Michio Yoshizawa

NEUTRON SPECTROMETER WITH POSITION-SENSITIVE PROPORTIONAL COUNTERS

A spherical neutron counter for spectrometry and dosimetry has been developed. The counter consists of a spherical polyethylene moderator and three slender 3He position-sensitive proportional counters inserted into the moderator. The thermal neutron distribution in the moderator, which is a function of the energy of the incident neutrons is obtained with the proportional counters. Characteristics of the counter have been studied with experiments and Monte Carlo simulations. Also shown are energy spectra of a 252Cf neutron source that are obtained by an unfolding method using the thermal neutron distributions as response functions.

Development of a microcalorimeter for measuring absolute intensity of synchrotron radiation

Abstract A total absorption calorimeter was developed to measure the absolute intensity of monoenergetic X-ray beams from 10 keV to 70 keV of synchrotron radiation. Experiments with synchrotron radiation have demonstrated that the heat power above about 1 μW due to monoenergetic synchrotron radiation is measured with an accuracy of about 1%.

Determination of Gamma-ray Efficiency Curves for Volume Samples by the Combination of Monte Carlo Simulations and Point Source Calibration

In general, radioactivity measurements for volume samples are performed by the γ-ray spectrometry method with germanium detectors. In the method, peak efficiency curves must be obtained for each kind of sample in advance. For the reason, usually, many standard volume sources have been made which have different shapes, densities, matrix compositions and so on. This paper describes about a simple method for determination of γ-ray efficiency curves by the combination of Monte Carlo simulations and a single-point calibration using a standard point source. This method enables us to determine the efficiency curves for various types of samples and detectors easily and precisely.

A Monte Carlo Program for Estimating Characteristics of Neutron Calibration Fields Using a Pelletron Accelerator

A Monte Carlo code system, MCNP-ANT, has been developed to simulate the transport of charged particles (<4MeV) and the neutron production in an accelerator neutron target as well as the neutron transport. The code system consists of a newly developed Monte Carlo code for charged particle transport and neutron production, and an existing neutron transport code, MCNP-4B. The available reactions in the code system are 2H(d,n)3He, 3H(d,n)4He, 3H(p,n)3He, 7Li(p,n)7Be and 45Sc(p,n)45Ti as main neutron sources. The calculated neutron spectra by MCNP-ANT were compared with the experimental ones in typical monoenergetic neutron fields.

Performance of a spherical neutron counter for spectroscopy and dosimetry

A spherical neutron counter for spectroscopy and dosimetry has been developed. The counter consists of a spherical polyethylene moderator and three slender /sup 3/He position-sensitive proportional counters. Rough measurements of neutron fluence spectra and ambient dose equivalents, H*(10), have been done in the moderated-neutron field of JAERI. Characteristics of the counter are examined and compared with those of a Bonner sphere neutron spectrometer.

Measurement of the photon energy-absorption coefficient for air, nitrogen and argon at 30 keV

Abstract A parallel plate free-air ionization chamber was filled with air, nitrogen and argon, and the ionization current was measured for 30-keV monoenergetic incident photons, using a synchrotron radiation source. The photon intensity was measured using a total absorption calorimeter placed behind the chamber. Because the W value for each gas is accurately known, photon mass energy-absorption coefficients for each gas were measured. The measured coefficients are in agreement with published tables by Hubbell within 2%.

Performance of the Hp(10) and Hp(0.07) Measurable Electronic Pocket Dosimeter for Gamma- and Beta-Rays

For individual monitoring for external exposure of workers, electronic dosimeters have been widely used as supplementary dosimeters in nuclear facilities. In this study, an electronic pocket dosimeter for measuring the Hp(10) and Hp(0.07), which was recently developed by Aloka Co., Ltd., has been examined in terms of (1) energy dependences for gamma- and beta-rays and (2) applicability in actual radiation workplaces. For the latter purpose, performance tests were carried out in hot cells of the Reactor Fuel Examination Facility in the Japan Atomic Energy Agency. From the examinations, the followings were found; (1) the energy responses vary within ±20 % in the intended energy range for both radiations. (2) At the actual radiation workplaces, the readings of the dosimeter tend to show slightly higher Hp(10) and lower Hp(0.07) than those of other types of dosimeters, i.e. glass dosimeters and thermoluminescence dosimeters.

Evaluation of Counting Efficiency of a Whole-body Counter using the EGS4 Code

Counting efficiencies of a whole-body counter installed in the Japan Atomic Energy Research Institute were evaluated using Monte Carlo simulation and measurement to validate the calculation method of the counting efficiency curves of the whole-body counter for the water-filled block-shape phantoms containing radionuclides. Consequently, the counting efficiency curves in the photon energy range of 300–1,461 keV were obtained using the Monte Carlo simulation. The calculations are in fairly good agreement with the measurements in the photon energy range of 662–1,461 keV. It was also found that the self-scattering effect of photons in the phantoms accounts for about 7% of the counting efficiency for 662 keV photons.

Development of Neutron-monitor Detectors Applicable to Energies from Thermal to 100 MeV

A method for evaluating the neutron dose from several MeV to 100 MeV was developed using a BC501 A-type liquid scintillator. The neutron dose is evaluated from the light output of the BC501A by applying the G-function, which relates the light output of the BC501A to the ambient dose equivalent. On the basis of the G-function method, a hybrid neutron detector consisting of the BC501A coupled with a 6Li glass scintillator was developed to extend the measurable energy to the thermal range. The energy response of the detector was studied with 252Cf and 241 Am-Be neutron sources and mono-energetic neutrons from 8 keV to 65 MeV produced by accelerators. The ambient dose equivalents were evaluated using the energy response characteristics of the detector for several neutron fields in accelerator facilities. It was concluded from the analysis that the hybrid detector can be applied to accurate neutron monitoring from thermal energy up to 100 MeV