An efficient albedo Monte Carlo method newly developed has been studied by analyzing two types of experiments on neutron streaming. The method is characterized by employing the energy-angle dependent doubly differential albedos for slab, which can be calculated in a short computer time with a one-dimensional transport theory, such as the Sn method and more efficient invariant imbedding method. This paper describes the features of the present albedo Monte Carlo method, including fundamental formulas. In the analyses of the neutron streaming experiments, the calculated results agreed with the measured data within a factor of 2 for a benchmark experiment at the YAYOI reactor and within a factor of 3 for an SNR sodium duct mock-up experiment.
It is concluded that the present albedo Monte Carlo method is practical and applicable to the reactor shielding analysis concerning radiation streaming.

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A simple formula which describes multi-scattered neutron flux in a spherical cavity was derived based on the albedo concept. The formura treats a neutron source which has an arbitrary energy-angle distribution and is placed at any point in the cavity.
The derived formula was applied to the estimation of neutron fluxes in two cavities, i.e. a spherical concrete cell with a 14-MeV neutron source at the center and the "YAYOI" reactor cavity with a pencil beam of reactor neutrons. The results of the analytical formula agreed very well with the reference data in the both problems. It was concluded that the formula is applicable to estimate the neutron fluxes in a spherical cell except for special cases that tangential source neutrons are incident to the cavity wall.

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The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage 1-1 Test.
The calculated results are in good agreement with the experimental results. The analysis indicates that fuel cladding temperature in a portion of the lowest one third of the test bundle would have reached the melting point of the ZrO₂ during a rapid temperature excursion driven by the zirconium-water reaction. The result is consistent with the result of metallographic examination. The crucibilization effect of the ZrO₂ layer played an important role in the reaction. Steam starvation condition would have occurred in contrast to the situation of the Scoping Test of the same test series. Zirconium-water reaction on the inner surface of the fuel cladding was found to have made a strong effect on the fuel rod temperature in the upper part of the test bundle.

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Characteristics of acceleration frequency spectra for an impact signal were studied and loose part mass was estimated, using variation in the frequency spectral patterns. The frequency spectral pattern for the acceleration induced by impact depends on the loose part mass and not on the impact energy and the loose part shape. It was found that the frequency spectral pattern for an impact which occurred at the lower plenum vessel wall is not influenced by the impact point. On the other hand, the maximum acceleration am-plitude depends not only on the loose part mass but also on the impact energy, the loose part shape and the distance from the impact point. It was also found that the foremost portion of the accelerometer output signal is influenced by the impact force between a loose part and the vessel. However, a second portion is influeuced by the sensor resonance and structural resonances. Loose part mass estimation, using "FR" values (ratio between low frequency component and high frequency component), is useful. The uncertainty range for the estimated loose part mass is about 0.7 decade.

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An experimental study has been conducted to investigate the effect of freezing of liquid on the heat transfer characteristics for laminar flow between two cooled parallel plates of 20 mm in distance. Both plates were maintained at the same uniform temperature which was lower than the freezing temperature of the working fluid, water. The plate wall temperature ranged -2-7°C and the inlet water temperature was varied from 1 to 8°C. The corresponding dimensionless wall temperature θw ranged 0.24.5. The Reynolds number was varied from 700 to 2, 300. The frozen layer was smooth and had a mono-tonously increasing thickness as water flowed downstream. In thin layer (θ_w ?? 1.1), the experimental results agreed well with the analytical calculations. In thick layer (θ_w>1.1), however, the measured layers were thinner than the calculated values since the fluid acceleration due to flow area reduction might produce an enhancement in heat transfer, which was neglected in the theory.

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Neptunium in high level radioactive wastes has to be retained in glasses through geological period from the point of biological toxicity for more than million years. Neptunium-237 diffusion in borosilicate glass with simulated wastes of 26.4w/0 was investigated in the temperature range of 400600°C by the use of α-degradation method. The energy loss rate dE/dx of α-particles, which is necessary in order to determine diffusion coefficients by the α-degradation method, was calculated for the waste glass.
The penetration depth of a-particle with 4.787 MeV from ²³⁷Np was 17 μm, which gives a limit in applying the α-degradation method in the waste glass. The temperature dependence of the diffusion coefficient of Np in the waste glass was given by
D_Np=3.67 exp(-55, 900/RT) (cm²/s), in which the activation energy of the diffusion was 55.9 kcal/mol. It was clarified that Np is one of the elements with the lowest mobility in waste glasses.

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An automatic measuring instrument for the radon daughters concentration of precipitation has been constructed and used successfully. The precipitation collected by a polyethylene funnel during 15 min is led to a glass cell set in a well-type NaI (T1) scintillator, and after 5 min counting the water sample is pumped out. A sequence of operations of sampling, measurement, recording, draining out and correction for the volume of the water sample is performed automatically, and is repeated at every 15 min. The instrument has sensitivity and durability enough for practical use at outdoors. This instrument enables the measurement of radon daughters concentration as low as 0.5 Bq/ml at precipitation rate as small as 0.5 mm/h.
For the improvement of accuracy of the monitoring system of environmental γ-radiation, the effect of radon daughters accumulated on the ground surface by precipitation can be eliminated on real time by means of providing it with the output of this instrument.
This instrument can be used for the measurement not only of the radon daughters but also of the other radioactive nuclides in precipitation.

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The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25% main recirculation pump suction line break (MRPS-B) experiments, the 21% single-ended jet pump drive line break (JPD-B) experiment and the 15% main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests.
In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop.

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