Neutron transmission measurements were carried out on the separated isotopes of silver using the time-of-flight facility at the Japan Atomic Energy Research Institute electron linear accelerator. Neutrons were detected with the ⁶Li-glass detectors at 56 and 191 m. The samples used were metallic powder enriched to 98.2% for ¹⁰⁷Ag and 99.3% for ¹⁰⁹Ag. Transmission data were analyzed with the multi-level Breit-Wigner formula incorporated in a least squares fitting program. Resonance energies and neutron widths were determined for the large number of resolved resonances in the neutron energy region of 400 eV7 keV. The s-wave strength functions and average level spacings were obtained to be ; S₀= (0.43±0.05)×10^-4, D₀=20±2 eV for ¹⁰⁷Ag and S₀= (0.45±0.05)×10^-4, D₀= 20±2 eV for ¹⁰⁹Ag.

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Transient behavior consequent to external loading faults is studied numerically on four configurations of high-interaction MHD generators-subsonic Faraday, supersonic Faraday, subsonic diagonal and supersonic diagonal, to provide a variable data base to serve in selecting the type of large-scale MHD generator. Time-dependent one-dimensional Navier-Stokes equations are solved with the 1969 MacCormack method, in combination with the Maxwell equations and the generalized Ohm's law. An artificial viscosity term is added to the Navier-Stokes equations to maintain numerical stability. It is shown that, with both supersonic and subsonic flows, the Faraday generator is liable to sustain more harmful effect from short than from open faults of the external loading circuit. For large-scale diagonal types, on the other hand, open faults are more dangerous. With subsonic flow, a shock wave propagating upstream is induced by short fault in the Faraday, and by open fault in the diagonal-type generator. In the case of supersonic flow, propagation upstream of the disturbance is completely obstructed. Larger electrical stress is foreseen for Faraday than for diagonal configuration.

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Decay heat removal capability under boiling condition was studied using an LMFBR fuel subassembly mockup loop. The sodium flow was driven by natural convection through the loop in which was installed a 37-pin bundle heated electrically over a length of 45 cm.
The heat flux furnished by the pins was increased stepwise, upon which the two-phase flow regime changed from bubble to slug flow and then to annular or annular mist flow. Dryout occurred even in slug flow regime, but only momentarily, and permanent dryout was not observed before establichment of annular flow. A suitable criterion for permanent dryout is considered to be 0.5 average exit sodium vapor quality. The results indicated that upon occurrence of sodium boiling, the coolability of fuel subassembly would be maintained by natural convection after reactor shutdown.

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Experimental results are presented on the flow behavior, phase distribution, average void fraction and slip ratio in air-water two-phase mixture flowing through an inverted U-bend. The curved test section of transparent acrylic resin tubing with 24 mm I. D. was used, comprising a riser, an inverted U-bend with radius of 96 mm, and a downcomer. From the visual observation, a flow map was derived, which indicates the relation between the flow patterns and velocities. Distributions of local void fraction along a diameter lying in the central plane of the bend and over the whole cross section of bend tube were measured, and those distribution mechanisms were explained by the effect of centrifugal and gravitational forces. The average void fraction around the bend was obtained by numerical integration of the measured local void fraction, and the slip ratio was calculated using the average void fraction.

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Void swelling in 10% cold-worked (10% CW) and 20% cold-worked (20% CW) type 316 stainless steels was investigated by 200 keV C⁺ ion irradiation and transmission electron microscope observation. Both 10% CW and 20% CW 316 steels show the swelling maximum at 923 K. Swelling in 10% CW 316 is much higher than that in 20% CW 316. The voids in the former material are larger and fewer than those in the latter material. The bilinear equation is applicable to describe swelling dose relation for both materials, except 10% CW 316 at higher doses than 50 dpa, where sharp swelling increase is observed. Heat to heat variability seems to exist in incubation dose, though it is not large. With regard to swelling rate, all three heats examined show good coincidence for both 10% CW and 20% CW 316 steels. Comparison of 20% CW 316 swelling rate for various irradiation projectiles indicates that the swelling rate is described as a simple function of the projectile mass, and there may exist a scaling law between the different projectile data.

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With a view to examining the failure-bearing capability of Zircaloy-4 cladding under postulated Loss-of-Coolant Accident condition in LWRs, integral tests of rod-burst, oxidation and thermal-shock were performed using simulated fuel containing Al₂O₃ pellets sheathed in Zircaloy-4 specimen cladding, filled with He gas, and sealed. This simulated fuel rod was oxidized in steam flowing at the isothermal oxidation temperatures between 920 and 1, 330°C for duration ranging of 3480 min after the cladding burst. After isothermal oxidation, the rod was quenched with bottom-flooding water under the condition of constraint or no constraint.
The failure boundary oxidation condition of the cladding on quenching under no constraint condition lay in the region of 3538% ECR for the isothermal oxidation temperatures between 1, 050 and 1, 330°C. For the temperatures ranging 9701, 050°C, the boundary value of ECR was somewhat lower than that obtained for higher temperatures.
The failure boundary oxidation condition of the cladding on quenching under constraint condition lay in the region of 19---24% ECR for the isothermal oxidation temperatures between 930 and 1, 310°C. It is sufficiently large compared with the criterion of 15% ECR in Japanese acceptance criteria for ECCS. Hydrogen absorbed by the Zircaloy-4 cladding as well as oxygen played a dominant role in the fracture behavior of the rod during flooding under constraint condition.

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The finite element method is applied in Galerkin-type approximation to three-dimensional neutron diffusion equations of fast reactors. A hexagonal element scheme is adopted for treating the hexagonal lattice which is typical for fast reactors. The validity of the scheme is verified by applying the scheme as well as alternative schemes to the neutron diffusion calculation of a gas-cooled fast reactor of actual scale. The computed results are compared with corresponding values obtained using the currently applied triangular-element and also with conventional finite difference schemes.
The hexagonal finite element scheme is found to yield a reasonable solution to the problem taken up here, with some merit in terms of saving in computing time, but the resulting multiplication factor differs by 1% and the flux by 9% compared with the triangular mesh finite difference scheme. The finite element method, even in triangular element scheme, would appear to incur error in inadmissible amount and which could not be easily eliminated by refining the nodes.

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