Takeo Onchi

The inhomogeneous deformation behaviour of neutron irradiated Zircaloy-2

Annealed Zircaloy-2 specimens irradiated to 3.2 × 1019 n/cm2 (E > 1 MeV) at ≈425 K were tensile-tested, together with unirradiated material, between 298 and 673 K in vacuum. The surface and microstructure of deformed specimens were observed using projector, optical and transmission electron microscope. Metallographie examinations or irradiated samples showed that localized deformation bands occur at intervals during deformation to the ultimate tensile stress between 473 and 623 K, while in the room temperature deformation, the inhomogeneity in metallographic features was characterized by microscopic dislocation channeling structure, without showing localized band on the surface of deformed specimen. From the shape of the stress-strain curves and metallographic features, it was concluded that the first localized band occurred after a small amount of strain, resulting in yielding and that the flow stress increased monotonically with further yielding. Evidences from temperature dependent mechanical properties of irradiated and unirradiated samples indicate that the radiation-anneal hardening phenomena are of significance in a range of temperature at which localized bands are formed, particularly pronounced at 553–593 K.

Creep deformation and rupture properties of unirradiated Zircaloy-4 nuclear fuel cladding tube at temperatures of 727 to 857 K

Abstract Creep deformation and rupture properties of an unirradiated Zircaloy-4 tube were examined at temperatures of 727 to 857 K to obtain data for evaluating spent fuel integrity under the off-normal dry storage condition. Creep tests were carried out on internally gas pressurized tubular specimens with end plugs welded to both ends. The hoop creep strain up to the steady-state creep region was given by the following equation: ϵ = 0.05{1- exp [−10( ϵ s t ) 0.51 ]} + ϵ s t , where ϵ s = 1.02 × 10 5 ( E / T ) exp (4060 σ / E ) exp (− Q / RT ) ( S −1 ), t is time (s), E is the elastic modulus (MPa), T is the temperature (K), σ is the applied stress (MPa), R is the gas constant (= 8.314 J/mol K), and Q the apparent activation energy (233 kJ/mol). The total creep deformation curve over the accelerated creep region could also be presented by the equation expressed as a function of true stress formulated on the basis of constant stress creep data. It is inferred that the transition from the steady-state to the accelerated creep region occurs without changing the deformation mechanism. It was also shown that most of the creep strain to rupture exceeded 100% and Zircaloy-4 had an extremely ductile rupture property.

Creep deformation of an unirradiated zircaloy nuclear fuel cladding tube under dry storage conditions

Abstract Measurements of creep deformation were made on an internally gas pressurized tubular Zircaloy-4 specimen with plugs welded to its ends. Creep tests were conducted at temperatures between 577 and 693 K for holding times of up to 26640 ks, to formulate the creep equation needed for predicting creep strain during dry storage of spent fuel. Discussion was also given to the difference of creep behaviour between irradiated and unirradiated fuel cladding, indicating that the equation derived is applicable for predicting creep strain of spent fuel cladding during dry storage.

Intergranular Cracking of Irradiated Thermally Sensitized Type 304 Stainless Steel in High-Temperature Water and Inert Gas

Abstract Specimens of solution-annealed (SA), commercial-purity type 304 (UNS S30400) stainless steel (SS) with a high carbon content were thermally sensitized by heat treatment in a two-step process: a high-temperature sensitization at 1,023 K followed by a low-temperature sensitization at 773 K. The specimens then were irradiated to 3 × 1023 n/m2 (E > 1 MeV) at 563 K in boiling pure water. After irradiation, the materials were tensile tested at room temperature (RT) and at 563 K in inert gas and then tensile tested by the slow strain rate (SSR) method in 563 K water with different dissolved oxygen (DO) concentrations and in inert gas. Results were compared with those for unirradiated sensitized materials. The irradiated thermally sensitized SS revealed intergranular cracking (IGC) when tensile tested at RT and at 563 K in argon gas. Intergranular stress corrosion cracking (IGSCC) susceptibility of thermally sensitized specimens in high-temperature water of different DO concentrations was enhanced by neu...

Irradiation effects of boron carbide used as control rod elements in fast breeder reactors

Abstract The paper describes radiation effects on 84C pellets used as control rod elements in the Enrico Fermi Fast Breeder Reactor. Pellet swelling (Δ V/V ) caused by irradiation was less than 1% in which crystal lattice swelling was less than 20%. Many microcracks, a main cause of pellet swelling, appeared in the irradiated pellets. The production of microcracks was related to graphite precipitation in the pellets before irradiation. Open pores which did not exist in the unirradiated pellets were formed in the irradiated ones. In a unit cell of B4C, the α-axis elongated by 0.025 A and the c- axis shrank by 0.07 A by irradiation. Moreover, we found three recovery stages which were from room temperature to 400°C, from 400 to 750°C and from 850 to 1100°C. The recovery mechanisms in the irradiated pellets are discussed in terms of the helium behavior.

Effect of fast-neutron irradiation on deformation twinning in zirconium deformed at 77 K

Abstract The effect of neutron irradiation on twinning deformation is investigated at 77 K using unirradiated and to a fluence of 4.1 × 1019n/cm2 irradiated zirconium. The slip line density Ns, the twin volume fraction Vt and the twin density Nt for both specimens were determined as a function of strain by means of optical microscopy. The twins were determined using the diffraction profiles of {1012} twins at the early stages, as well as, {1122} and {1124} twins at the latter stages of deformation. It is shown that the nucleation of twins was affected more strongly than the growth by neutron irradiation.

Crack Initiation Mechanism in Non-ductile Cracking of Irradiated 304L Stainless Steels under BWR Water Environment

The deformation behavior and initiation mechanisms of intergranular (IG) and transgranular (TG) cracks in irradiated 304L stainless steel were studied by slow-strain-rate tensile tests in inert gas and simulated BWR water environments, followed by fractographic and microstructural examinations. Neutron irradiation was made in test reactors to fluences of up to 6.2x1020 n/cm2 (E>1 MeV). Intergranular cracking occurred in water above a critical neutron fluence of around 1 × 1020 n/cm2, based on the results of the SSRT tests and SEM fractography. That critical fluence is mechanistically supported by irradiated, deformed microstructures exhibiting dislocation channeling at that fluence, while radiation-induced Cr depletion at the grain boundaries was minor. Transgranular cracking of the irradiated material occurred in water below the critical fluence, initiating in the non-uniformly strained surface region of the test bar in the later stages of plastic deformation. The initiation of TG cracking is hypothesized to be related to a high density of deformation twins. Intergranular cracking is proposed to have initiated where localized slip bands terminated at grain boundaries, while TG cracking is inferred to have initiated at deformation twin boundaries. High stress and strain concentrations at grain/twin boundaries would be the common cause of non-ductile crack initiation.

Correlation of Intergranular Stress Corrosion Cracking Susceptibility with Mechanical Response of Irradiated, Thermally Sensitized Type 304 Stainless Steel

Abstract Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated, thermally sensitized type 304 (UNS S30400) stainless steels (SS) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 5.4 × 1023 n/m2 (E > 1 MeV) at 290°C and to 1.1 × 1024 n/m2 (E > 1 MeV) at 340°C in the Japan Material Test Reactor. Irradiated specimens were examined by slow strain rate tensile testing (SSRT) in 290°C pure water of 0.2 ppm dissolved oxygen (DO) concentration, by SSRT at 290°C in argon gas, and by microhardness measurements. IGSCC susceptibility increased with neutron fluences up to 1.1 × 1024 n/m2 regardless of irradiation temperatures. The 0.2% proof stress (PS) and the grain-boundary (GBHV) and grain-center microhardness (GCHV) increased. The uniform elongation (UE) decreased with neutron fluences up to 5.4 × 1023 n/m2, but 0.2% PS and GBHV and GCHV decreased, and UE increased at 1.1 × 10...

The applicability of the strain-hardening rule to creep deformation of Zircaloy fuel cladding tube under dry storage condition

Abstract Variable stress and temperature creep tests were carried out at temperatures between 626 and 693 K for the Zircaloy-4 tubular specimens in the hoop stress regime 55 to 125 MPa to examine the applicability of the strain hardening rule to creep deformation of Zircaloy-4 fuel cladding tube. Results were compared with those of the constant temperature/stress tests, indicating that the strain hardening rule but not the time hardening rule can be applicable for predicting creep deformation of fuel cladding tube at temperatures tested. It is also suggested that although Zircaloy-4 fuel cladding tube tends to deform under multiaxial condition, it can satisfy the requirements for the strain hardening rule at temperatures of interest during dry storage condition.

Mechanical response of irradiated thermally-sensitized type 304 stainless steels

Abstract Thermally sensitized type 304 stainless steels, together with solution annealed materials, were irradiated in inert gas environment to neutron fluences up to 5 × 1023 n/m2 (E > 1 MeV) at 563 K, and to 1 × 1024 n/m2 at 613 K, and then examined by a slow strain rate (SSR) tensile test and a microhardness test. The SSR tensile test results showed that 0.2% proof stress of the thermally sensitized materials unirradiated and irradiated to lower neutron fluences of ∼ 2 × 1023 n/m2 was larger than that for the solution annealed one, whereas the former irradiated to the high neutron fluence above ∼ 2 × 1023 n/m2 was smaller than the latter at the same neutron fluences. Microhardness at grain boundary regions was larger than that for grain centers in the irradiated as well as unirradiated thermally sensitized materials. The larger grain boundaries hardening could be attributed to radiation induced segregation of interstitial impurity elements and their interaction with radiation defects. The hardening of the grain boundaries relative to the grain centers increased with neutron fluence. The above neutron fluence dependent 0.2% proof stress of the thermally sensitized and solution annealed materials would primarily be ascribable to the larger hardening of the grain boundary area than that for the grain center. Radiation anneal hardening appeared at 600–650 K and 700–750 K upon post irradiation annealing of thermally sensitized materials.