Fuel Cladding Research Articles

Automatic Detection of Spills and Gas Pockets upon Ultrasonic Checking of Welds in Zirconium Fuel Claddings

The problem of automating detection of flaws (gas pockets and spills) in welds obtained by electron-beam welding, using the ultrasonic echo method which is implemented by slanted introduction of ultrasonic waves into a thin zirconium fuel shell, is considered. Algorithms for automatic detection of flaws and evaluation of their relative dimensions are proposed. Experimental results confirming the efficiency of the algorithms proposed are reported.

Crystallography of Zirconium Hydrides in Recrystallized Zircaloy-2 Fuel Cladding by Electron Backscatter Diffraction

Precipitation morphology and habit planes of the δ-phase Zr hydrides, which were precipitated within the a-phase matrix grains and along the grain boundaries of recrystallized Zircaloy-2 cladding tube, have been examined by electron backscatter diffraction (EBSD). Radially-oriented hydrides, induced by residual tensile stress, precipitated in the outside region of the cladding, and circumferentially-oriented hydrides in the stress-free middle region of the cladding. The most common crystallographic relationship for both types of the hydrides precipitated at the inter- and intra-granular sites was identical at (0001)α // {111}δ, with {10&1macr;7}α // {111}δ, being the occasional exception only for the inter-granular radial hydrides. When tensile stress was loaded, the intra-granular hydrides tended to preferentially precipitate in the grains with circumferential basal pole textures. The inter-granular hydrides tended to preferentially precipitate on the grain faces opposite to tensile axis. The change of prioritization in the precipitation sites for the hydrides due to tensile stress could be explained in terms of the relaxation effect of constrained elastic energy on the terminal solid solubility of hydrogen at hydride precipitation.

The Post-Test Analyses of CCTF C2-4 using TRAC-M and RELAP Codes Code Modeling Capability on the Downcomer Boiling

The void generation in downcomer during the late reflood phase and its effect on overall transient is evaluated using the post test analyses of the CCTF C2-4 (Run 62) experiment. C2-4 test, which was performed at JAERI, is a reflood test for the typical Westinghouse 4 loop plants. TRAC-M and RELAP 3.2.2γ codes are chosen in this study. Results show that both codes overpredict downcomer water level reduction. To examine the sensitivity of the interfacial drag model in the downcomer region, Blasius model option is selected for TRAC-M for the sensitivity study. Also, Blasius model is implemented in RELAP code. The sensitivity runs using Blasius model show that it enhances the steam venting in the downcomer. With the Blasius interfacial drag model, both codes show a good agreement with the experiment for downcomer level. Fuel cladding temperatures predicted by the codes are still higher than the measured data.

Assessment of the Combined Effects of Irradiation and Hydrogenation on the Fracture Behavior of Zircaloy Fuel Claddings by Fracture Mechanics

In order to study the combined effects of irradiation and hydrogenation on the failure behavior of high burnup light water reactor (LWR) fuel cladding rods during reactivity initiated accidents (RIAs), an analysis based on fracture mechanics has been performed at room temperature for the following two types of claddings: irradiated Zircaloy-2 claddings with precipitated platelet planes of zirconium hydride oriented in the radial direction of the claddings rather than circumferential direction (Type-A), and irradiated Zircaloy-4 claddings with zirconium hydride accumulated locally beneath the outer surface of the claddings (Type-B). A failure assessment diagram (FAD) was constructed using the R6 approach first produced by the former Central Electricity Generating Board (CEGB) in the United Kingdom to estimate the failure stress of the Type-A and Type-B models. It was concluded from the FAD that hydrogenation could be the dominant factor in affecting the strength degradation of the Zircaloy claddings rather than irradiation effect, depending on the distribution and orientation states of the hydride precipitates.

A Simple Model for Hydrogen Re-distribution in Zirconium-lined Fuel Claddings.

A simple model of hydrogen diffusion and hydride precipitation was developed to estimate the radial re-distribution of hydrogen in zirconium-lined Zircaloy-2 claddings for light water reactors. Numerical calculation based on the model agreed well with experimental data and clearly showed that a slight difference of the terminal solid solubility of hydrogen in the two materials caused a significant flow of hydrogen from Zircaloy-2 to zirconium during slow cooling. The model would also evaluate the hydrogen distribution in the claddings of which the terminal solid solubility is not uniform.

A Simple Model for Hydrogen Re-distribution in Zirconium-lined Fuel Claddings

A simple model of hydrogen diffusion and hydride precipitation was developed to estimate the radial re-distribution of hydrogen in zirconium-lined Zircaloy-2 claddings for light water reactors. Numerical calculation based on the model agreed well with experimental data and clearly showed that a slight difference of the terminal solid solubility of hydrogen in the two materials caused a significant flow of hydrogen from Zircaloy-2 to zirconium during slow cooling. The model would also evaluate the hydrogen distribution in the claddings of which the terminal solid solubility is not uniform.

Assessment of the Combined Effects of Irradiation and Hydrogenation on the Fracture Behavior of Zircaloy Fuel Claddings by Fracture Mechanics.

Assessment of the Combined Effects of Irradiation and Hydrogenation on the Fracture Behavior of Zircaloy Fuel Claddings by Fracture Mechanics.

Influence of Precipitated Hydride on the Fracture Behavior of Zircaloy Fuel Cladding Tube

In order to clarify the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes, the stress-strain distribution of the cladding was estimated by finite element method (FEM) analysis. The mechanical properties of α-phase of zirconium and zirconium hydride required for the analysis were examined by means of an ultrasonic pulse-echo method and a tensile test. It was found from the analysis that the non-hydrided cladding has the highest equivalent plastic strain at the inner surface of the cladding, suggesting that the fracture initiated at the inner surface of the cladding. Since the hydride accumulated layer located in the outer surface of the hydrided cladding fails at a lower internal pressure, the crack appears to initiate at the outer surface of the cladding. The fracture behavior estimated from the stress states of the hydrided cladding was in good agreement with the experimental results obtained by pulse irradiation tests of the Nuclear Safety Research Reactor (NSRR) and high-pressurization-rate burst tests in the Japan Atomic Energy Research Institute (JAERI).

Influence of Precipitated Hydride on the Fracture Behavior of Zircaloy Fuel Cladding Tube.

Influence of Precipitated Hydride on the Fracture Behavior of Zircaloy Fuel Cladding Tube.

高速炉燃料被覆管用ＯＤＳフェライト‐マルテンサイト鋼の開発

高速炉燃料被覆管用ＯＤＳフェライト‐マルテンサイト鋼の開発

Long Term Creep Behavior of Spent Fuel Cladding for Storage and Disposal

Long Term Creep Behavior of Spent Fuel Cladding for Storage and Disposal

Development of the Ultra-microhardness Technique for Post Irradiation Examination of Fuel Cladding Tubes

Hardness measurements are potentially valuable for a quantitative discussion of embrittlement in the inner portions of fuel cladding tubes. The size of the indentation, however, is not negligible compared to the measuring region, even when a micro Vickers hardness tester is employed. This limits the measuring technique, and very little has been studied about degradation phenomena in the inner portion of the tubes.A hardness measurement system, equipped with a depth-sensing indentation instrument, and the necessary post irradiation examination technique for specimens with high radioactivity were successfully developed and the following observations were obtained from the system's application example. The diffusion coefficient of oxygen obtained from the hardness of an unirradiated zirconium lined cladding with simulated oxidation in the fuel rod showed good agreement with literature data. The calculated diffusion coefficient from hardness in the inner portion of irradiated Zircaloy-2 fuel rods was almost the same value as that of unirradiated zirconium, which implied that neither neutron irradiation nor fission fragment bombardment enhanced the oxygen diffusion in the inner portion of cladding tube.

Development of the Ultra-microhardness Technique for Post Irradiation Examination of Fuel Cladding Tubes.

Hardness measurements are potentially valuable for a quantitative discussion of embrittlement in the inner portions of fuel cladding tubes. The size of the indentation, however, is not negligible compared to the measuring region, even when a micro Vickers hardness tester is employed. This limits the measuring technique, and very little has been studied about degradation phenomena in the inner portion of the tubes. A hardness measurement system, equipped with a depth-sensing indentation instrument, and the necessary post irradiation examination technique for specimens with high radioactivity were successfully developed and the following observations were obtained from the system's application example. The diffusion coefficient of oxygen obtained from the hardness of an unirradiated zirconium lined cladding with simulated oxidation in the fuel rod showed good agreement with literature data. The calculated diffusion coefficient from hardness in the inner portion of irradiated Zircaloy-2 fuel rods was almost ...

Effect of Lithium Hydroxide on Stability of Fuel Cladding Oxide Film in Simulated Pressurized Water Reactor Primary Water Environments

Abstract The trend in pressurized water reactors (PWR) toward higher burnups, increasing lithium concentrations, and higher coolant temperatures imposes a demand for better fuel cladding corrosion ...

Copper Ion Deposition on Fuel Cladding under BWR Conditions

Copper ion deposition behavior on Zircaloy cladding surfaces under simulated BWR conditions was investigated. The results indicated that Cu ion deposition rate on the fuel surface increased and deposited particle size decreased as pH increased. Oxidation of Zircaloy, especially by nodular corrosion, enhanced Cu ion deposition. The Cu deposits were concentrated on the nodular cracks. Copper ions were deposited in the form of CuO. They did not depend on iron crud coexistence.

Copper Ion Deposition on Fuel Cladding under BWR Conditions.

Copper ion deposition behavior on Zircaloy cladding surfaces under simulated BWR conditions was investigated. The results indicated that Cu ion deposition rate on the fuel surface increased and deposited particle size decreased as pH increased. Oxidation of Zircaloy, especially by nodular corrosion, enhanced Cu ion deposition. The Cu deposits were concentrated on the nodular cracks. Copper ions were deposited in the form of CuO. They did not depend on iron crud coexistence.

Extending the Life of Reactor Fuel Cladding Tubes

Extending the Life of Reactor Fuel Cladding Tubes

Microstrueture and Corrosion Resistance of Irradiated Fuel Cladding Tubes

Nodular corrosion on the surface of Zircaloy-2 fuel cladding tubes was investigated and the effects of irradiation time, linear heat rate, and fast neutron flux on the nucleation and growth of nodules were statistically studied. The nucleation and growth rate of nodules were found to decrease with irradiation time. Then, the nodular corrosion resistance of neutron irradiated Zircaloy-2 was examined in an out-of-pile corrosion test, and neutron irradiation was seemed to improve the Zircaloy-2 nodular corrosion resistance. Finally, microstructural evolution of Zircaloy-2 during neutron irradiation was investigated and radiation-induced dissolution of inter-metallic precipitates was confirmed. It was proposed that an increase of the dissolved alloying element concentration should improve the nodular corrosion resistance of Zircaloy-2 and re- strain the nucleation and growth of nodules on Zircaloy-2 fuel cladding tubes during irradiation in BWRs.

高燃焼度（長寿命）高速炉混合酸化物燃料の設計成立性評価

The potential of the Mixed Oxide (MOX) fuel to achieve a high burnup and long life core was studied for four type fuels with the fuel performance code LIFE-IV. Operating conditions, uncertainties and tolerances were varied in calculations and the results were used to establish multi-dimensional response surfaces from the important performance parameters. The Monte Carlo statistical simulation method was applied to determine the effect of statistical tolerance and uncertainty distributions on the fuel performance and to obtain statistical chracteristic of the fuel behavior.(1) From low to medium burnups of 0-100 GWd/t, fission gas swelling dominates Fuel Cladding Mechanical Interaction (FCMI).(2) At high burnup of 150 to 200GWd/t, solid fission product swelling dominates FCMI and fuel performance characteristics are simplified because cladding strain proceeds approximately at a constant rate with burnup.(3) Statistical fuel performance characteristics were evaluated and it was ascertained that MOX has a potential to realize high burnup with 2-σ reliability.(4) The operatioal and fabricational tolerances have little effects on fuel performances.

Intergranular fracture of fast reactor irradiated stainless steel

A mechanistic understanding of the failure of fuel pins in a Fast Reactor, in normal operation or in accident conditions, is very important in assesssing safety. Progress in gaining such an understanding has been slow and greater emphasis has been placed on empirical failure rules. This paper presents a simple set of physical models describing the intergranular fracture of Fast Reactor irradiated stainless steel fuel pin cladding. The approach emphasizes the importance of both the nucleation and growth of intergranular cavities as well as the effects of helium bubble formation during irradiation. The models allow a unified theoretical approach to be taken in describing unirradiated, in-pile and post-irradiation creep rupture experiments, as well as Fuel Cladding Transient Tests (FCTT).