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.     

Determination of Trace Impurities in High Purity Lutetium Oxide by Neutron Activation with Aid of Cation-Exchange Separation

Trace impurities including Yb in commercially available LU₂O₃ of high purity were determined by neutron activation followed by cation-exchange separation with α-hydroxyisobutyrate.

After irradiation in reactor, the LU₂O₃, together with about the same quantity of Yb carrier, was dissolved in HC1, and fed to a cation-exchange resin column. Lutetium, Yb, Tm and some other rare earths were eluted successively, using α-hydroxyisobutyrate and citrate solutions as eluants. The addition of Yb carrier facilitated the analysis of Yb which is difficult to separate from Lu, and this permitted the determination of Yb of quantities as small as a few ppm.

The impurities found present in this sample of LU₂O₃ were: 0.26ppm Sc, 395ppm Yb, 150 ppm Tm, 12 ppm Tb and 2.3 ppm Eu.     

Compatibility of Eu2O3 and (Eu,Gd)2O3 with Type 316 Stainless Steel

The compatibility of Eu₃O₃ and (Eu,Gd) ₂O₃ mixed oxides which contain 0, 1, 5, 10, 50 and 100 mol% of Gd₂O₃, with type 316 stainless steel was examined by out-of-pile heatings. The heating temperature ranged from 550 to 1,000°C and heating time from 500 h to up to 5,000 h. Metallographic examinations of the stainless steel indicated that chemical reaction between Eu₂O₃ and stainless steel occurred at 650°C and that grain boundary penetration extended up to 40 μ m after heating at 1,000°C for 500 h. The penetration depth Δ was expressed as a function of heating temperature T as

Δ = 3.08 × 10⁴ exp(?1.72 × 10²RT)

where R is the gas constant and the activation energy is given in cal/mol. Electron probe microanalysis and X-ray diffraction analysis indicated that main reaction product is europium silicate. When Gd, which is a decay product of Eu, is mixed with Eu₂O₃, the degree of reaction between (Eu,Gd) ₂O₃ and stainless steel decreases as increase in concentration of Gd₂O₃. It was found that actually no grain boundary attack was observed even after heating at 1,000°C for 500 h when concentration of Gd₂O₃;i in (Eu,Gd) ₂O₃ exceeded 10 Discussion was made of reaction mechanism of Eu₂O₃ and (Eu,Gd) ₂O₃ with stainless steel.     

Reactor Neutron Capture Cross Section of 60.3-day 124Sb

The radioactive nuclides ¹²⁴Sb (T _1/2=60.3d) and ¹²⁵Sb (T_1/2=2.77yr) were produced from natural antimony by JRR-3 reactor irradiation of 283.5 h through the single and double capture processes. After cooling of 3.50 yr, the γ-ray spectrum of the antimony sample irradiated was measured by a 50 cc coaxial type Ge(Li) detector, and the photo-peak yield ratio of ¹²⁵Sb (E _r=428keV) to ¹²⁴Sb (E _r=1.691 MeV) was obtained. By using a relation between this photo-peak yield ratio and the ¹²⁴Sb (n, γ) ¹²⁶Sb cross section, the reactor neutron capture cross section of 60.3-day ¹²⁴Sb was obtained as 17.4:5:^+2.8 _?2.5b. The thermal neutron flux at the position of antimony sample irradiated was estimated as (4.92±0.38) ×10¹²n/cm²·s by measuring the 1.333-MeV photo-peak yield of ⁶⁰Co, which was activated by reactor irradiation of cobalt impurity contained in the antimony sample.     

Comparison of Calculated Values with Measured Values on the Amount of TRU and FP Nuclides Accumulated in Gadolinium Bearing PWR Spent Fuels

Destructive analyses for five spent fuel samples taken from a Gd bearing fuel assembly were done. The measured amounts of actinides of ^234-238U, ²³⁷Np, ^238-242p_u 241,242m,243Am ^242,244Cm, and fission products of ¹³⁴Cs and ¹⁵⁴Eu were used for evaluating the accuracy of calculation made by CASMO-MICBURN and ORIGEN-2 codes. The effect of Gd on the neutron spectrum was taken into account in the CASMO-MICBURN calculation.

The amounts of ²³⁵U, ²³⁹Pu and ²⁴¹Pu calculated by CASMO-MICBURN agreed well with the observed values within about 3%. On the other hand, the amounts obtained from ORIGEN-2 calculation showed lower values than those observed, especially by —12% in average in ²³⁵U for Gd₂0₃U0₂ fuel. The main cause of this large difference may be attributed to the effect of Gd on the neutron spectrum. The amounts of the other actinides by both calculation codes revealed no significant difference in nearly 10% except for ^242mAm, in which a large fluctuation among the samples was observed. About 10% difference between the measured values and the calculated values was also observed for ¹³⁴Cs, but the calculated values for ¹⁵⁴Eu showed a significant difference from measured values.     

Effect of Neutron Irradiation on Mechanical Properties of Al-Li Alloys

An increase in yield stress at room temperature was observed in Al-0.6W/0 Li alloy irradiated to thermal neutron doses of 2.9 × 10¹⁹ to 7.2 × 10¹⁹ cm^?2. The hardening of as-irradiated specimens is accompanied with yield point followed by jerky yield-elongation in the stress-strain curve. The radiation hardening could not be annealed out by heating for 30 min at temperatures up to 350°C, whereas the yield-elongation disappeared gradually with increasing heating temperature in the l mm diam. specimens; with the 2 mm diam. specimens the yield-elongation still remained even after post-irradiation heating for 30 min at 350°C. Strengthening accompanied by jerky yield-elongation is considered to be due to He atom clusters precipitated along the dislocation. The hardening observed in the specimens heat-treated after irradiation at temperatures above 250°C is caused by randomly distributed gas bubbles.

In heavily cold-worked Al-0.6%W/o Li specimens, recovery of work hardening occurred during neutron irradiation to 4.2 × 10¹⁹ cm^?2. Hardening due to gas bubbles was also observed in the cold-worked specimens. In Al-2.7W/0 Li alloy, an increase in yield stress took place in the specimens irradiated to 4.2 × 10¹⁹ cm^?2 and heated for 30 min at temperatures of 155° to 260°C. The hardening is thought to be due to re-precipitation of β-phase resolved during the neutron irradiation.     

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.     

Experimental study of pulse neutron irradiation damage in SiGe HBT

The response to transient irradiation of npn SiGe HBT (BG35 SiGe BiCMOS), i.e. device under test (DUT) was studied with online measurement of 1 MeV equivalent pulse neutron fluence of 0.8 × 10¹³n/cm². The differently biased DUT1 and DUT2 in test circuit were irradiated in the first day with neutron fluence (0.8 × 10¹³n/cm²) termed as Fluence1 and with an additional neutron fluence (0.8 × 10¹³n/cm²) in the second day to make Fluence2 equals to 1.6 × 10¹³n/cm². The experimental results show that pulsed neutron irradiation causes voltage surges in the DUTs exhibited by a negative and positive peak known to be radiation damage (RD). The RD in DUTs induced by pulse neutron Fluence1 initially created unstable displacement defects and the defects later reordered (cluster defects) to form more stable configurations via neutron Fluence2. The irradiated DUTs experienced online instantaneous annealing after 2.03 × 10^?9 s and offline measurement (i.e. without irradiation) of DUTs showed recovery to normal mode of operation after 24 h annealing. The level of pulse peaks in the base voltage terminals of DUT1 and DUT2 were compared as V_b2Fluence1 > V_b1Fluence1 and V_b2Fluence2 > V_b1Fluence2. A comprehensive analysis of RD region in DUTs with reference to Area (A₁, A₂), Peak (P₁, P₂), Height (H), and Full width at half maximum (FWHM) were investigated.     

Gadolinium Isotope Separation by Cation Exchange Chromatography

Displacement chromatographies of Gd adsorption band in cation exchange resin were performed to observe the isotope effects in the Gd ion exchange processes involving complex forming reagents. The heavy isotope of ¹⁶⁰Gd was found to be enriched at the front boundary of Gd adsorption band and the lighter isotopes of 1MGd, ¹⁵⁶Gd and ¹⁵⁷Gd were enriched at the rear boundary in both cases of 20.1m elution with EDTA and 14 m elution with malic acid, as predicted in the theoretical relations. Observed separation coefficients are 4.9×10^?5, 4.0×10^?5 and 2.5×10^?5for isotopie pairs of ^{156 160}Gd, ¹⁵⁸Gd and ¹⁶⁰Gd, respectively, in the case of EDTA elution. In the case of malic acid elution, smaller separation coefficients were observed as 1.8×10^?5, 1.6 ⁵O^?5 and 0.92×10^?5 for isotopie pairs of ^{156 160}Gd, ¹⁵⁷Gd and^{158 160} respectively.     

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.     

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.     

In-Pile and Out-of-Pile Grain Growth Behavior of Sintered UO2 and (U,Gd)O2 Pellets

Grain growth behavior of UO₂ and (U, Gd)O₂ fuel pellets was investigated with the data from the out-of-pile isothermal heating experiments and the irradiation test at the Halden Boiling Water Reactor. The laboratory data gave best-fitted equations by employing the following fourth power rate equations :

UO₂ : D²-D⁴ ₀=3.79×10¹⁸ exp(-142,000/RT)t,

(U, Gd) : D²-D⁴ ₀=4.98×10¹⁷ exp(-140,000/RT)t,

where, D ₀ and D are initial and final three-dimensional diameters (μm), respectively, R the gas constant (=1.987 cal/mole/K), T the absolute temperature (K) and t the time (h) (gadolinia content: 3~10%, temperature range: 1,700~2,000°C).

The calculated grain diameter with the above equations revealed an overestimation on specimens which involved noticeable fission gas bubbles on their grain boundaries. It was demonstrated that the in-pile grain growth model, as was given in the following equation, which took account of the retarding effects of growth by precipitated intergranular bubbles could describe the grain growth of the irradiated samples :

where f: Grain boundary fractional coverage (-).     

Neutron Irradiation Effects on Superconducting Nb-Ti Wires for Fusion Reactor

A systematic study has been carried out on Nb-Ti wire of two different kinds of metallurgical structure, to examine the effects of neutron irradiation on the critical density J_c. The samples used were Nb-47.6 a/0 Ti (sample #A) and Nb-59.8a/0 Ti (sample #B), which were aged at 380°C for 0~10⁴min and irradiated to 1.3×10¹⁸n/cm² (E _n0.1 MeV). The sample temperature during the irradiation is believed to have been below 70°C. The values of J_c of both #A and #B aged up to 50 min were found to increase with irradiation. But when aged beyond 100 min, #B had its value of J_c lowered by the irradiation. The presence of Ti enriched precipitates such as α and ω phases in the samples was surmised from the behavior shown by the critical temperature T_c. The T_c of f #A and #B changed little by irradiation when aged not longer than 100 min, but with aging beyond 500 min, #B showed a decrease in its value of T_c. This decrease indicates that the Ti concentration in the matrix may have increased through radiation-induced breakup of the above-mentioned precipitates, which, in turn, would have brought about the reduction observed in J_c upon irradiation. It is concluded that superconducting Nb-Ti wire with J_c endanced by precipitation does not appear very resistant to neutron irradiation. This underlines the importance of the choice of superconducting materials to be used in fusion reactor magnets.     

Steady-State Power Coefficients of Fast Reactor Considering the Effects of Partial Load and Flow Control

High resolution neutron transmission and low background capture measurements were carried out on the separated rubidium isotopes, using the time-of-flight facility of the linear accelerator of Japan Atomic Energy Research Institute. Resonance parameters and associated quantities were deduced as follows:

For ⁸⁵Rb, gΓ _n values were determined for 138 resonance levels in the energy region below 18.5 keV. s-wave strength function was obtained to be S₀=(0.94±0.11)×10^?4, average level spacing ≤D>=133±11 eV and average radiative width ≤Γ _r >=328±18 me V. For ⁸⁷Rb, gΓ _n values were determined for 30 resonance levels in the energy region below 48.6 keV and the following quantities were deduced:

S₀=(1.15±0.3)×10^?4, D=1,380±250 eV and ≤Γγ>=166±30 meV.

For ⁸⁵Rb average properties of resonances are in good agreement with the prediction of the statistical model. On the other hand, for ⁸⁷R9b the average properties of resonances deviate from the prediction of the statistical model; four strong s-wave resonances cluster within an energy interval of 5 keV, and they carry about 37% of s-wave strength below 48.6 keV.     

Neutron capture cross section measurements of 151,153Eu using a pair of C6D6 detectors

We have measured the neutron capture cross sections of ¹⁵¹Eu and ¹⁵³Eu by the time-of-flight (TOF) method in the range from 0.005 eV to keV region using the Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC). We employed a pair of C₆D₆ liquid scintillators for the prompt capture γ-ray measurement. The pulse-height weighting technique was employed to obtain the capture yields from the γ-ray spectra of ^151,153Eu. The obtained thermal cross sections at 0.0253 eV are 9051 ± 683 b for ¹⁵¹Eu and 364 ± 44 b for ¹⁵³Eu, respectively. The resonance integrals have been derived as 3490 ± 162 b for ¹⁵¹Eu and 1538 ± 106 b for ¹⁵³Eu.

The obtained capture cross sections were compared with the previously reported experimental data and the evaluated data. The evaluated data in JENDL-4.0 and JEFF-3.2 show good agreement with the present experiment results of ¹⁵¹Eu, however, the evaluated data in ENDF/B-VII.1 are larger than the present experiment results of ¹⁵¹Eu about 10% to 20% in the energy region from 0.03 to 0.2 eV. For the neutron capture cross sections of ¹⁵³Eu, the evaluated data in ENDF/B-VII.1 and Widder's data are in good agreement with the present results in the energy region below 0.35 eV.     

Irradiation Effect on the Tensile Properties of Magnox AL80

A study has been made on the effect of neutron irradiation on the tensile properties at room temperature of Magnox AL80, Mg-Al alloy and unalloyed magnesium as a function of grain size. Heat treated wire specimens were encapsulated and then irradiated in the JRR-2 reactor up to exposures of 1.2x10¹⁹nvt (epi-Cd energy). As a result the parameters σ₀ and k in the Petch relationship increased in magnitude after irradiations exceeding 1x10¹⁷nvt. Among the materials tested Magnox AL80 was found to be the most sensitive to neutron irradiation and unalloyed magnesium the least effected.

The effect of neutron irradiation on the strain rate (ε) dependence of the tensile stress (σ) was also investigated. The effect of neutron irradiation was to increase A and to decrease B in the equation

σ=A+B log ε,

when σ represented the yield stresses.     

Heat Capacity of Thorium Nitrides from 450 to 850K

The throium nitrides (ThN and Th₃N₄) were prepared by solid-gas reaction with thorium hydride and nitrogen.

The heat capacity of these samples were measured from 450 to 850K by a high-temperature double- adiabatic calorimeter and were determined as functions of temperature as follows:

Th₃N₄: C _p=41.30–8.47×10^?3 T-6.37×10⁵ T ^?2

ThN: C _p=12.50–2.66×10^?3 T-2.44×10⁵ T ^?2     

Enrichment of Phosphorus-32 by Activation of Tripolyphosphate

Enrichment of ³²P was attempted by Szillard-Chalmers effect on a neutron irradiated tripolyphosphate hexahydrate in the first eluate from anion-exchange resin column by chromatographic separation with gradient elution. Forty to fifty % of radioactive ³²P was foand as orthophosphate, having a specific activity of around 100 mCi/g of phosphorus at the end of about 20 min irradiation with a neutron flux of 5 × 10¹²n/cm²-sec in the KUR 1 MW reactor. The increase of specific activity is straightforward for a short period of irradiation. The enrichment factor was several thousand when recrystallized tripolyphosphate hexahydrate was used as target compound containing 0.02% of orthophosphate-P as impurity. Radiation effect, radiolysis by γ-rays and fast neutron, thermal degradation and isotopie exchange effect were examined as factors that might affect the enrichment. The relation between the specific activity of the isolated phosphates and the exposure time was estabished from a series of experiments in which the time of irradiation was varied.     

Construction of Neutron Guide Tubes in Upgraded JRR-3

Five neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No. 3). Two of them are for thermal neutrons and the other three are for cold ones. The characteristic wavelength of the thermal neutron guide tubes is 2 Å, and those of the cold neutron guide tubes are 4 and 6 Å. The longest guide tube is 59.9 m long and the total length of guide tubes is 232.1 m.

The beam sizes are 2 cm × 20 cm for the thermal neutron beams and 2 cm × 12 cm for the cold neutron beams. A curved part of the neutron guide is assembled by a polygonal approximation with use of 85 cm long straight units. The neutron mirrors of these units are made of natural Ni deposited borosilicate glasses. The Ni layer is about 2,000 Å in thickness.

The mean fabrication error of guide tube units is 4 μm. The mean installation errors are 8 μm for the positional abutment error and 5 × 10^?6 rad for the angular error. The neutron losses by these errors will be about 5%, and the neutron fluxes at the exits of the neutron guides are estimated to be about 2 × 10⁸ n/cm²·s.     

Void Swelling of Solution-Annealed Type 316 Stainless Steel under Long Time Proton Irradiation

Solution-annealed type 316 stainless steel was irradiated by 150 keV proton to a dose of about 6 dpa at the irradiation temperature ranging 450–700°C. To examine the effect of aging during irradiation, the present proton irradiation was carried out for about 25 h at a low dose rate of 7×10–^?5dpa/s. The specimens without He preinjection showed much smaller void swelling than those preinjected with He to the content of 10 at.ppm. Similarly to the case of neutron irradiations, the void swelling in the He preinjected specimens showed the temperature dependence with double peaks, and the peak swelling temperatures were about 550 and 650°C. In these specimens with He preinjection. void number density decreased and average void diameter increased with the increase of irradiation temperature in the range of 450–600°C, but these trends were reversed between 600 and 650°C. The volume of the grain boudary M₂₃C6 precipitates increased with the increase of irradiation temperature from 600 to 700°C, and it was concluded that the decrease of soluble carbon due to the precipitation of M₂₃C6 caused the second swelling peak at 650°C.