Recovery behavior in neutron irradiated β-sic

Reaction sintered β-SiC specimens were neutron-irradiated in the fast breeder reactors, JOYO, Rapsodie and Phenix, at temperatures around 430 to 550°C to a fluence from 4.0 × 10²³to 1.0 × 10²⁷ n/m² (E > 0.1 MeV). A change in the specimen length was accurately measured using a conventional micrometer after isochronal annealing at high temperatures. The specimens irradiated to a neutron fluence of about 10²⁵ n/m² showed a larger dimensional change on anneals than those irradiated to a fluence of 10²⁶ n/m². A change in lattice parameters by annealing was also measured. It showed that the temperature dependence is nearby identical to that of the macroscopic length change. The transmission electron microscopic study of the neutron-irradiated β-SiC specimens showed the formation of irradiation induced defects, considered to be dislocation loops. An increase in the neutron fluence resulted in the growth of the dislocation loops. The dislocation loops in strongly irradiated β-SiC interacted with each other, forming the heavily disturbed dislocation structure. The effect of the neutron fluence on the microstructure and dimendional change was discussed.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72?1.92 × 10²⁰ n/cm²($\text{E > 1}$ MeV) and 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV)at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72–1.92 × 10²⁰n/cm² (E>1 MeV) and 2.03×10²¹n/cm² (E>1 MeV) at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 ×10²¹ n/cm² (E>1 MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

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.     

Diffusion of Fission-Produced Xenon in Uranium Monocarbide

The diffusion of Xe atoms in UC at temperature from 800° to 1,400°C was studied on samples irradiated to various doses up to 2.9×10¹⁸ nvt. The apparent activation energy of the diffusion was found to vary with fluence, which corresponded to the extent of neutron or fission fragment damage in the specimen.

The diffusion was observed to be enhanced at exposures below 10¹⁷ nvt. At the higher doses, on the other hand, trapping of the gas atoms by defect clusters became evident, which inhibited the gas release.

When the irradiated samples were once heated up to 1,400°C, the activation energy of Xe diffusion through them showed a constant value of 83±5 kcal/mol, which was independent of the neutron fluence.

The scattered values of activation energy reported in literature are explained from the present results as resulting from differences in the activation energy of Xe diffusion, which depends on the extent of radiation damage retained in the UC.     

Combined effects of neutron irradiation and hydrogen absorption on mechanical properties of pure iron

To obtain a fundamental knowledge of the combined effect of neutron-irradiation and hydrogen, mechanical properties and the fracture mode were studied for pure neutron irradiated iron, followed by hydrogen charging. The effect of interaction between neutron irradiation and hydrogen absorption for a pure iron could be clarified. Under the hydrogen charged condition, the ductility is higher in the neutron irradiated specimen than in the unirradiated. The cause could be sought in hydrogen trap sites of the iron and the fracture mode. As a result of interaction between many irradiation defects and hydrogen atoms, the fracture mode of a hydrogen charged specimen after irradiation, is a mixed mode of quasi-cleavage crack and dimple pattern. That of a hydrogen charged unirradiated specimen is predominantly intergranular cracking.     

Measurement of Neutron-Production Double-Differential Cross Sections for Intermediate Energy Pion Incident Reaction

Neutron-production double-differential cross sections for 870-MeV π⁺ and π ^- and 2.1-GeV π⁺ mesons incident on iron and lead targets were measured with NE213 liquid scintillators by time-of-flight technique. NE213 liquid scintillators 12.7cm in diameter and 12.7cm thick were placed in directions of 15°, 30°, 60°, 90°, 120° and 150°. The typical flight path length was 1.5 m. Neutron detection efficiencies were derived from the calculation results of SCINFUL and CECIL codes. The experimental results were compared with the cascade-evaporation calculation code NUCLEUS. The calculation results are higher typically by a factor of two than the experimental data at neutron energies below about 30MeV. NUCLEUS overestimates π^+-incident neutron-production cross sections in forward angles at neutron energies of 100 to 500 MeV.     

Ternary Compounds PuNiC2 and PuCoC2

In order to determine a crack propagation rate of less than 10^-8 mm/s in a 24-hour integrated measurement, major parameters of a coupled system of a constant tension specimen and crack depth measurement, based on potential drop method, have been optimized. Influences of sensor geometry, location for detecting potential drop and data processing of the ratio of signal to noise (S/N) were optimized by applying Taguchi's Method. Then a suitable sensor geometry and data processing method were proposed to get a robust measurement system with higher sensitivity and lower susceptibility for geometrical and procedural fluctuations.

By applying the optimal crack propagation rate measurement system, it was confirmed that a crack propagation rate of lxlO^-8 mm/s can be measured under a low concentration condition of hydrogen peroxide with less than a 20% error by a 24-hour integrated measurement.     

Measurement of Thermal Neutron Capture Cross Section and Resonance Integral of the 129I(n, γ) 130I Reaction

To obtain fundamental data for research on the transmutation of nuclear wastes, the thermal neutron cross section and the resonance integral of the ¹²⁹I(n, γ)¹³⁰I reaction have been measured using an activation method. The neutron cross sections for the formation of the ground (5⁺) state and the isomeric (2⁺) state of ¹³⁰I were measured separately.

Six ¹²⁹I targets were irradiated for 10 min with thermal reactor neutrons; three of them containing 2.55- 2.61 kBq of ¹²⁹I were irradiated within a Cd capsule, and the other three targets containing 259–261 Bq of ¹²⁹I were irradiated without it. The Co/Al and Au/Al alloy wires were used to monitor the neutron flux and the fraction of the epithermal part (Westcott's epithermal index). The gamma-ray spectra from the irradiated samples were measured with a Ge detector.

The thermal neutron capture cross section (the 2,200 m/s neutron cross section) and the resonance integral of the ¹²⁹I(n, 7)¹³⁰I reaction were determined to be 12.5±0.5b and 15.6±0.7b for the formation of the ground state ^130gI(5+), 17.8±0.7b and 18.2±0.8b for the formation of the isomeric state ^130mI(2⁺), and 30.3±1.2b and 33.8±1.4b for the formation of ¹³⁰I (the sum of the 2⁺ and the 5⁺ states), respectively. The sum of the thermal neutron capture cross sections forming the 2⁺ and the 5⁺ states was 12% larger than the evaluated values of JENDL-3.2 and ENDF/B-VI and that reported by Roy et al. This discrepancy is explained by the population of the isomeric level.     

Neutron Irradiation Effects on Thermally Activated Process of Tensile Properties of Titanium

This paper deals with the relationship between mechanical properties and irradiation, effects for titanium irradiated to fast neutron fluxes. The neutron fluences applied are 6.9×10¹⁸, 8.6 × 10¹⁸ and 3.0 × 10¹⁹ n/cm². Tensile deformation is carried out over the temperature range of 77–about 600°K retaining the strain rate constant on one hand and changing the strain rate by a factor of about 5 and 10 on the other.

The fluence (φ) dependence of the yield stress at room temperature for an athermal component of the stress, σμ is greater than that for a thermal component σ* which does not change remarkably after irradiation. Their increments Δσ, Δμ and Δ σ* are proportional toσ ^1/3, σ^1/2σ^1/4 and, respectively.

The relationship between activation volume V* and effective shear stress τ* is investigated for both the unirradiated and irradiated specimens. In terms of the τ*/τ*₀ analysis (τ*_o is the value of τ* at T = 0°K), V* shows a tendency to decrease with increase in neutron fluence.

Irradiation defects observable by electron microscopy seem to be related to the athermal activation stress (σ_u) and those too small to be observed by electron microscopy to the thermal activation stress. The yield stress in the thermal activation can be given by Conrad's formula. The activation energy H₀ shows a constant value of about 1.8 eV irrespective of the neutron fluence applied. This value is 0.3–0.4eV higher than that for unirradiated specimens.     

Interpretation of the influences of irradiation upon fatigue crack propagation in austenitic stainless steels

An interpretation of the influences of neutron irradiation upon fatigue crack propagation in austenitic stainless steels is given. The approach has been to extend a previously developed rationalisation of the effects of various test and materials variables upon fatigue crack propagation in unirradiated stainless steels to include irradiated stainless steels.Irradiation has diverse influences upon the rate of fatigue crack propagation depending on the exact irradiation and test conditions. It has been shown that, by considering the underlying mechanisms of failure, some confidence is established in trends in data in a subject where information is very scarce and difficult to obtain.     

Effects of neutron irradiation on polymer matrix composites at 5 K and at room temperature: I. Absorbed-dose calculation

The spatial distribution of absorbed dose in a composite specimen irradiated in the Intense Pulsed Neutron Source (IPNS) was calculated for four kinds of cloth-filled polymer-matrix composites (filler: E-glass or carbon fiber; matrix: epoxy or polyimide resin). This calculation was performed by taking into account the range of recoil particles and the array of fibers in the composite. The average ratio of the energy of recoil protons deposited in a matrix of a composite to that deposited in an infinite matrix is 0.55–0.79, depending on the IPNS neutron spectra and on the kinds of composite materials. For E-glass fiber composites which have a ¹⁰B(n,α)⁷Li reaction taking place in the fiber, the average ratio of the energy of α and ⁷Li particles deposited in the matrix to that deposited in an infinite fiber material is about 0.79. On the basis of these ratios, the conversion factor from total neutron fluence to absorbed dose for a matrix of a composite is calculated for composite materials irradiated in IPNS.     

Evaluation of impact properties of weld joint of reactor pressure vessel steels with the use of miniaturized specimens

The effects of specimen size and location of V-notch on the Charpy impact properties were investigated with different sizes of specimens, standard, CVN-1/2, CVN-1/3, and CVN-1.5 mm, for A533B steel, low Mn, high Cu, high phosphorus (P), and high Cu/P steel weld joint. A part of the specimens was irradiated with neutron at 563 K up to 8 × 10¹⁹ n/cm². The heat affected zone (HAZ) specimen is the best in the impact properties among the specimens of base metal, HAZ, and weld metal in the steels with 0.003 wt.% P, while it is the worst in the steels with ～ 0.3 wt.% P. This indicates that the surveillance test of HAZ specimen can be represented by base metal in the case of A533B steels with lower P content (～ 0.003 wt.%). The effects of notch location and chemical contents on ductile to brittle transition temperature (DBTT) are almost independent of specimen size within an error of ±5 K, indicating that the miniaturized Charpy specimens are applicable and effective in the surveillance tests of reactor pressure vessel steel of extended operation period. After irradiation, the highest DBTT was observed for the specimen with V-notch in base metal in the case of A533B steel and high Cu steel with 0.003 wt.% P.     

The effect of fast neutron irradiation upon the fatigue-crack propagation behavior of Alloy 718 plate and weldments

The fatigue-crack propagation (FCP) behavior of Alloy 718 plate and gas-tungsten-arc weldments was studied at 427°C for specimens irradiated up to a total neutron fluence of $\text{7.6 × 10}{}^{22}\text{n/cm}{}^2$ (28 dpa). Two different precipitation heat-treatments were utilized: a “conventional” treatment and a “modified” treatment. No significant effect of neutron irradiation upon the FCP behavior was noted for any of the material/heat-treatment combinations, except for the conventionally-treated weldments where FCP rates were higher after irradiation.     

Electron Microscope Observations of Gas Bubbles in Neutron Irradiated Al-Li Alloys

The electron microscope has been used to observe the behavior of He gas bubbles in neutron irradiated Al-Li alloys. In the case of high He concentration, the gas bubbles were observed as small white or black dots in specimens as irradiated. The bubbles initiated appreciable growth upon heating to 400°C. They precipitated preferentially along the subgrain boundaries and dislocations, as well as along the grain boundaries. The size of the bubbles, observed in a specimen heated to 550°C, ranged from about 10 Å to 1 μ.

The shape of the bubbles in the specimen heated to 400°C was hexagonal or octagonal in the two-dimentional projection and a polyhedral image of the larger bubbles was clearly observed. The number of planes that bound the polyhedral bubble increased with increasing temperature of heating. Spherical bubbles were also observed.     

Effect of Neutron Irradiation on Iodine Stress Corrosion Cracking Susceptibility of Zircaloy-2 Tubing

The effect of neutron irradiation on the iodine stress corrosion cracking (SCC) of Zircaloy-2 tubing of 8×8 type design was studied by means of ring tension test, using specimens unirradiated and irradiated to 3.2×l0¹⁹ and 3.0×10²⁰ n/cm² (E>lMeV). The SCC threshold stresses were determined from constant load tests and the SCC initiation stresses were defined from constant cross-head rate tests. Both stresses increased with increasing neutron fluence, reaching a maximum at a neutron fluence between 10¹⁹ and 10²⁰ n/cm² and then decreased. The tendency is qualitatively in good agreement with the hypothetical conclusion derived by Lunde & Videm, for SCC failure stresses from internal gas pressurization tests on various Zircaloy cladding tubes irradiated at different reactor conditions. The cause of the increase in the SCC susceptibility at neutron fluences above 10²⁰ n/cm² would be ascribed to radiation anneal hardening phenomenon and resultant inhomogeneous incipient deformation characterized by dislocation channels.     

Simplified Method for Calculation of Neutron Capture Transformation Effects of Fission Products on Decay Power

A simplified method is proposed for the calculation of the effects of neutron capture transformations of fission products (FPs) on the decay power of FPs. The decay power of FPs after shutdown changes by the neutron capture transformations of FP nuclides during reactor operation. It is proposed to calculate the neutron capture transformation effects considering the production of the following 7 nuclides ¹⁰³Ru, ¹³⁴Cs, ¹³⁶Cs, ^148mPm, ¹⁴⁸Pm, ¹⁵⁴Eu and ¹⁵⁶Eu by the neutron capture reaction of the direct mother nuclide alone giving a cumulative fission yield for the mother nuclide. The present method was assessed by com-paring the calculation results with the rigorous calculation results for the thermal-neutron fission of ²³⁵U irradiated between 1 and 5 yr in a light water reactor with thermal-nentron flux between 3 x 10¹³ and 6 x 10¹³ n/cm²·s and for the fast-neutron fission of ²³⁹Pu irradiated between 1 and 5 yr in a fast breeder reactor with total neutron flux between 3 x 10¹⁵ and 6 x 10¹⁵ n/cm²·s. It has been clarified that the present method can calculate the neutron capture transformation effects within the accuracy of ±1% of the decay power for the irradiation of 1yr and cooling time less than 10⁹s irrespective of fission type and neutron flux. The accuracy varies little with neutron flux but considerably with irradiation time. For a irradiation of 5 yr the present method can calculate the capture effect within the accuracy of +1% and -5% of the decay power. The accuracy can be improved to ±1% of the decay power with the simple correction factors.     

Experimental Study on the Importance of Accurate Neutron Angular Distribution Measurement for Determination of Neutron Yield in a 2.5 kJ Mather type PF

Three activation counters were constructed by thin wall Geiger and were placed in different angels with respect to anode axis; 0°, 45°, 90°; where the Geigers have 20 cm distance from anode top of the 2.5 kJ SBUPF1 plasma focus device. The counters were calibrated by a 5 Ci Am–Be neutron source with source removal method. A computer program receives the Results of each plasma focus experiment and background count rates, estimates the angular distribution of neutron emission and calculates the neutron yield of each shot. The neutron yield of the device was measured about 6 × 10⁷ neutrons per shot. The results, indicate that using three counters (instead of one counter in 0°) in different angles for determination of total neutron yield, gives a more accurate measurement (up to 28% is measured in sample shots), and this error is bigger in those shots that thermonuclear fusion mechanism have greater share in neutron production.     

Combined effects of radiation damage and hydrides on the ductility of Zircaloy-2

Interest remains high regarding the effects of zirconium hydride precipitates on the ductility of reactor Zircaloy components, particularly in irradiated material. Previous studies have reported that ductility reductions are much greater at room temperature compared to reactor component temperatures. It is often concluded that the effects of irradiation dominate the ductility reduction observed in test specimens, although there is no consensus as to whether hydriding effects are additive. Many of the tests reported in the literature are difficult to interpret due to variations in test specimen geometry and material history. In this paper, we present the results of an experimental program aimed at clearly describing the combined effects of irradiation and hydriding on ductility parameters under conditions of a realistic test specimen design and well characterized hydride content, distribution and orientation. Experiments were conducted at 295 and 605 K, respectively on Zircaloy-2 tubing segments containing 10–800 ppm hydrogen and neutron fluences between 0–9×10²⁵ n m⁻² (E>1 MeV). Tests utilized the well proven localized ductility specimen which applies plane strain tension in the hoop direction of the tubing segment. In all cases, hydrides were also oriented in the hoop or circumferential direction and were uniformly distributed across the tubing wall. Results indicate that at 605 K, the ductility of irradiated material was almost independent of hydride content, retaining above 4% uniform elongation and 25% reduction in an area for the highest fluences and hydrogen contents. Even at 295 K, measurable ductility was retained for irradiated material with up to 600 ppm hydrogen. In the paper, results of fractographic analyses and strain rate are also discussed. We conclude that at reactor component operating temperatures, radiation damage controls the ductility of Zircaloy-2 for conditions of these tests up to hydride levels of at least 800 ppm, and probably much higher. At room temperature the effect of hydride content and radiation damage appear to be additive.     

Particle Release from Vanadium Metal Surface by Neutrons

Sputtering yields from vanadium metal surface due to neutron irradiation were studied. A carefully prepared Pyrex glass tube, containing a vanadium foil as target and a polyethylene film pasted on a nickel plate as catcher, was sealed after evacuation, irradiated in a reactor, disassembled to take up the film, and the ⁵²V activity on it was counted for estimating the thermal neutron sputtering yield due to the recoil by(n, γ) reaction. The reactivation of the film gave the fast neutron sputtering yield. These values were found to be 2.3×10^?9 and 2.1×10^?1 respectively.