A multi-functional and small-scale ion accelerator has been developed for surface analysis. The maximum power of the accelerator is 200 kV×1 mA and an extremely low-energy beam is produced by using a deceleration power supply in place of an acceleration one.
A beam of 1 ns in minimum pulse width can be obtained by means of an electrostatic einzel lens and a fast electric deflector in order to improve the signal to noise ratio in beam measurement and also enable energy measurement using a TOF method.
In addition, the accelerator is equipped with a microbeam apparatus. It delivers a micro beam around 3 itim in mini mum diameter and can be used with a scanner fabricated from XY deflectors or a movable stage for surface analysis.
Also the accelerator can be controlled remotely through the Internet with safety connection. A low-energy proton PIXE surface analysis experiment on a corpuscular sample was successfully performed with this accelerator system.

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Radionuclides for PET are mainly produced using on-site small cyclotrons. During the production of the radionuclides significant quantities of neutrons are produced. Gold foil activation was used as a method to detect the neutrons at several points around a cyclotron. Thermal neutron fluxes in the target box were estimated as 9.3×10⁶cm^-2 s^-1 and 1.7×10⁶cm^-2s^-1 during ¹⁸F and ¹¹C production, respectively. Those in the cyclotron room were estimated as 4.1×10⁵cm^-2s^-1and 1.2×10⁵cm^-2s^-1, respectively. Those outside the room were estimated as being equal to or less than -3cm^-2s^-1, which corresponded to 0.1μSv h^-1 in effective dose. The effective dose outside the cyclotron room was proved to be safety level.

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For decommissioning of the medical linear accelerator facilities, thethermalneutron dose in three hospitals equipped with 10 MV medical linear accelerator was determined by activation of gold foil and CR-39. It was found that the maximum neutron fluence rate at the surface of inside wall was 7.6 ×10⁶±1.8×10⁶cm²d^-1 (mean±standard deviation). ¹⁹⁶Au was not detected . These results indicate that the quantity of the radio-activation nuclides generated in the concrete will not exceed the clearance level of IAEA RS-G-1.7.

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When the generated activity of 18F was 100 GBq, about 1015 neutrons are emitted by the nuclear reaction in target of the medical cyclotron. These neutrons induce activity in the cyclotron and the indoor concrete of the cyclotron room, and will contribute to the exposure of the staff maintaining the cyclotron.
This paper describes the basic characteristics of the thermal neutron measurement method of 23Na activation detector by auto radiography (ARG) using the medical imaging plate (IP).
Simple linear regression lines were able to describe the relationship between the scanner unit and the activity of ²³Na. The optimal S value and exposure time of ARG method was found to be 1, 000 and 24 hours. This method that uses the salt instead of gold foil allows hospitals to measure the thermal neutron fluencies easily at many locations for the radiation safety management of routine work and the decomissioning of the cyclotron facility.

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It is known that a radioisotope disperses in the air from a radioisotope labeled compound in an aqueous solution via the isotopic exchange reaction. In this research, in order to examine the dispersion mechanism of tritium in the air from a tritium labeled compound, the model room which imitated a working room in the controlled area was made, and the radioactivity of the tritium contained in the air of the model room was measured by sampling the air in the model room. It was found that the dispersion rate of tritium in the air increased with the passage time from its purchase. The dispersion rate of tritium from 3H-ATP changed from 0.10% to 0.76% after 17.7 months. Furthermore, the two-dimensional distribution of tritium on the surface of the whole walls in the model room was obtained using an imaging plate technique.

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All radiation facilities using unsealed radioisotopes are required to measure the concentration of radioactivity in indoor airborne every month on the basis of the“Industrial Safety and Health Law” and the“Ordinance on Prevention of Ionizing Radiation Hazards”. The indoor concentrations of radioactive substances were also calculated from the actual amounts of radioisotopes handled in Nagasaki University. Both the calculation data and the measurement data in practice were extremely well below the control level and even lower than the detection limits. These results suggest that the evidence-based efforts should be made to relax the present regulations on measurements of indoor airborne radioactivity. The possible scenario of exemption rules are discussed and proposed here.

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It is difficult to detect a low level contamination with ⁵¹Cr using a scintillation survey-meter because background counting rate of the survey-meter is high and emitted rate of λ-rays (0.320 MeV) from ⁵¹Cr is less than 10%. We examined whether a surface contamination with ⁵¹Cr can be detected directly with a GM survey-meter. As the result, the detection efficiency and detection limit of the GM survey-meter against the surface contamination were more than 0.6%(4π) and 4.6 13q/cm², respectively. From measurement of transmittance of radiations from ⁵¹Cr against poly (vinylidene chloride) seat, it became clear that the majority (about 98%) of detected radiations with the GM survey-meter is characteristic X-rays. These results show that the GM survey-meter can be used as a detector to check a surface contamination with ⁵¹Cr in controlled area.

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To look over the current measurement of radioactivity concentration in working environment of many radioisotopes facilities, a questionnaire survey was carried out under the auspices of the Planning Committee of the Japan Society of Radiation Safety Management. 64 responses were obtained in 128 radiation facilities, which the questionnaires were sent to. The main results were obtained by aggregate analysis of the answers for questionnaires as the followings.
Major nuclides subject to measurement were ³H, ¹⁴C, ³²P and ¹²⁵I Sampling of radioisotopes in air was mainly performed using collectors like dust samplers and HC-collectors. Liquid scintillation counters and gamma counters were used to measure β and γ radioactivity contained in airborne particles or gas samples. Contamination by radioactivity was not detected in 55% facilities surveyed, but in 40% facilities at the same level as or at lower levels than a hundredth part of the regulated concentration limit of each nuclide. Almost all facilities is found to consider that the measurement of radioactivity concentration in working environments is not always necessary.

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It has been popular to use the computer in almost all of radiation facilities. However the commercial software for radiation control is expensive and is not often suited for each facility.Especially, I have a particular problem that the two buildings of control area and the radiation control office are separated each other.Therefore, I developed the homemade software for radiation control by using commercial tools.I realized a favorite system characterized by the real-time acquisition of user input data from control area and automatic preparation of several legal reports for radiation control

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