Fabrication of DUPIC fuel pellets using high burn-up spent PWR fuel

Technology for the direct usage of a spent PWR fuel in CANDU reactors (DUPIC) was developed in KAERI to reduce the amount of spent fuel. DUPIC fuel pellets were fabricated using a dry processing method to re-fabricate CANDU fuel from spent PWR fuel without any intentional separation of fissile materials or fission products. The DUPIC fuel element fabrication process satisfied a quality assurance program in accordance with the Canadian standard. For the DUPIC fuels with various fuel burn-ups between 27,300 and 65,000 MWd/tU, the sintered pellet density decreased with increasing fuel burn-ups. Fission gas releases and powder properties of the spent fuel also influenced the DUPIC fuel characteristics. Measurement of cesium content released from green pellets revealed that their sintered density significantly depended on sintering temperature history. It was useful to establish a DUPIC fuel fabrication technology in which a high-burn-up fuel with 65,000 MWd/tU was treated.     

Fabrication of Sintered Annular Fuel Pellet for HANARO Irradiation Test

The fabrication method of an annular pellet with highly precise diametric tolerances, same dimensions, and various sintered densities has been investigated. To examine the in-pile densification and swelling of the annular pellet, 5 different types of annular pellet were prepared for a HANARO irradiation test. In order to obtain annular fuel pellets with the same dimensions and various sintered densities, we control the green density of an annular compact, the sintering temperatures, and the sintering time. For a diametric tolerance control, we have introduced a new compaction process that combines the usual double-acting pressing and cold isostatic pressing. Annular fuel pellets with the same dimensions and various sintered densities were fabricated successfully, and all the pellets satisfied the pellet specification of the HANARO irradiation test. Sintered annular pellets show an excellent inner diametric tolerance of less than ±12 μm without an inner surface grinding.     

Remote Fabrication of DUPIC Fuel Pellets in a Hot Cell under Quality Assurance Program

The Korea Atomic Energy Research Institute (KAERI) has been developing the Direct Use of Spent Pressurized Water Reactor (PWR) Fuel in the CANada Deuterium Uranium (CANDU) Reactors (DUPIC) fuel fabrication technology since 1992, and the basic DUPIC fuel fabrication process was established in 2002. In order to demonstrate the robustness of the DUPIC fuel fabrication process through the irradiation test, it is important that a Quality Assurance (QA) program should be in place before a fabrication of the DUPIC fuel. Therefore, the Quality Assurance Manual (QM) for the DUPIC fuel was developed on the basis of the Canadian standard, CAN3-Z299.2-85. This manual describes the quality management system applicable to the activities performed for the DUPIC fuel fabrication at KAERI. In order to demonstrate the DUPIC fuel fabrication technology and produce qualified DUPIC fuel pellets, the process qualification tests were performed, which include three pre-qualification tests and three qualification tests. The characteristics of the DUPIC fuel pellets such as the sintered density, grain size, and surface roughness were measured and evaluated in accordance with the QA procedures. The optimum fabrication process of the DUPIC fuel pellet was also established based on the qualification results. Finally a production campaign was carried out to fabricate the DUPIC fuel pellets at a batch size of 1 kg following the QA program. As a result of the production campaign, qualified DUPIC fuel pellets were successfully produced and, therefore, the remote fuel fabrication technology of the DUPIC fuel pellet was demonstrated.     

Production and investigation of thorium-containing fuel compositions

A technology has been developed for obtaining fuel tablets with the compositions (U, Th)O₂, (U, Th, Ca)O₂, and (U, Th)O₂+MgO by combined precipitation of uranium, thorium, magnesium, or calcium components from inert solutions, followed by heat treatment of the powders, compression into pellets, and sintering of the pellets. Work on optimizing the technological processes for obtaining fuel pellets so as to obtain good pellet quality was performed. The effect of the properties of the precipitates and powders, fabricated using different technological regimes on the properties of the finished objects was studied. The work includes detailed investigations of powders (x-ray phase analysis, electron-microscopic investigation) and sintered fuel tablets (change in the geometric dimensions as a result of sintering, determination of the density, and study of the microstructure). The behavior of fuel compositions (U, Th)O₂ and (U, Th)O₂+MgO in contact, with coolants under conditions where the fuel elements become unsealed was studied: with water at 300°C and sodium at 700°C. 3 figures, 3 tables, 6 references. State Science Center of the Russian Federation-A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 5, pp. 346–353, May, 2000.     

Effect of Relative Humidity on Growth of Sodium Oxide Aerosols

A hollow pellet is proposed as fuel for high performance fuel rods. It is difficult, during pellet fabrication, to accurately center the hole in the pellet. A fabrication tolerance must be allowed for the center of the hole and that of the pellet.

The present paper seeks to analytically obtain the temperature and heat flux distribution in a fuel rod containing pellets with eccentric holes. The temperature distribution in power reactor fuel, based on practical assumption, can be obtained within reasonable limits by solving the steady-state heat conduction equation for the pellets with eccentric holes.

The solution is simple, and appropriate to fuel design or safety evaluation.     

Development of Impregnated Agglomerate Pelletization (IAP) process for fabrication of (Th,U)O2 mixed oxide pellets

Impregnated Agglomerate Pelletization (IAP) technique has been developed at Advanced Fuel Fabrication Facility (AFFF), BARC, Tarapur, for manufacturing (Th,²³³U)O₂ mixed oxide fuel pellets, which are remotely fabricated in hot cell or shielded glove box facilities to reduce man-rem problem associated with ²³²U daughter radionuclides. This technique is being investigated to fabricate the fuel for Indian Advanced Heavy Water Reactor (AHWR). In the IAP process, ThO₂ is converted to free flowing spheroids by powder extrusion route in an unshielded facility which are then coated with uranyl nitrate solution in a shielded facility. The dried coated agglomerate is finally compacted and then sintered in oxidizing/reducing atmosphere to obtain high density (Th,U)O₂ pellets. In this study, fabrication of (Th,U)O₂ mixed oxide pellets containing 3–5 wt.% UO₂ was carried out by IAP process. The pellets obtained were characterized using optical microscopy, XRD and alpha autoradiography. The results obtained were compared with the results for the pellets fabricated by other routes such as Coated Agglomerate Pelletization (CAP) and Powder Oxide Pelletization (POP) route.     

Intra-pellet neutron flux distribution measurements in LWR critical lattices

We have developed inexpensive and easy-handling measurement methods on intra-pellet neutron flux. A foil activation method with metallic foils, which were fabricated by punching out technique and etching technique to reduce fabrication error and positioning error, was used for the intra-pellet neutron flux distribution measurement. The developed method was applied to measure intra-pellet neutron flux distributions in a reduced–moderation light water reactor (LWR) lattices, and uncertainty of the distributions was estimated to be 1% to 2%. Measured values were analyzed with a continuous energy Monte Carlo code. Comparison of measurements and analyses revealed that the developed method is useful for the validation of an advanced fuel design method considering neutron behavior in fuel pellets.     

Swelling under irradiation of MgO pellets containing americium oxide: The ECRIX-H irradiation experiment

The ECRIX-H irradiation experiment studied the behaviour of pellets containing americium dispersed in MgO. The purpose of the irradiation was to demonstrate the capacity of magnesia to provide an efficient support matrix. After fabrication, the sintered pellets contained 16.65 wt.% of Am microdispersed in the inert matrix. The ECRIX-H pellets were irradiated under a locally moderated neutron flux in the Phénix sodium-cooled fast reactor (SFR) for 318 Effective Full Power Days (EFPD). Post-test calculations indicated that the fission and transmutation rates of americium at the maximum flux plane reached 33.9% and 92.6% respectively at the end of the irradiation phase. The results of the post-irradiation examinations - both non-destructive and destructive - are discussed in this paper. These results indicate a satisfactory behaviour of the MgO matrix. Particularly, a moderate swelling occurs in the pellets under irradiation even with significant quantities of helium generated and at high transmutation rate.     

Applicability of CeO2 as a surrogate for PuO2 in a MOX fuel development

The applicability of cerium oxide, as a surrogate for plutonium oxide, was evaluated for the fabrication process of a MOX (mixed oxide) fuel pellet. Sintering behavior, pore former effect and thermal properties of the Ce–MOX were compared with those of Pu–MOX. Compacting parameters of the Pu–MOX powder were optimized by a simulation using Ce–MOX powder. Sintering behavior of Ce–MOX was very similar to that of Pu–MOX, in particular for the oxidative sintering process. The sintered density of both pellets was decreased with the same slope with an increasing DA (dicarbon amide) content. Both the Ce–MOX and Pu–MOX pellets which were fabricated by an admixing of 0.05 wt% DA and sintering in a CO₂ atmosphere had the same average grain size of 11 μm and a density of 95%T.D. The thermal conductivity of the Pu–MOX was a little higher than that of the Ce–MOX at a lower temperature but both conductivities became closer to each other above 900 K. Cerium oxide was found to be a useful surrogate to simulate the Pu behavior in the MOX fuel fabrication.     

Fabrication of Mixed Urianium–Plutonium Fuel on the Paket Setup and Waste Handling Methods

The results of fabrication of fuel elements with mixed uranium–plutonium oxide fuel are presented. The experimental fuel assemblies assembled from the fuel elements were tested in BN-350 and -600 reactors. Postreactor investigations of the fuel elements showed that there was no substantial difference in the behavior of the fuel cores consisting of the mixed fuel as compared with UO₂ fuel. Solid and liquid radioactive wastes are produced during the fuel fabrication process. A classification of the wastes and methods for handling them is given. It is shown that the off-grade sintered pellets should be pulverized and returned to the beginning of the mixed-fuel fabrication process.     

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.     

The effect of processing on the thermal diffusivity of MgO-Nd2Zr2O7 composites for inert matrix materials

Oxides possess many of the required properties suitable for an inert matrix fuel in light water reactors, however, their primary disadvantage is low thermal conductivity. Composites are being investigated to maximize the thermal conductivity of the inert matrix fuel by using thermally conductive MgO as the primary phase while improving its hot water corrosion resistance through the addition of a second phase acting as a hydration barrier. Inert matrix fuel candidate MgO-Nd₂Zr₂O₇ composites were synthesized with multiple processing methods, the composite powders were characterized, the resulting microstructures quantitatively analyzed, and the thermal diffusivity of the composites was measured. Among the four processing methods investigated, ball milling and high-energy shaker blending produced the most homogeneous microstructures with a negligible amount of MgO and Nd₂Zr₂O₇ heterogeneities. An effect of processing on the properties of the composites manifests as a larger variation in the thermal diffusivity in pellets processed by methods that produce a higher quantity and frequency of MgO and Nd₂Zr₂O₇ heterogeneities than in methods that produce negligible amounts of heterogeneities.     

Fuel elements based on spherical fuel pellets with a protective coating for enhanced-safety reactors

The principal investigations performed at the Scientific and Industrial Association Luch on the development of fuel elements based on spherical pellets of nuclear fuel with protective ceramic and metallic coatings for HTGR. VVéR, and other types of reactors are reviewed. The main solutions concerning the construction and technical and materials-science aspects of the fabrication of fuel micropellets, fuel microelements, and fuel elements of different modifications (spherical, rod-shaped) are examined. The characteristics of fuel elements and their components at the fabrication and preliminary and reactor testing stages are presented. It is shown that because of additional protective barriers the fuel elements which have been developed effectively confine the fission products and ensure safety. The directions of possible practical utilization of the fuel elements developed for improved enhanced-safety reactors are described. 7 figures, 6 tables, 42 references. Scientific and Industrial Association Luch, Translated from Atomnaya énergiya, Vol. 87, No. 6, pp. 451–462, December, 1999     

UO2-不锈钢燃料棒的光纤激光切割研究

为验证光纤激光用于燃料组件解体和燃料棒切割的可行性,研究光纤激光用于热物性差别很大的UO2芯块 不锈钢包壳管复合结构的切割和铀芯块的切割质量,本文采用光纤激光切割UO2芯块 316Ti包壳管元件棒,并通过扫描电子显微镜、能谱和X射线衍射对UO2芯块的切断面进行微观表征分析,研究激光切割过程对铀芯块切断的表面微观形貌、元素组成及物相的影响。研究结果表明,光纤激光可用于切割UO2芯块 316Ti包壳管元件棒,激光切割过程虽会造成铀芯块切断面出现大量微孔和碎渣,但不会造成UO2的相变。以上结果表明,光纤激光可用于UO2芯块 316Ti包壳管元件棒的切割,通过后续对激光切割系统的抗辐射屏蔽防护,可应用于乏燃料组件解体和乏燃料棒切割。     

AC-3-irradiation test of sphere-pac and pellet (U,Pu)C fuel in the US Fast Flux Test Facility

The objective of the AC-3 bundle experiment in the Fast Flux Test Facility (FFTF) was to evaluate a fuel fabrication method by ‘direct conversion’ of nitrate solutions into spherical uranium–plutonium carbide particles and to compare the irradiation performance of ‘sphere-pac’ fuel pins prepared at Paul Scherrer Institute (PSI) with standard pellet fuel pins fabricated at Los Alamos National Laboratory (LANL). The irradiation and post test examination results show that mixed carbide pellet fuel produced by powder methods and sphere-pac particle fuel developed by internal gelation techniques are both valuable advanced fuel candidates for liquid metal reactors. The PSI fabrication process with direct conversion of actinide nitrate solutions into various sizes of fuel spheres by internal gelation and direct filling of spheres into cladding tubes is seen as more easily transferable to remote operation, showing a significant reduction of process steps. The process is also adaptable for the fabrication of carbonitrides and nitrides (still based on a uranium matrix), as well as for actinides diluted in a (uranium-free) yttrium stabilized zirconium oxide matrix. The AC-3 fuel bundle was irradiated in the Fast Flux Test Facility (FFTF) during the years 1986–1988 for 630 full power days to a peak burn up of 8 at.% fissile material. All of the pins, irradiated at linear powers of up to 84 kW/m, with cladding outer temperatures of 465 °C appeared to be in good condition when removed from the assembly. The rebirth of interest for fast reactor systems motivated the earlier teams to report about the excellent, still perfectly relevant results reached; this paper focusing on the sphere-pac fuel behaviour.     

亚化学计量UO2-x燃料芯块的制备机理

亚化学计量UO_2-x芯块是一种设计新颖的特殊核反应堆用核燃料,很难采用传统压水堆超化学计量UO_2+x+U芯块工艺进行制造。本工作采用UO_2+x+U混合粉末为原料制备了UO_2-x芯块,研究了铀粉表面包覆处理方法、铀粉含量、成型压力、烧结气氛等工艺参数对芯块O/U比、烧结密度和微观结构的影响,探讨了UO_2-x环形芯块的亚化学计量形成机理。研究表明,当铀粉加入量（质量分数）分别为0、3%、6%时,芯块O/U比分别为2.010、1.991、1.982,平均晶粒尺寸分别为10、15、20μm;当铀粉加入量为50%时,O/U比为1.943,样品发生熔化。亚化学计量UO_2-x芯块必须在干燥惰性气氛中密封保存。     

UO2复合燃料芯块导热性能有限元模拟

目前各国均在开发适用于压水堆的含有高导热性第二相材料的新型先进UO₂复合燃料芯块。本文通过有限元计算方法分析了新型先进UO₂复合燃料芯块关键结构参数对其导热性能的影响。结果表明：少量高导热性第二相材料的添加可显著降低燃料芯块服役过程中的中心线温度;第二相的种类、含量、分布形式等均对新型先进UO₂复合燃料芯块的导热性能有重要影响。     

Online integrated visual inspection and sorting system for fast reactor fuels

Mixed oxide (MOX) fuel for prototype fast breeder reactor (PFBR) is designed to have initial burn up of 100,000 MWD/T. The major differences from thermal reactor fuel are relatively smaller dimension with central hole and higher plutonium concentration (21% and 28% of PuO₂) MOX pellets which are loaded into 2.5 m long clad tubes with depleted UO₂ blanket pellets at either end of the MOX stack. The relatively smaller dimension of fuel pellets for PFBR results in large volume at fabrication and inspection. To ensure fast and accurate inspection and sorting of as sintered pellets with less radiation exposure to personnel an integrated on line pellet inspection system for remote visual inspection and sorting of pellets based on diameter has been developed. Details of the integrated pellet inspection system developed at Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur along with the results of the performance trials has been described in this paper.     

New concept of designing Pu and MA containing fuel for fast reactors

New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO₂ powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO₂ powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.     

Fuel performance evaluation of rock-like oxide fuels

The concept of the rock-like oxide (ROX) fuel has been developed for the annihilation of excess plutonium in light water reactors. Irradiation tests and post-irradiation examinations were carried out on candidate ROX fuels. The ternary fuel of YSZ–spinel–corundum system, the single-phase fuels of YSZ, the particle-dispersed fuels of YSZ in spinel or corundum matrix, and the blended fuels of YSZ and spinel or corundum matrix were fabricated and submitted to irradiation testings. The fuels containing spinel showed chemical instabilities with the vaporization of MgO component, which caused fuel restructuring. The swelling behavior was improved with the particle-dispersed fuels. However, the particle-dispersed fuels showed higher fractional gas release (FGR) than blended type fuels. The FGR of YSZ single-phase fuels were comparable to what would be expected for UO₂ fuel at the similar fuel temperatures. The YSZ single-phase fuel showed the best irradiation performance among the ROX fuels investigated.