Plutonium redistribution in fast reactor mixed oxide fuel pins

Some redistribution effects of uranium and plutonium, caused by thermal diffusion and evaporation-condensation processes in mixed oxide fuels, are discussed by means of autoradiographs of sections of fuel pins irradiated in the fast flux of the RAPSODIE reactor. The change in the stoichiometric state as a function of burnup and the radial distribution of oxygen are described and their influence on the redistribution processes is discussed. A model and suitable data are given to calculate redistribution effects on the basis of thermal diffusion in fast reactor fuels. In fuel pins with power ratings of 500 W/cm and 600 W/cm the enrichment of plutonium around the central cavity produces an increase in the central temperature of about 100°C and 250° C, respectively.     

Cesium-uranium-oxygen chemistry in uranium-plutonium oxide fast reactor fuel pins

The Cs-U-O system has been reinvestigated in light of recently reported thermodynamic data for Cs₂U₄O₁₂ and recent phase data showing the existence of a new Cs-U-O compound with an $\text{O}\text{(U + Cs)}$ atom ratio less than that of Cs₂UO₄. Our experiments have confirmed the existence of a new phase, allowed the formula Cs₂UO_3.56 to be assigned, and generated thermodynamic data for this new compound. This new phase exists only at oxygen potentials that are too negative to be encountered in uranium-plutonium oxide fast reactor fuel pins. The compound Cs₂UO₄ appears to be the most likely one to be formed, with the formation occurring at the fuel-blanket interface.     

Rupture and fragmentation of pressurized pipes and fast reactor fuel pins

A model of axial crack propagation in a pressurized tube is developed which predicts the crack velocity and deformation geometry and the minimum driving pressure. Emphasis is placed upon the stability of propagation. The model also offers a criterion for the appearance of multiple cracks and subsequent fragmentation of the tube wall due to excessive axial bending strains. The model is applied to the rupture of gas pipelines, PWR coolant pipes and fast reactor fuel pins.     

Power-to-melts of uranium-plutonium oxide fuel pins at a beginning-of-life condition in the experimental fast reactor JOYO

Power-to-melts of uranium-plutonium oxide fuel pins at an initial startup condition were experimentally obtained from the B5D-2 test in the experimental fast reactor JOYO in Oarai Engineering Center. MCNP code calculations were combined with burnup measurements to determine linear heat rating of the test fuel pins. To identify the axial incipient melting positions corresponding to the power-to-melts, solidified grain morphology and molten fuel axial movements were characterized. Extensive observations on longitudinal ceramographs allowed classifying molten fuel settlements near bottom and top extents of axial fuel melting into three types. The power-to-melts depended slightly on fuel-to-cladding gap sizes and clearly on both oxygen-to-metal ratios and densities of fuel pellets. These dependencies resulted from the fuel pellet cracking and relocation behavior, which fairly improves heat transfers across the gaps. Also, the power-to-melt at the bottom position was higher than that at the top position due to an axial gradient of cladding temperatures in each fuel pin.     

Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 10²⁶ n/m² as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

ODS cladding fuel pins irradiation tests using the BOR-60 reactor

In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burn-up of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed and indicated good application prospects for high burn-up fuel. On the other hand, anomalous irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the anomalous irradiation behaviors.     

Fuel-to-cladding gap evolution and its impact on thermal performance of high burnup fast reactor type uranium-plutonium oxide fuel pins

Drastic evolution of fuel-to-cladding gap is observed in high burnup JOYO Mk-II driver and MONJU type uranium-plutonium oxide fuel pins. The effect of the evolution is examined from viewpoints of fuel restructuring, gaseous FP release and retention and cesium migration behaviors. Its thermal impact on fuel pin performance is also studied by one-dimensional steady state thermal analysis. Threshold condition of the evolution depends on fuel pellet characteristics, burnup and probably temperature. The evolution directly relates to as-fabricated microstructures and to gaseous FP release and retention behavior. A comparison of fuel restructuring with predicted temperature profiles indicates that, even where large residual gaps are observed, non-gaseous filler always improves the heat transfer across the gaps.     

Oxygen redistribution in fast reactor oxide fuel

Out-of-pile tests were carried out in order to investigate the oxygen redistribution in uranium-plutonium mixed oxides exposed to a thermal gradient. In hypostoichiometric oxide fuel the oxygen migrates towards the low temperature region of the pellet and in hyperstoichiometric fuel the oxygen migrates in the opposite direction. The oxygen transport is explained on the basis of solid-state thermal diffusion and occurs via vacancies and interstitials. It has been shown that the heats of oxygen transport are a function of plutonium and uranium valencies for hypo- and hyperstoichiometric oxides, respectively. The experimental results allowed to construct a practical example in which oxygen profiles in fuel pins were calculated as a function of initial stoichiometry and burnup.     

Influence of pellet length on interaction between fuel and cladding in simulated reactor fuel pins

An electrically heated fuel pin test apparatus has been developed for out-of-pile investigations of fuel pin parameters with a view to supplementing in-pile experiments. Sixty per cent of reactor heat ratings has been achieved with a hollow pin having an axially located electrical heater, the limitation being the melting of the UO₂ pellets. The theoretical unconstrained shapes of a heated pellet and a fuel pellet under elastic conditions were calculated. Both showed an ‘hour glass’ form suggesting that permanent circumferential ridges would occur in the cladding of a heated pin as they do in the cladding of fuel pins. These ridges were subsequently produced in heated pins, the pins being heated while immersed in cooling water at typical reactor temperatures and pressures. From a series of such tests using different pellet lengths it was found that a significant reduction in ridge height occured when the pellet ratio was one-third of the value in a typical reactor. The temperatures reached in the UO₂ pellets were estimated from a metallographic examination of a pin cross section after test. Using published data of ∫kdT for UO₂ over various temperature ranges the pin heat output at that cross section was determined.     

Molten fuel dispersal in hypothetical fast reactor accidents

Some aspects of molten fuel dispersal in hypothetical fast reactor accidents are considered, ranging from the two-phase flow fluid equation forms appropriate to modelling molten fuel dispersal to an analytical self-similar solution as a function of space and time for the dispersal of a two-phase molten fuel/fission gas mixture by fission gas pressures. The analytical solution provides both scaling laws for fuel dispersal velocities as a function of gas content and a solution with which to check code results. A discussion of the COMCYL program for molten fuel dispersal in a sodium volded channel is used to illustrate the types of problems that need to be tackled in a molten fuel dispersal program and typical results obtained from the application of COMCYL to hypothetical loss-of-coolant flow accidents are presented.     

Fuel-cladding mechanical interaction in fast breeder fuel pins; observations and analysis

The problems of fuel-cladding mechanical interaction are considered, and a survey is given of the causal processes in oxide fuel pins. Their importance is judged by relevant results from irradiation experiments in thermal and fast test reactors, and by corresponding modelling computations. It is demonstrated that critical cladding strain has to be expected only in case of large, fast rod power increase, primarily from reduced power to full power, and maybe also in case of serious cesium accumulation at the fuel/blanket interface. Steady-state fuel swelling does not make any considerable contribution to cladding strain by thermal creep or plastic flow.     

基于CSA算法的快堆堆芯燃料管理研究

CSA(Character Statistic Algorithm)算法是由清华大学核研院研究开发的特征统计算法,目前已可用于压水堆的堆芯燃料管理。采用CSA优化算法结合快堆堆芯计算程序HNDB,对快堆的平衡循环换料进行优化,计算结果说明CSA算法可以很好地用于快堆的平衡循环换料,可为快堆堆芯燃料管理程序的开发提供借鉴。     

A CFD simulation process for fast reactor fuel assemblies

A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-? and SST (Menter) k-ω were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.     

A perspective on fast reactor fuel cycle in India

The growing energy needs of India can be fulfilled only by judicious mix of all the fuel resources. It is possible to achieve energy security and sustainability through the introduction of fast reactors in an expeditious manner and closing the fuel cycle. This approach is inevitable in view of the limited uranium resources in India. The Fast Breeder Test Reactor (FBTR) built by India uses mixed carbide as fuel and the 500 MW(e) Fast Breeder Reactor Project (PFBR), to be operational in 2010, will use mixed oxide as fuel. It has also been decided that fast reactors beyond 2020, with enhanced safety features and having better economy, will use metallic fuel. Having successfully operated FBTR with carbide fuels, we need to develop the fuel cycles for both the mixed oxide fuel in the near future and the metallic fuel expeditiously. The progress achieved so far and the plans for implementation are discussed in this paper.     

Physicochemical interaction of oxide fuel with the cladding of the fuel elements of A fast reactor

Conclusions The investigations of fuel elements with mixed oxide fuel, used in BOR-60, revealed four types of corrosion damage to OKh16N15M3B steel cladding due to the action of fission products. It was shown that the general corrosion develops as a result of the interaction of the cladding with cesium with an oxygen potential created by the mixed oxide fuel. Precipitation of carbides, formed as a result of radiation-thermal aging of the steel, on grain boundaries leads to intercrystallite corrosion of the cladding in the presence of cesium. When mixed oxide fuel with the starting ratio O/M=1.98–2.00 is used in the fuel elements, iodide transport of the components of the steel, giving rise to intercrystallite corrosion of the cladding, occurs. It was demonstrated that chemical activity relative to the stainless steel, leading to corrosion damage to the cladding, is high.Translated from Atomnaya Énergiya, Vol. 56, No. 4, pp. 195–199, April, 1984.     

On a fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel

A fast reactor cycle scheme that incorporates a thoria-based minor actinide-containing cermet fuel is given. The present cermet fuel consists of an oxide solid solution of Th and minor actinides and Mo-inert matrix. It has been proposed as a high-performance device that can enhance minor actinide incineration in a fast reactor cycle. It is used in an independent small sub-cycle, whereby dedicated cycle technologies are adopted. Two-step reprocessing process was proposed for the present cermet fuel; it consists of a pre-removal of Mo-inert matrix and an actinide recovery. A preliminary test for the pre-removal of Mo-inert matrix was carried out using a surrogate cermet fuel. Burnup characteristics of a fast reactor core loaded with the cermet fuel were investigated by using neutronic calculation codes. It was revealed that a heterogeneous composition of Mo-inert matrix in inner and outer cores may lead to an effective transmutation of minor actinides and a flattened power density. It was concluded that the present cermet fuel was potentially promising as a high-performance incineration device of minor actinides for fast reactors.     

深燃耗球床式气冷快堆研究

气冷快堆是未来发展的第四代先进核能系统候选堆型之一,它可以满足核能的可持续性、安全可靠性和经济性要求.从反应堆物理和热工水力学的角度出发,设计了热功率300 MW的球床式气冷快堆,选择了碳化物燃料作为气冷快堆的燃料.用耦合燃耗计算程序COUPLE2.0模拟得到了深燃耗气冷快堆的铀燃料循环的平衡态.平衡态研究结果表明基于深燃耗的300 MW球床式气冷快堆可以提高铀资源的利用率同时降低乏燃料中的次锕系核素的含量.当燃料球直径为6 cm,燃料区的直径为5.5 cm,燃料占燃料区的体积的70%,燃料形式为UC,其中235U的初始富集度为12%时,燃料球通过堆芯的时间可以达到12 600 d,重金属燃耗深度为164.38 GWd/t,总的铀资源的利用率可以达到为28.03%.