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.     

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.     

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.     

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens’ theory and reported thermal conductivities of unirradiated (U, Pu) O₂ and irradiated UO₂ pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.     

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.     

On the Cs,Te fission product-induced attack and embrittlement of stainless steel cladding in oxide fuel pins

Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), attempts are made to rationalize the observed out-of-pile Cs: Te dependences of FCCI and FPLME incidence and severity, and their particular Cs, Te synergisms, in terms of Cs-Te thermochemistry and phase equilibria. Successful rationalization in the case of FPLME is taken to point up the critical importance of Te activity and Cs-Te physical state in the FPLME mechanism. A similar conclusion is reached for CCCT, the nonoxidative mode of FCCI, however oxidative modes of FCCI are concluded to rely more on the physical or catalytic properties of Cs-Te mixtures than on specific thermodynamic properties such as Te or Cs activities. The possibility of synergistic coupling between oxidative FCCI and FPLME in irradiated fuel pins is also examined, and it is concluded that although the available evidence does not support such coupling under monotonie loading, it is suggested — as intergranular notch-sensitivity in FPLME — under cyclic loading conditions.     

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.     

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.     

Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up

In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ∼30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap.     

Laser cladding of Colmonoy 6 powder on AISI316L austenitic stainless steel

Stainless steels are widely used in nuclear power plant due to their good corrosion resistance, but their wear resistance is relatively low. Therefore, it is very important to improve this property by surface treatment. This paper investigates cladding Colmonoy 6 powder on AISI316L austenitic stainless steel by CO₂ laser. It is found that preheating is necessary for preventing cracking in the laser cladding procedure and 450 °C is the proper preheating temperature. The effects of laser power, traveling speed, defocusing distance, powder feed rate on the bead height, bead width, penetration depth and dilution are investigated. The friction and wear test results show that the friction coefficient of specimens with laser cladding is lower than that of specimens without laser cladding, and the wear resistance of specimens has been increased 53 times after laser cladding, which reveals that laser cladding layer plays roles on wear resistance. The microstructures of laser cladding layer are composed of Ni-rich austenitic, boride and carbide.     

Influence of silicon on swelling and microstructure in Russian austenitic stainless steel EI-847 irradiated to high neutron doses

Void swelling and microstructural development of niobium-stabilized EI-847 austenitic stainless steel with a range of silicon levels were investigated by destructive examination of fuel pin cladding irradiated in three fast reactors located in either Russia or Kazakhstan. The tendency of void swelling to be progressively reduced by increasing silicon concentration appears to be a very general phenomenon in this steel, whether observed in simple, single-variable experiments on well-defined materials or when observed in multivariable, time-dependent irradiations conducted on commercially produced steels over a wide range of irradiation temperatures, neutron spectra and dpa rates. The role of silicon on microstructural development is expressed both in the solid solution via its influence on dislocation and void microstructure and via its influence on formation of radiation-induced phases that in turn alter the matrix composition. Surprisingly, increases in silicon level in this study do not accelerate the formation of silicon-rich G-phase, but act to increase the formation of Nb (C,N) precipitates. Such precipitates are known to be associated with delayed void swelling.     

A non destructive technique for determination of U and Pu content in MOX fuel pins for fast reactors

MOX fuel pins containing both U²³³O₂ and PuO₂ have been fabricated for making an experimental subassembly for irradiation in Fast Breeder Test reactor (FBTR) at Kalpakkam, India. This unique composition of the fuel pin is chosen to simulate the thermo-mechanical conditions of the upcoming Prototype Fast Breeder Reactor (PFBR) in the existing Fast Breeder Test Reactor. Since the fertile matrix is natural UO_2, it was difficult to monitor the percentage of U²³³O₂ through chemical methods and neutron assay methods. During the fabrication of MOX fuel pins at Advanced Fuel Fabrication Facility; Bhabha Atomic Research Centre, Tarapur, Passive Gamma Scanning (PGS) was employed as one of the characterisation tools for verifying the fuel composition. PGS was found to be effective in estimating the percentage composition of both U²³³O₂ and PuO₂ and also in ensuring the uniform distribution of the fissile material in MOX fuel pins. PGS is also found effective in monitoring the correct loading of natural UO₂ insulation pellets and MOX fuel pellets in welded MOX pins.     

Correlation of yield stress and microhardness in 08Cr16Ni11Mo3 stainless steel irradiated to high dose in the BN-350 fast reactor

The relationship between the microhardness and the engineering yield stress in 08Cr16Ni11Mo3 steel after irradiation in the BN-350 reactor has been experimentally derived and agrees with a previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K_Δ format where the correlation involves changes in the two properties, excellent agreement is found with a universal K_Δ correlation developed by Busby and coworkers. Additionally, this report points out that microhardness measurements must take into account that sodium exposure at high temperature and neutron fluence alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing.     

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U–Zr alloy fuel elements irradiated in the Experimental Breeder Reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining.     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.     

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of 25–93 MWd kg⁻¹ UO₂, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.     

Modelling the high temperature mechanical behaviour of helium embrittled austenitic stainless steel

Current theories for the growth of grain-boundary defects during plastic deformation have been used to model the ductility of an irradiated austenitic steel as a function of strain rate and thermal-neutron fluence. Failure of irradiated steels is via the growth and linkage of helium bubbles on the grain boundaries produced by an (n, ) reaction between the thermal neutrons and boron present in the steel as a controlled trace element.

Reasonable agreement is obtained between theoretically predicted and experimentally measured ductilities over a range of strain rates between 10⁻¹⁰ and 10⁻²/s. However, in order to obtain such agreement over the range of strain rates examined, variables such as defect size and spacing had to be carefully selected. The values used were internally self-consistent with the amount of helium known to be availablc to nucleate grain-boundary bubbles.

The models of defect growth have also been used to make some predictions on the creep-fatigue performance of such materials.     

The microstructure of neutron irradiated 20Cr/25Ni/TiN austenitic stainless steel

A nitrided, titanium-stabilised, 20Cr/25Ni austenitic stainless steel was examined in the electron microscope, after 1000 h irradiation at 783 K to doses of 5.0× 10²⁴ n/m² (thermal) and 2.5× 10²⁴ n/m² (fast), in the PLUTO reactor. Microstructures were compared with those of as-received and thermal control samples. The austenitic matrix and M₂₃C₆ particles were free of irradiation-induced damage, while the TiN particles contained loops 2 nm in diameter which coarsened into a network on annealing at 1083 K. Annealing also resulted in a low density of transmutation-induced helium bubbles, ~ 4 nm in diameter, located in precipitate-free regions of grain boundaries. We conclude that 20/25/TiN is relatively unaffected by irradiation at these dose levels and that helium bubble embrittlement is unlikely under normal stresses.