DEVELOPMENT OF A NEW DETECTOR SYSTEM TO EVALUATE POSITION AND ACTIVITY OF PLUTONIUM PARTICLES IN NASAL CAVITIES.

Plutonium dioxide (PuO2) is used to fabricate a mixed oxide fuel for fast breeder reactors. When a glove box containing PuO2 fails, such as by rupture of a glove or a vinyl bag, airborne contamination of plutonium (Pu) can occur. If a worker inhales PuO2 particles, they will be continually irradiating their lung tissue with alpha particles, and this could cause lung cancer. The nasal smear and nose blow methods are useful for checking workers for PuO2 intake in the field. However, neither method can evaluate the quantitative activity of Pu. No alpha-particle detector that can be used for direct measurements in the nasal cavity has been developed. For direct and quantitative measurement, it is required that a shape of the detector should be a fine bar which inserts itself in the nose to measure the accurate activity of Pu. Therefore, we developed a nasal monitor capable of directly measuring the activity of Pu in the nasal cavity to estimate the internal exposure dose of a worker. Prismatic-shaped 2 × 2 acrylic light guides were used to compose a detector block, and a ZnS(Ag) scintillator was adhered to the surface of these light guides. Silicon photomultiplier (SiPM) arrays with 8 × 8 channels were used as a photodetector. Actual PuO2 particles were measured using the nasal monitor. The nasal monitor could be directly inserted in the nasal cavities, and the activity distribution of Pu was obtained by the nasal monitor. The average efficiencies in 4-pi were 11.4 and 11.6% for the left and right nasal cavities, respectively. The influence of gamma and beta rays from Cesium-137 (137Cs) Strontium-90 (90Sr) on the detection of the alpha particles of Pu was negligible. The difference in the measured Pu activity between the ZnS(Ag) scintillation counter and the nasal monitor was within 4.0%. Therefore, it was considered that the developed nasal monitor could be used in direct Pu determination to estimate the internal exposure dose of workers.

A diamond 14 MeV neutron energy spectrometer with high energy resolution.

A self-standing single-crystal chemical vapor deposited diamond was obtained using lift-off method. It was fabricated into a radiation detector and response function measurements for 14 MeV neutrons were taken at the fusion neutronics source. 1.5% of high energy resolution was obtained by using the (12)C(n, α)(9)Be reaction at an angle of 100° with the deuteron beam line. The intrinsic energy resolution, excluding energy spreading caused by neutron scattering, slowing in the target and circuit noises was 0.79%, which was also the best resolution of the diamond detector ever reported.

Superheated emulsions as high-energy neutron dosemeters.

Superheated emulsions being inexpensive, easy to fabricate, and having tissue equivalent composition make them as one of the popular neutron dosemeters. One more advantage is that they can be made insensitive to gamma rays by the choice of the sensitive liquid. It is observed that the response of commercially available bubble detector to neutron decreases above 20 MeV while its response is roughly flat in the 0.1-15 MeV region. This restricts its application as a dosemeter to high-energy neutrons. The response of bubble detector from Bubble Technology Industries, has been observed by using Pb-breeder for high-energy neutrons from different facilities in Japan. It is observed that 2-3 cm Pb-breeder is effective in increasing the response of the detector to the nominal value. Theoretical calculation using MCNPX code indicates an increase in neutrons in the energy range of 0.1-10 MeV with Pb-breeder. The present work indicates the possibility of using the bubble detector as a dosemeter to high-energy neutron using a Pb-breeder of proper thickness.

Propagation characteristics of neutrons leaking from the accelerator facilities.

In this study spatial and time distribution of neutrons leaking from Hokkaido University 45 MeV Electron Linac facility have measured and compared with the Monte Carlo simulations. The neutron transport processes inside and outside the facility building has been simulated using MCNP. The neutrons have measured by BF3 counters and 3He counters with polyethylene moderators up to the distance of 330 m from the facility. The spatial distribution of ambient dose equivalent converted from the counts has been compared with the simulation. The distribution estimated from the counts by the BF3 counter has been shown fairly good agreement with the calculation. The spatial distribution of counts obtained at the 45 MeV Electron Linac facility has been compared with that obtained at the Fusion Neutronics Source (FNS) facility of JAERI. The difference between the propagation characteristics of neutrons leaking from those facilities has been discussed.