A measurement of neutron dose rate on iron-polyethylene shielding structure was carried out by 252Cf source. Simulated geometry was slit-like opening of polyethylene in iron slab and polyethylene slab shielding. These experiment was done at research facility of Hazama Co,. Iron slab and polyethylene slab thickness were 10cm each. A gap of the polyethylene was simulated. Neutron REM-counter, polyethylene covered BF3 counter (STUDSVIK 2202-D), was used for measurement of streaming neutron dose equivalent. The solid state track detector (SSTD), allyl-diglycol-carbonate, were used for measurement of fast neutron dose equivalent in the range of l 70Kev to l 5Mev. The experimental data was obtained against gap width, source location and detector location. Obtained data shows strong correlation between dose rate and above parameters. These data was investigated in the view of to make use of actual facility design and compared with calculation such as MCNP48. From the result of gap streaming experiment and calculation, we obtained allowable gap width as 6mm for this case (10cm polyethylene thickness).
The neutron streaming for torus duct was investigated experimentally and analytically by using actual scale of experimental setups and 252 Cfneutron source, in order to obtain the basic data of the shielding design such as the nuclear fuel facilities. The torus ducts with 1 m of radius of curvature and 3cm and 5cm of the duct diameters were used. Neutron doses were measured by REM counter and solid state track detector (SSTD). The MCNP4A calculation was done for the comparison with the experimental results. In the case where the arrangement of the source and the detector is comparatively able to foresee through the duct, the doses were high as a matter of course. In the case where the other side could not be foreseen directly through the torus duct, the dose became 4E-2 as compared with the straight duct. The dose for 3 cm of the duct diameter was generally smaller than that of 5 cm. The doses for 3cm increased only 7% as compared with the bulk case (90 cm thickness, source position: disk, detector position: center).
Nuclear facilities with strong radioactivity need massive concrete shields. In the view of shielding design, there arise some difficulties in the estimation of radiation doses from a gap or a void between the hatch and the wall. So, there is a need of experimental data and calculational procedures on such geometry for a reasonable shielding design. The experiments were carried out by 152 Cfneutron source. A gap between the shielding hatch and the wall, and the offset of the shielding hatch were simulated by piling up the concrete slabs to the height of30 cm. The thickness of the slabs are I O cm each. Neutron dose was measured by REMcounter and CR-39 plate. To simulate various types of the shielding hatch, gap width and offset length of each gap were changed.Results of the experiments shows that the dose rate goes up with increment of gap width when the offset was fixed. The experimental data were also compared with the calculations by MCNP4B. We obtain the guide of design for radiation shielding hatch.
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