total number of gadolinium-containing fuel elements in the fuel assemblies. This decreases the maximum power released in the gadolinium-containing fuel elements and the temperature of the fuel during the entire run. __________ Translated from Atomnaya énergiya, Vol. 100, No. 1, pp. 8–13, January, 2006.     

宇宙线μ用于反应堆燃料元件成像的尝试性模拟研究

建立了用于Monte-Carlo模拟的μ抽样模型。使用此模型,采用Geant4程序和Root软件,对核燃料元件的μ成像进行模拟研究。模拟成像结果显示,基于将多次库伦散射等效为单次散射的径迹重建方法,可实现核燃料元件的μ成像。     

VVER-440 fuel cycles with inert matrices for burning plutonium

The possibility of forming special fuel loads for VVER-440 that would make it possible to reduce the amount of transuranium elements in spent nuclear fuel by burning transuranium actinides is examined. Preliminary calculations are performed using the HELIOS spectral code to assess the effectiveness of the transmutation. Transmutation is modeled for two load variants: mixed uranium–plutonium oxide fuel and fuel with inert diluents. The main criterion for effective transmutation is a decrease of the mass of the transuranium elements as compared with an open fuel cycle.     

Fabrication and study of the properties of PuO2−MgO fuel pellets

Conclusions Defect-free PuO₂−MgO pellets with a density of 4.4 g/cm³ (90% of the computed density of the composition, in which the volume fractions of PuO₂ and MgO equal 15 and 85% respectively), were obtained. Work with plutonium-containing material showed that the technology developed for fabricating UO₂−MgO fuel pellets is suitable for fabricating PuO₂−MgO fuel pellets. Main Science Center of the Russian Federation — A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 82, No. 5, pp. 355–358, May, 1997.     

CFD prediction of flow and phase distribution in fuel assemblies with spacers

This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.     

缓发中子计算燃料元件破损方法研究

当燃料元件发生破损时,裂变产物会释放到主冷却剂中,引起主冷却剂放射性水平增加。根据燃料元件破损的监测数据,采用一定的计算方法,计算燃料元件破损数目,可为核电厂处理元件破损事故、确保反应堆和人员安全提供重要依据。本文对缓发中子先驱核产生、释放、迁移和探测器响应等过程进行深入研究,并对每个过程建立了数学计算模型,形成了1套根据缓发中子监测数据来计算燃料元件破损数目的方法。该方法可适用于多数反应堆的燃料元件破损数目计算。     

High power ramping in commercial PHWR fuel at extended burnup

Argentinean Atucha and Embalse NPP, both PHWR with on line refuelling with approximately 68,000 fuel elements, irradiated at standard burnup, during the refuelling the fuel changes the power close to a mathematical ideal step. Recently, Atucha core started to operate fully with slightly enriched uranium (SEU, 0.85% of enrichment) increasing the burnup more than 50%, thus producing the first irradiated data base of on line refuelling at extended burnup using commercial fuel elements.The BACO code was extensively used in Argentina to model the physical behaviour of both fuel elements, natural UO₂ and SEU. The hoop stress predicted by BACO at the inner surface of the cladding correlate very well with the fuel failure probability over a wide range of tests. Using the data base of present commercial extended burnup fuel, a simple criterion of fuel failure taking into account probabilistic and parametric analysis correlate very well with the present irradiation experience, in agreement with previous BACO experience which confirm the expected behaviour and the SEU fuel element design.     

Effect of fuel burnup on the effectiveness of ultrasonic detection of leaky fuel elements in VVER-1000 fuel assemblies

The effect of fuel burnup in VVER-1000 fuel elements on the utilization effectiveness of ultrasonic detection of leaky fuel elements is examined. It is determined that the limitations of this method are due to the interaction of fuel-element cladding with the fuel pellets. Threshold for fuel burnup in VVER-1000 fuel elements with E-110 alloy cladding, determining the application limits of ultrasonic detection of leaky fuel elements in fuel assemblies, is determined.