Mechanical Properties of Neutron Irradiated Fuel Cladding Tubes

There was few post irradiation examination data on the mechanical properties of domestic fuel cladding tubes used for light water reactors, then those data obtained abroad have been often used in the fuel design or fuel performance codes. Although, many reports discussed the deformation mechanism of the tube, almost all the data were not obtained from irradiated specimens but unirradiated ones. In recent years, systematic post irradiation examinations on domestic fuel elements used in Japanese light water reactors and the related studies were performed.

This report first summarizes briefly the crystallographic texture which characterizes the properties of Zircaloy fuel cladding tubes, followed by an explanation of basic properties such as elasticity, plasticity, creep and fatigue. Finally, the up-to-date results are introduced.     

Neutron Irradiation Effect on Magnetic Properties of Cr-, Mo- and Si-Doped FeNi-Alloys

Sample of FeNi, FeNi: Cr, FeNi: Mo, and FeNi: Si alloys were irradiated by neutron beams in the vicinity of a 2 MW reactor core and their magnetic hysteresis curves, magnetic after effects as well as variations of magnetic permeability with temperature were determined. From these curves Curie temperature shift, spectrum widthk, and activation energy for self-diffusion of Fe and Ni atoms were obtained.

The presence of Si impurities in the FeNi alloy produces a considerable attenuation in the vacancy supersaturation and the analysis of the spectrum width in this sample indicates that the defect structure due to neutron irradiations is more complex than that due to thermal effect.     

Effect of Neutron Irradiation on Deformation Behavior of Zirconium

The effect of neutron irradiation on the tensile deformation behavior of zirconium was examined at room temperature at various strain rates ranging of 2.2×10^?4~2.2× 10^?2 sec^?1. The microstructure of the deformed specimens was observed by transmission electron microscopy. It was established that neutron irradiation diminishes the uniform elongation and the strain hardening rate, and hastens the onset of plastic instability. These phenomena are attributed to inhomogeneous deformation in the dislocation channels in the irradiated and deformed zirconium.

From the relation between strain rate and tensile properties (yield stress, ultimate tensile stress, uniform elongation and strain hardening rate), it was established that in unirradiated zirconium deformation is controlled by slip at strain rates below 6×10^?3 sec^?1, while above this threshold, twinning as well as slip contribute to deformation.

Neutron irradiation markedly inhibits deformation twinning in zirconium at room temperature. At 77 K, on the other hand, deformation by twinning is more prominent in irradiated specimens. The mechanism of twinning inhibition due to neutron irradiation is discussed.     

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.     

Effect of Absorbed Hydrogen on the Stress Corrosion Cracking (SCC) Susceptibility of Unirradiated Zircaloy Cladding

Effect of absorbed hydrogen on the stress corrosion cracking (SCC) susceptibility of unirradiated Zircaloy cladding was examined. The data obtained from literatures showed that the normalized ratios of SCC threshold stress (σ_th ) to 0.2% yield stress (σ_0.2.) claddings, from which the influence of σ_0.2 had been eliminated, increased with increasing hydrogen contents below 50ppm in unirradiated Zircaloy-2 and -4. For Zircaloy-4, the break point was observed in the relationship between normalized ratios of σ_th to σ_0.2 and hydrogen content in sample at hydrogen content of approximately 50ppm. Thermodynamic calculations were carried out for the reaction between iodine gas and zirconium containing hydrogen. The results suggested that the reactions hardly occurred at increased hydrogen content and zirconium reacted with iodine gas only below 100 ppm of hydrogen. Since these tendencies correspond to those of the normalized ratios of σ _th to σ_0.2 on the hydrogen content, it is considered that hydrogen affects the reactions between iodine gas and zirconium and reduces the SCC susceptibility of Zircaloy cladding.     

Effect of Fast-Neutron Irradiation on Twinning Deformation of Zirconium

Unirradiated zirconium specimens and specimens irradiated to a fluence of 4.1 ×10¹⁹ n/cm² were subjected to tensile deformation at room temperature. During the straining, the specimens were removed at intervals from the testing machine for microphotographie inspection, to determine the progress of slip line density Ns, the twin volume fraction Vt and the twin density Ni, as function of strain. The effect of neutron irradiation on these values were studied in detail. The resulting data indicated that the neutron irradiation inhibited the growth of twins in the early stage of deformation, up to a plastic strain ?_p of about 11%, beyond which level the irradiation tended instead to inhibit nucleation. Between growth of twins and nucleation, it was the latter phenomenon that was found more effective in promoting slip deformation. The break in the slope of plots relating the reciprocal work hardening de/da to the true plastic stress σ_p gave the twin stress σ_t, which proved to represent a good approximation of the stress of twin nucleation.     

Effects of Temperature,Strain Rate,and Specimen Orientation on Localized Plastic Deformation of Irradiated Zircaloy-2

Both transverse and longitudinal Zircaloy-2 specimens irradiated up to 1.2 × 10²⁰ n/cm² (E> 1 MeV) were tested in tension with strain rates ranging 1.1 × 10^-4~1.1 × 10^-2 s^-1 in the temperature range 200~400°C. Detailed observations of the specimen wall surface and microstructure were also made on samples deformed to various amounts of plastic strain, with a projector and an optical microscope.

It was found that localized plastic deformation bands occurred in the temperature range approximately 280~330°C during straining to the ultimate tensile stress. Results also showed that the strain rate dependence of tensile properties, particularly the strain to the ultimate tensile stress, was associated with changes in the number and width of the localized deformation band with strain rates at a temperature of 300°C at which localized bands occurred. From a break of the straight line tracing the true stress-true plastic strain relationship, it was established that the onset stress and strain of the localized deformation band could be estimated.

The effect of specimen orientation on localized deformation band was also discussed on the basis of differences in the onset stress and strain between the transverse and longitudinal specimens.     

Separation of Transplutonium Elements from Neutron Irradiated Americium-241, (II)

The isotopes of transplutonium elements were produced by irradiation of 62 mg of ²⁴¹AmO₂ in JMTR during the period July 1971 through July 1972. The transplutonium elements were separated from plutonium, fission products and cladding materials by a combination of precipitation and ion exchange. Americium, curium, berkelium and californium were mutually isolated by ordinary procedure. The isotopic ratios of americium and curium were determined by mass spectrometry, and berkelium and californium by radiochemical method. In addition to the nuclides detected in the previous work, ²⁴⁶Cm, ²⁴⁹Bk and ²⁴⁹Cf were also identified.     

Irradiation Behaviors of Nuclear Grade Graphite in Commercial Reactor, (III) Mechanical Properties

Bending and compressive strengths, and Young's modulus were measured for Pechiney nuclear grade graphite irradiated in the temperature range 220～400°C in the environment of CO₂ in a commercial reactor, up to the neutron fluence 6.2 × 10¹⁹ and 2.2 × 10²⁰n/cm² (E>0.85 MeV), respectively.

All of them increased owing to neutron irradiation, and the changes in both strengths were almost similar in the whole range of irradiation temperature, however the changes in Young's modulus depended on irradiation temperature.

It was clarified in the present experiment that both strengths were related with Young's modulus and the relation could be expressed by the formula σ=kE ⁿ, where σ and E are strength and Young's modulus, respectively, and n is constant which has different value for bending or compressive strength and also for their measured direction.     

Gamma-Ray Irradiation Effect on Loss Increase of Single Mode Optical Fibers, (II)

In order to examine the reliability of optical fiber systems in nuclear power plants, we estimated the induced losses of pure silica core and GeO₂ added core single mode optical fibers (Si-SMF and Ge-SMF) exposed over a long period to low dose rate γ-rays (less than 100 R/h, 40 yr). Also γ-ray induced losses under a high dose rate and high temperature condition such as 10⁶R/h and 200°C were estimated in view of an accident.

In both cases, the induced losses were concluded to depend mainly on total dose from the results of low dose rate irradiation experiments and a study on temperature dependences of induced losses

It was ascertained from the estimation that the Si-SMF had less radiation sensitivity than the Ge-SMF in the radiation environment of nuclear power plants The induced loss of the Si-SMF exposed for 40 yr to 1 R/h γ-rays was estimated to be about 2 dB/km and almost five times lower than that of the Ge-SMF.     

Investigations on Sodium Boiling Accidents in a Large Fast Breeder Reactor, (I)

The radial propagation of sodium boiling by thermal process in a 1,500 MWe FBR is analyzed with the BOIP-T code. Since the model used embodies some uncertainty, several modes of molten fuel behavior are considered. The time required for boiling to propagate from one channel to its neighbor is calculated. Even on the most conservative basis, the boiling propagation requires at least 10 sec, which indicates that the radial propagation of sodium boiling could be prevented by a suitable core protection system.     

Development of Oxide Dispersion Strengthened Steels for FBR Core Application, (II)

Previously manufactured oxide dispersion strengthened (ODS) ferritic steel cladding tubes had inferior internal creep rupture strength in the circumferential hoop direction. This unexpected feature of ODS cladding tubes was substantially ascribed to the needle-like grain structure aligned with the forming direction. In this study, the grain morphology was controlled by using the martensite transformation in ODS martensitic steels to produce an equi-axial grain structure. A major improvement in the strength anisotropy was successfully achieved. The most effective yttria addition was about 1 mass% in improving the strength of the ODS martensitic steels. A simple addition of titanium was particularly effective in increasing the strength level of the ODS martensitic steels to that of ODS ferritic steels.