Inert matrix fuel — A new challenge for material technology in the nuclear fuel cycle

An innovative nuclear fuel concept for the utilization as energy resources and for the incineration of excess Pu arisings as well as for an effective transmutation of minor actinides (MA's; Am, Np and Cm) is discussed from the aspect of material technology. Stabilized cubic phase ZrO₂ and other potential candidate materials for the Inert Matrix are compared in terms of the material properties and other behaviors such as the behavior against irradiation with the relevant information currently available. Strategies for the use of the Inert Matrix Fuel concept in various countries are discussed and compared for their options in nuclear fuel cycle technology. This article based on a presentation made in the “Symposium on Nuclear Materials and Fuel 2000” held at the Korea Atomic Energy Research Institute (KAERI), Taejon, Korea, August 24–25, 2000 under the auspices of the Ministry of Science and Technology (MOST).     

Biofouling on austenitic stainless steels in spent nuclear fuel pools

The objective of this study was to investigate the biofilm formation on three different types of austenitic stainless steel (UNS S30400, S30466 and S31600) submerged in a spent nuclear fuel pool. The presence of microorganisms in coupons was characterised using standard culture microbiological methods, microscopic techniques (epifluorescence microscopy and scanning electron microscopy), and molecular biology techniques (denaturing gradient gel electrophoresis and sequencing fragments of 16S rDNA). The microscopy techniques showed signs of colonisation of stainless steels in spite of these extreme conditions. Based on sequencing of cultured microorganisms, different bacteria belonging to α, β, γ‐Proteobacteria, Bacilli, and Actinobacteria classes have been identified. The biofilm radioactivity was measured using gamma‐ray spectrometry and, according to the data gathered, the radionuclides present in the water pool were entrapped in the biofilm increasing the amount of radiation at the surface of the different materials.     

Isolating wastes in the electrometallurgical treatment of spent nuclear fuel

The main hazards in spent nuclear fuel are fission products and transuranic radionuclides. An electrometallurgical treatment is designed to isolate these elements by electrorefining and then place them in waste forms suitable for geologic disposal. In the highly reducing chemical environment used for electrometallurgical treatment, fuel cladding and transition-metal fission products remain as metals; these metals are collected and melted to form a highly corrosion-resistant waste form. Other fission-product elements and transuramic elements collect in the molten-salt process fluid and are removed by ion exchange into zeolite, which is further processed to make a durable-composite ceramic waste form.     

Research on graphite powders used for HTR-PM fuel elements

Different batches of natural graphite powders and electrographite powders were characterized by impurity, degree of graphitization, particle size distribution, specific surface area, and shape characteristics. The graphite balls consist of proper mix-ratio of natural graphite, electrographite and phenolic resin were manufactured and characterized by thermal conductivity, anisotropy of thermal expansion, crush strength, and drop strength. Results show that some types of graphite powders possess very high purity, degree of graphitization, and sound size distribution and apparent density, which can serve for matrix graphite of HTR-PM. The graphite balls manufactured with reasonable mix-ratio of graphite powders and process method show very good properties. It is indicated that the properties of graphite balls can meet the design criterion of HTR-PM. We can provide a powerful candidate material for the future manufacture of HTR-PM fuel elements.     

A review of TRISO-coated particle nuclear fuel performance models

The success of high temperature gas cooled reactor depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires the development of an integrated mechanistic fuel performance model that fully describes the mechanical and physicochemical behavior of the fuel particle under irradiation. In this paper, a review of the analytical capability of some of the existing computer codes for coated particle fuel was performed. These existing models and codes include FZJ model, JAERI model, Stress3 model, ATLAS model, PARFUME model and TIMCOAT model. The theoretic model, methodology, calculation parameters and benchmark of these codes were classified. Based on the failure mechanism of coated particle, the advantage and limits of the models were compared and discussed. The calculated results of the coated particles for China HTR-10 by using some existing code are shown. Finally, problems and challenges in fuel performance modeling were listed.     

Actinides in metallic waste from electrometallurgical treatment of spent nuclear fuel

Argonne National Laboratory has developed a pyroprocessing-based technique for conditioning spent sodium-bonded nuclear-reactor fuel in preparation for long-term disposal. The technique produces a metallic waste form whose nominal composition is stainless steel with 15 wt.% Zr (SS-15Zr), up to ∼ 11 wt.% actinide elements (primarily uranium), and a few percent metallic fission products. Actual and simulated waste forms show similar eutectic microstructures with approximately equal proportions of iron solid solution phases and Fe-Zr intermetallics. This article reports on an analysis of simulated waste forms containing uranium, neptunium, and plutonium. For more information, contact Dawn Janney, Argonne National Laboratory-West, P.O. Box 2528, Idaho Falls, ID 83403-2528; (208) 533-7478; fax: (208) 533-7863; e-mail dawn.janney@anlw.anl.gov.     

核燃料组件骨架点焊装置

介绍了核燃料组件骨架点焊装置的结构和主要工作原理，并给出了该机构的总示意图、控制柜示意图、气动原理图及控制原理图，扼要说明了与操作有关的问题。     

Uranium processing during the treatment of sodium-bonded spent nuclear fuel

Several key technical areas have been developed during a demonstration program at Argonne National Laboratory to advance the electrometallurgical treatment of spent nuclear fuel, including equipment operating conditions, uranium-product purity and recovery, and processing rates. Operating conditions for three integral processing steps were optimized based on minimizing the impurities to the final uranium product while meeting program goals for processing rates. The overall recovery of uranium during the process steps was influenced by process conditions as well as the production rates set forth for the demonstration program. For more information, contact B.R. Westphal, Argonne National Laboratory, Nuclear Technology Department, P.O. Box 2528, Idaho Falls, Idaho 83403-2528; (208) 533-7398; fax (208) 533-7735; e-mail brian.westphal@anl.gov.     

焊接核燃料棒的TIG焊接电源

论述的TIG焊接电源是应用在核燃料棒端塞焊接的专用焊接电源.介绍了这种电源的工作原理,其中包括了晶体管弧焊电源的工作原理以及高频引弧器的工作原理.并对焊接设备的总体结构和总体电路做了说明.该电源系统的设计成功地解决了普通TIG焊接电源存在的引弧困难、电网电压波动大、电流控削精度低等问题.     

核燃料包壳锆合金表面涂层研究进展

锆合金表面涂层是提高核燃料包壳事故容错能力的重要途径之一。本文综述了锆合金表面涂层的研究进展,包括涂层种类、制备工艺、微观组织以及抗水蒸气氧化性能、耐腐蚀性能等,介绍了锆合金表面涂层种类选择的依据,探讨了涂层的制备工艺、微观组织与性能之间的关系,分析了当前研究中存在的若干问题及未来涂层的发展方向,为进一步促进核燃料包壳锆合金表面涂层的研究提供了有价值的参考。     

A detecting method for spherical fuel elements in pebble-bed HTGR using eddy current detection

In pebble-bed high temperature gas-cooled reactors (HTGRs), spherical fuel elements move inside pipelines of a handling system, which should be controlled precisely. A detecting method for these elements is proposed in this paper, which is achieved by a detecting system with self-diagnosis function. Detecting signals are obtained by sensors installed outside pipes. A signal identification algorithm was designed for graphite ball detection. Electromagnetic simulations and detecting experiments were performed for system optimization and development of the method. The results show that the proposed method is capable for spherical fuel elements detection, and can be successfully used in practical application.     

Stainless steel-zirconium waste forms from the treatment of spent nuclear fuel

Stainless steel-zirconium waste-form alloys have been developed for the disposal of metallic wastes recovered from spent nuclear fuel using the electrometallurgical process developed by Argonne National Laboratory. The metal waste comprises the spent-fuel cladding, noble-metal fission products, and other metallic constituents remaining after electrorefining. Two nominal waste-form compositions have been slected: stainless steel-clad fuels and zirconium-8 wt.% stainless steel for Zircaloy-clad fuels. These alloys are very corrosion resistant. Tests performed with these alloys indicate favorable behavior for use high-level nuclear waste forms.     

Ceramic-composite waste forms from the electrometallurgical treatment of spent nuclear fuel

Argonne National Laboratory is developing a method to treat spent nuclear fuel in a molten-salt electrorefiner. Glass-bonded zeolite and sodalite are both being developed as ceramic waste forms. The ceramic waste form will contain the fission product (e.g., rare earth, alkali and alkaline-earth metals, halogens, and chalcogens) and transuranic radionuclides that accumulate in the electrorefiner salt. Zeolite A can fully incorporate both the salt and the radionuclides into its crystal structure. Salt-loaded zeolite A is mixed with glass frit; the blend undergoes hot isostatic pressing to produce a monolithic leach-resistant waste form. Alternatively, the salt-loaded zeolite may be converted to sodalite simply by heat treating first, then adding the glass and hot pressing.     

Application of infrared thermography to the fabrication process of nuclear fuel containers

In recent years, the use of infrared thermography based non-destructive test (NDT) methods to characterize and control manufacturing processes has been progressively increased. This method can be employed in the fabrication process of containers for nuclear fuel transport. In order to avoid undesirable bubbles, thermal gradients must be controlled during cooling process of the lead shield included in the container. Using infrared thermography the cooling process behaviours are analyzed and monitored on real time, and enabling its control. Satisfactory achievements in two different ways of cooling process are presented in this paper.     

Nondestructive technique for identification of the dished end of nuclear fuel pellets

Use of an infrared technique for identifying the dished/flat ends of nuclear fuel pellets for PHWR is suggested. The studies show that this method can be used for on-line checking of the orientation of the pellets.     

Nanosized polycrystalline diamond cladding for surface protection of zirconium nuclear fuel tubes

A 300 nm thick polycrystalline diamond layer has been used for protection of zirconium alloy nuclear fuel cladding against undesirable oxidation with no loss of chemical stability and preservation of its functionality. Deposition of polycrystalline diamond layer was carried out using microwave plasma enhanced chemical vapor deposition apparatus with linear antenna delivery (which enables deposition of PCD layers over large areas). Polycrystalline diamond coated zirconium alloy fuel tubes were subjected to corrosion tests to replicate nuclear reactor conditions, namely irradiation and hot steam oxidation. Stable radiation tolerance of the polycrystalline diamond layer and its protective capabilities against hot steam oxidation of the zirconium alloy were confirmed. Finally, the use of polycrystalline diamond layers as a sensor of specific conditions (temperature/pressure dependent phase transition) in nuclear reactors is suggested.