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.     

高燃耗下快堆燃料与包壳的化学相互作用模型建立及验证

快中子反应堆二氧化铀燃料元件在高燃耗、高中子注量率、高线功率和高温状况下运行,燃料与包壳材料会发生复杂的物理化学相互作用。燃料元件化学相互作用模型的建立对高燃耗快堆燃料元件的设计非常重要。针对快中子反应堆氧化物燃料元件与包壳材料发生的化学相互作用,采用动力学模型建立了二氧化铀与奥氏体不锈钢、铁素体-马氏体钢包壳材料的化学相互作用模型,并通过实验数据验证该模型。结果表明：建立的快堆二氧化铀燃料与奥氏体不锈钢的腐蚀模型可以成功预测最大燃耗10.8at%、辐照损伤87.5 dpa的包壳腐蚀;建立的快堆二氧化铀燃料与铁马钢的腐蚀模型可以成功预测最大燃耗9.3at%、辐照损伤76.6 dpa的包壳腐蚀。研究结果为高燃耗二氧化铀辐照元件及示范快堆燃料元件的设计和性能预测提供重要的参考价值。     

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.     

Attack upon the cladding of oxide fuel pins by fuel and fission products

In parallel with post-irradiation examinations, a comprehensive out-of-pile experimental programme has been performed to determine the most important fission product reactions with four austenitic stainless steels at different oxygen potentials. Single as well as groups of fission products (simulated burn-up systems) have been used. Only the elements cesium, iodine and tellurium cause dangerous reactions with the cladding of an oxide fuel pin. The others are either not reactive or produced in such small quantities that their attack on the cladding is insignificant. Molybdenum is often found in the reaction zone of an irradiated oxide pin. However, according to our out-of-pile results it does not look as if molybdenum is a dangerous fission product. A decisive factor for the occurrence of reactions with the cladding is the oxygen potential in the fuel pin. As long as the O/M ratio of the fuel is markedly below 2.00, there are no dangerous reactions, neither with cesium nor with tellurium and iodine. The post-irradiation investigations (burn-up 1 to 10 at %) have shown that the cladding attack below 750 °C is most dependent on the inner wall temperature. Other factors, including fuel density, rod power and burn-up, seem to play a minor role. A noticeable reduction of the cladding attack was observed when the initial O/M ratio of the fuel was less than 1.98. A kinetic evaluation of some of the reactions observed in the out-of-pile tests has been attempted. At temperatures above 700 °C, the influence of temperature decreases markedly and the fission product concentration in the fuel becomes more important. There are indications that this also holds true for in-pile conditions.     

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.     

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.     

船用堆燃料棒包壳疲劳寿命分析

船用堆瞬态变工况下燃料棒包壳温度和冷却剂压力波动较大,引起包壳的疲劳损伤,因此包壳疲劳寿命分析至关重要。本文利用ANSYS软件模拟船用堆瞬态变工况下燃料棒的热机械行为,结合锆包壳疲劳寿命设计曲线,考察包壳温度、冷却剂压力、燃料棒内压以及辐照对船用堆燃料棒包壳疲劳寿命的影响。计算结果表明,瞬态变工况使得包壳疲劳寿命有很大降低;包壳温度变化与冷却剂压力变化相比,前者对包壳疲劳寿命的影响小;辐照会降低包壳疲劳寿命。在不影响核动力船舶机动性的前提下,可采取一些必要的措施来降低包壳的疲劳损伤。     

Pellet-cladding interactions and length changes in SGHWR fuel pins

Fuel pins in the Steam Generating Heavy Water Reactor increase in length during irradiation, due largely to a mechanical interaction between the cladding and the fuel pellets. The length change produced by this process depends critically on the deformation behaviour of the fuel pellet under the interaction stresses. This paper presents a model of the pellet deformation which has been incorporated into a computer programme, SLIM, capable of predicting extensions from this and other causes in SGHWR fuel pins subjected to any power history. SLIM is shown to give good agreement with measured length changes for all variants of fuel rating, burn-up and pin type tested, and predicts that extension of the narrower pin designs now in use should be reduced by a factor of 0.7 compared with similarly rated standard pins.     

On the failure of fast reactor fuel pins

The prediction of the timing and position of fuel pin failures is an important task in the modelling of fast reactor fuel behaviour. The range of processes that can provoke failure of fast reactor fuel pins in normal operating conditions and during hypothetical accidents is reviewed. Some of the mechanisms of failure are examined in more detail and the effect of hot spots and local stress concentrations is discussed. A review of failure criteria used in fast reactor fuel pin codes is given elsewhere, but the difficulties in applying various types of criteria are examined. Some discussion is also given on probabilistic approaches. Recommendations are given for a future approach to the problem of failure prediction, resolving the dilemma between inadequate empirical criteria and over-complex physically based approaches.     

Anisotropic growth of reactor fuel pin cladding

Anisotropic growth of 316 stainless steel reactor fuel pin cladding was found to occur after irradiation in the Experimental Breeder Reactor-II (EBR-II). Pressurized tube specimens were irradiated to a peak fluence of 10²³n/cm² (E >0.1 MeV) at temperature ranging from 430°C to approximately 590°C. Growth was observed in both the annealed and 20% cold worked conditions and was found to decrease with increasing hoop stress. The anisotropic growth is more pronounced in the cold worked condition. The growth is attributed to a preferred orientation of Burgers vectors in the preirradiated cold worked dislocation structure.     

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.     

A comparison between reactor experiments and SLEUTH-SEER code predictions of pellet-clad interactions in AGR fuel pins

Power cycling endurance experiments have been conducted in order to establish the relative endurance capabilities of commercial AGR type fuel pins and to determine the significance of the prime operating parameters. The information accumulated relates to 20/25/Nb and nitrided 20/25/Ti stainless steel clad fuel pins with hollow fuel pellets, operating at coolant pressures of 400 and 600 psi and clad temperatures in the range 680–840°C. The data has provided a basis for the validation of the SLEUTH-SEER model for this fuel type, and in the analysis the experimental observations are compared with model predictions in order to give an impression of the reliability of the code in commercial AGR applications. It is concluded that in the clad temperature range 700–800°C the SLEUTH-SEER code predicts the power cycling endurance of 20/25/Nb stainless steel clad fuel pins to within a factor 2.     

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.     

Interactions between fuel pins and assembly components

The structure of a fuel element has the primary function of holding fuel pins in a regular array throughout their irradiation life, and secondary functions of permitting loading and unloading operations and transportation to be carried out without damage. Differential thermal expansions, irradiation-induced dimensional changes and flow-induced phenomena have to be accommodated. These factors can lead to fuel pin damage and ultimate loss of cladding integrity, or to distortions which may affect the thermal performance either in normal operation or in a loss-of-coolant accident. This paper discusses the various types of interaction that have been experienced and their consequences, as well as the design principles that should be followed to avoid them.     

Influence of iodine on the burst strain of Zircaloy-4 cladding tubes under simulated reactor accident conditions

The Stress Corrosion Cracking (SCC) behavior of short tubular Zircaloy-4 (Zry) specimens by the action of iodine was investigated out-of-pile between 600 and 1100°C under inert gas conditions. The Zry cladding tubes were used as-received, with an inner preliminary oxidation and pre-damaged, respectively. Moreover, the influence of the UO₂ oxygen potential on the SCC behavior was studied. The burst and creep-rupture tests (time-to-rupture ?15 min) clearly show that the deformation behavior of Zry cladding tubes below 850°C is influenced by the presence of iodine. A low ductility failure of Zry tubes takes place that is characterized by little deformation as compared to reference specimens without iodine. Moreover, in the isothermal, isobaric experiments, the time-to-failure of the iodine containing specimens is markedly shorter as compared to the iodine-free reference specimens. Internal preoxidation of the cladding tube or UO₂ in the specimens exert an additional influence on the mechanical properties of Zry. SEM examinations of the rupture surface of the cladding tubes show that in the presence of iodine and at burst temperatures below 850°C the cracks in the cladding material are mainly intergranular followed by ductile residual rupture.     

A proposed mechanism for internal cladding corrosion in LMFBR mixed oxide fuel pins

A model is proposed which describes a possible chronological sequence of events during inner cladding corrosion in mixed oxide fuel pins. The corrosion reaction is visualized as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the inter-granular form of attack and the distribution of corrosion products in the fuel/cladding gap.     

Gas-cooled fast breeder reactor designs with advanced fuel and cladding

This paper examines the potential impact of some alternative cladding and fuel materials being considered for the liquid metal fast breeder reactor (LMFBR) on the performance and design of large commercial gas-cooled fast breeder reactors (GCFRs). Mixed carbide fuel and Inconel 718 cladding material were examined. Another cladding alternative considered was silicon carbide (SiC), which presents some interesting possibilities in high-temperature performance. Design concepts based on the above fuel and claddings were examined and compared with a reference oxide/316 stainless steel design based on a commercial 4000 MW(th) [1500 MW(e)] system. Substantial benefits can be derived from a high-temperature cladding material such as Inconel 718 or ¹⁶Pe; core volume and steam generator heat transfer area could be reduced by 20% or more, and significant reductions in core inventory and doubling time are possible. Carbide fuels would reduce the number of fuel rods by 50% because of higher linear power, and doubling time would be lowered.     

Oxidation of aluminum alloy cladding for research and test reactor fuel

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.