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We developed a thermal neutron source for performance tests of boron agents for Boron Neutron Capture Therapy (BNCT) at Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) in National Institute of Radiological Sciences, Japan. The thermal neutron flux at the experimental position was measured to be 2.18× 108 n/cm2/s, the cadmium ratio was 9.21, and gamma ray dose rate was 2.11 Gy/h. The neutron dose rate was estimated to be 2.96 Gy/h. These values indicate that the neutron source can be applicable for tests of boron agents although application range is limited. An example of irradiation experiment is described.

Dopant free eutectic scintillators 6LiF–SrF2 and 6LiF–CaF2 were developed by the vertical Bridgeman method for the purpose of thermal neutron detection. The molar ratio of LiF and Ca/SrF2 was 4:1 on its eutectic composition. The α-ray induced radioluminescence spectra of the scintillators showed intense emission peak at 300 nm due to the emission from the self-trapped exciton in Ca/SrF2 layers. When the samples were irradiated with 252Cf neutrons, 6LiF–SrF2 and 6LiF–CaF2 exhibited the light yields of 4700 and 9400 ph/n, respectively. Scintillation decay times of 6LiF–SrF2 and 6LiF–CaF2 were accepted for scintillation detectors, 90 and 250 ns, respectively.

We have developed a novel micro-size dosimeter using an optical fiber probe coupled with a photostimulable storage phosphor, such as BaFBr:Eu 2+ . Our micro-size dosimeter is based on the concept in which the radiation dose response is temporally compressed by using integrating-type dosimeter elements, such as a photostimulable storage phosphor, and the signal-to-noise ratio is improved to downsize the dosimeter probe. We fabricate the prototype dosimeter probe which has the detectable volume of 100 μm thick and 400 μm diameter and the outer diameter of 900 μm. The lower limit of detectable dose, which is defined as three times the noise level, is evaluated to be 0.9 mGy. This value is much smaller than the typical irradiation dose (a few Gy per treatment) in the radiotherapy. We, therefore, conclude that the proposed micro-size dosimeter can be a promising candidate of in-vivo and on-line dose monitor in the radiotherapy.

Capability of thermal neutron detection was examined for LiCaAlF 6 (LiCAF) scintillators doped with 3d-transition metal ions. Their radioluminescence spectra were measured with an 241-Am source to simulate 6 Li(n, α) 3 H reaction. The sufficiently intense radioluminescence was observed for the Mn, Co and Cu dopants, while only a weak one was observed for Ti, V, Fe and Ni. A Mn doped LiCAF crystal, which showed the highest radioluminescence intensity, was coupled with a Si avalanche photodiode for the examination of its neutron response. It was confirmed that the average current of the photodiode clearly increased under excitation with 13.5 meV neutron flux.

Ce 0.1% doped LiF–SrF2 eutectic scintillators were produced by vertical Bridgman method at various solidification rates of 1, 5, 20, 80, 320, and 1280 mm/hr. The LiF contained 95% of 6Li. The SEM images of the samples solidified at 1–80 mm/hr demonstrated clear lamellar structures. The α-ray induced radioluminescence spectra of the scintillators had intense emission peak at 310 and 330 nm due to the emission from Ce3+ 5d–4f transition of Ce:SrF2 layers. When the samples were irradiated with 252Cf neutrons, they exhibited almost the same light yields of 4500–5500 ph/n and typical decay times of 120–160 ns. The optimal layer thickness of LiF–SrF2 was determined to be 0.9 μm.

LiCaAlF6 scintillators are one of the attractive scintillators for neutron detection. To reduce the effect on gamma-rays, the Ce doped LiCaAlF6 scintillators can discriminate the neutron and gamma-ray events based on the pulse shape discrimination technique. To apply the scintillators for the oil logging, the high temperature characteristics must be investigated. In this paper, the temperature dependence of the neutron-gamma discrimination based on pulse shape discrimination technique in the Ce:LiCaAlF6 scintillator is investigated as one of the high temperature characteristics. The property of pulse shape discrimination in the Ce:LiCaAlF6 has small temperature dependence ranging from 25°C to 150°C. We concluded that the Ce:LiCaAlF6 scintillators can discriminate neutron and gamma-ray events under high temperature condition up to 150°C .

Eu2+ 0.05%, 0.1%, and 0.2% activated LiF–SrF2 eutectic scintillators were prepared by the Bridgman method using 6Li enriched (95%) raw material. The α-ray-induced radio luminescence spectra showed intense emission peak at 430 nm due to an emission from Eu2+ 5d–4f transition in the Eu:SrF2 layers. When excited by 252Cf neutrons, all the samples exhibited almost the same light yields of 5000–7000 ph/n with a typical decay times of several hundreds ns.

We propose a novel micro-size dosimeter using an optical fiber probe based on the PSL from the BaFBr:Eu 2+ . Through basic experiments using the prototype dosimeter, the feasibility of the proposed dose monitoring method has been confirmed. We confirm that increasing the stimulus laser power is effective to improve the signal to noise ratio. The lower limit of detectable dose of the prototype dosimeter probe having sensitive region with the diameter of 200 μm is preliminarily estimated to be 11 mGy for a Co-60 gamma-rays. This value is much smaller than the typical irradiation dose (a few Gy per treatment) in the radiotherapy. We can measure the dose in a few hundred subdivisions during a treatment. We, therefore, conclude that the proposed micro-size dosimeter can be a promising candidate of in-vivo and on-line dose monitor in the radiotherapy.

Two topics were focused. The first one was about the gamma-ray scintillator, Pr 3+ :Lu 3 Al 5 O 12 (LuAG). The second one was about neutron scintillator, Ce 3+ : 6 LiCaAlF 6 and Eu 2+ : 6 LiCaAlF 6 ( 6 LiCAF). Those scintillators have been developed very recently for modern imaging applications in the medical and homeland security fields. In both cases, the rare earth ions are playing the crucial role as emission centers. Pr 3+ in LuAG provided fast 5d→4f transition providing noticeably shorter decay time than that of Ce 3+ . Among several candidate hosts, LuAG showed the best performance. Bulk crystal growth, basic scintillation properties, two-dimensional gamma-ray imaging and positron emission mammography (PEM) application were demonstrated. Due to the international situation, the homeland security was compromized by illicit traffic of explosives, drugs, nuclear materials, etc. and the ways to its improvement became an important R&D topic. For this purpose the Ce and Eu doped LiCAF appeared competitive candidates. Especially, when substitution of 3 He neutron detectors was considered, the discrimination ability of gamma-ray from alpha-ray was important. Bulk crystal growth, basic scintillation properties and two-dimensional neutron imaging were demonstrated.

We demonstrate neutron–gamma discrimination based on a pulse shape discrimination method in a Ce:LiCAF scintillator. We have tried neutron–gamma discrimination using a difference in the pulse shape or the decay time of the scintillation light pulse. The decay time is converted into the rise time through an integrating circuit. A 252Cf enclosed in a polyethylene container is used as the source of thermal neutrons and prompt gamma-rays. Obvious separation of neutron and gamma-ray events is achieved using the information of the rise time of the scintillation light pulse. In the separated neutron spectrum, the gamma-ray events are effectively suppressed with little loss of neutron events. The pulse shape discrimination is confirmed to be useful to detect neutrons with the Ce:LiCAF scintillator under an intense high-energy gamma-ray condition.

We propose the novel gamma-camera with a functional collimator, that has detection ability on the collimator wall surface. The gamma-ray detection sensitivity of the proposed gamma-camera is improved by a factor of 70 compared with a conventional one. Through Monte Carlo simulations, the required gamma-ray emissions for image reconstruction is confirmed to dramatically to be reduced by using the information collected on the collimator wall surface. The spatial resolution is estimated to be 10 mm. In addition, we experimentally demonstrate image reconstruction procedure including reading out the collimator wall surface information. We successfully acquire the reconstructed images of a point-like Am-241 gamma-ray source from the projection image of the prototype gamma-camera with the functional collimator. We conclude that a much larger contribution from the gamma-ray emitters than background results in reduction on the lattice-shaped artifact.

We propose a novel electron density imaging technique based on a Compton scattered X-ray CT (CSX-CT) technique. We design fundamental configuration of the CSX-CT system, which consists of a fan-shaped X-ray beam, two-dimensional sensors for scattered X-ray detection, parallel plate collimators for limitation of the direction of scattered X-rays and a line sensor for the transmission of X-ray CT (TX-CT). An image obtained by the TX-CT can be used to correct the attenuation effect of scattered X-rays. Through Monte Carlo simulation modeling studies of the CSX-CT system, we demonstrate that the utilization of the information of scattered X-rays is useful to obtain the electron density image. We additionally confirm that the medical exposure irradiated in the CSX-CT is estimated to be lower than the maximum dose recommended in the guideline of some committees.

Eu 2+ 0.1, 0.5, 1, and 2 mol% doped LiCaAlF 6 single crystalline scintillators were grown by the micro-pulling down (μ-PD) method. Eu 2+ 2 mol% doped LiCaAlF 6 was also prepared using the Czochralski method. In the transmittance spectra, 4f–5d absorption lines appeared around 200–220 and 290–350 nm. An intense emission at 375 nm due to Eu 2+ 5d–4f transition was observed under 241 Am α-ray excitation. When 252 Cf excited pulse height spectra were measured, Eu 2% doped one showed the highest light yield of 29,000 ph/n with 1.15 μs decay time. Using the 2 inch φ Czochralski grown one coupled with the position sensitive photomultiplier tube covered by Cd mask with various size (1, 2, 3, and 5 mm) pin holes, thermal neutron imaging was examined. As a result, the spatial resolution turned out to be better than 1 mm.

A small-sized thermal neutron detector based on a (6)Li-glass scintillator and a plastic optical fibre has been developed for precise measurement of the spatial distribution of thermal neutron fluence rate. This detector was tested in experiments performed using thermal neutrons. The detector is useful in a thermal neutron field. However, the gamma-ray discrimination ability of the (6)Li-glass scintillator is not good in a low-intensity thermal neutron field. A new idea using a gamma-ray suppression method is proposed to reduce the uncertainty in the neutron counts due to the gamma-ray background. A novel small-sized thermal neutron detector consists of a (6)Li-glass scintillator, a hollow CsI(Tl) scintillator and plastic optical fibres. The evaluation of the gamma-ray suppression ability of the detector using the EGS4 code indicates that the gamma-ray suppression is effective.

The thermal neutron fluence rate is determined by the gold activation method. The absolute activity of the irradiated gold foil is measured by a 4πβ-γ coincidence counter. Using this method, corrections for the detection of conversion electrons and gamma rays by a 4πβ counter are very important to obtain accurate absolute activity. In this work, Monte Carlo simulations were performed to derive the correction factor K. The absolute measurement of (198)Au activity for 20-100 μm thickness Au foils were performed to verify the calculating model of the 4πβ-γ coincidence counting system.

NMIJ provides national standards on neutron fluence as well as neutron emission rate in Japan. The neutron fluence standards are established for thermal neutrons, monoenergetic neutrons at seven energies from 24 keV to 14.8 MeV, radionuclide-source neutrons from 252 Cf and 241 Am-Be. The neutron emission standards are provided for these radionuclide sources. These are basically prepared in accordance with ISO 8529-1. Worldwide consistency of the standards has been supported by participating in key comparisons organized by BIPM. NMIJ issues calibration certificates accepted worldwide within the range specified in our calibration and measurement capabilities listed in the BIPM web page. The development of the 19 MeV neutron field is in progress to complete standardization in 2011. The radionuclide-source neutron fluence standards were extended in 2008 so that calibrations are also available in terms of ambient/personal dose equivalent. The Japan calibration service system has launched to improve the traceability of neutron measurement in Japan. A new thermal neutron calibration field is developing with a guided beam at the research reactor JRR-3M of JAEA. Aiming at the establishment of the 20-100 MeV high-energy neutron standard, comparison studies are currently being conducted between the cyclotron facilities TIARA of JAEA and CYRIC of the Tohoku University. -1 ) in Japan. We report the present status and recent activities of these neutron standards.

Long counters are widely used as neutron fluence monitors by utilizing their flat-response feature. A bulky neutron moderator is required however to achieve the flat-response feature, which can be a disadvantage from the viewpoint of portability as well as an undesirable source of scattered neutrons. Just downsizing the moderator deteriorates the flat response especially for high-energy neutrons. We proposed a new method to downsize the moderator with keeping the flat-response feature and developed a compact flat-response neutron detector in the present study. The detector consists of two small spherical 3He proportional counters embedded in appropriate positions in a compact cylindrical polyethylene moderator. The sensitivity ratio between the two counters was adjusted to provide a flat response in the total outputs from the two counters. It was demonstrated by MCNPX calculations and response measurements for Cf and Am-Be neutrons and monoenergetic neutrons that the present detector had a good flat response in the wide energy range up to 20 MeV while being downsized by one fifth compared with long counters. The positions of the effective center were determined as a function of the neutron energy. We also proposed a method to evaluate the neutron energy using the detector, which will allow various new applications by taking advantage of measuring the neutron fluence and energy at the same time.

Neutron source emission rates are determined by absolute or relative measurement. The relative measurement is performed by comparing between a sample source and a calibrated reference source of the same type. In this calibration, the energy spectra of the sample and reference sources should be identical, since normally there exists energy dependence in the response of neutron detectors. We propose a method of relative calibration of the neutron source emission rate between different source types and spectra. In the present method, a sample source or a calibrated reference source is located at the center of a graphite pile and thermal neutron flux at a special position is monitored and compared. We performed gold foil activation measurements at the NMIJ/AIST graphite pile and revealed that the thermal neutron flux at that position per unit source emission rate is independent of the types of neutron sources. The details of the physical mechanism were examined by calculating the spectral fluence of the neutrons thermalized inside the graphite pile using the MCNP4C code. We demonstrated this method using Am-Be, Ra(/spl alpha/)-Be, Cf, Pu-Be neutron sources and obtained the successful results which agreed within 1%.

In this paper we propose a dynamic 3-D gamma camera consisting of multi-pinhole collimators and flat panel detectors. Through simulation studies on various pinhole configurations we determined that from a spatial and time resolution consideration, pinhole collimators with a 60° opening angle achieved acceptable dynamic 3-D imaging results. With regard to basic system performance, we estimated that the spatial and time resolution exceeds 3.5 mm and 4 s, respectively, for a small animal injected with 99mTc with an activity of 37MBq. In addition, from further experiments we concluded that the proposed system would be acceptable and practical as a dynamic 3-D imager for nuclear medicine applications although position calibration concerns require further study.

Our laboratory provides national standards on fast neutron fluence. Neutron fields are always accompanied by gamma rays produced in neutron sources and surroundings. We have characterised these gamma rays in the 5.0 MeV standard neutron field. Gamma ray measurement was performed using an NE213 liquid scintillator. Pulse shape discrimination was incorporated to separate the events induced by gamma rays from those by neutrons. The measured gamma ray spectra were unfolded with the HEPRO program package to obtain the spectral fluences using the response matrix prepared with the EGS4 code. Corrections were made for the gamma rays produced by neutrons in the detector assembly using the MCNP4C code. The effective dose equivalents were estimated to be of the order of 25 microSv at the neutron fluence of 10(7) neutrons cm(-2).

A new fast neutron spectrometer has been developed. The spectrometer is composed of a silicon surface barrier detector and three position-sensitive proportional counters with methane gas working as counting gas and a radiator. A collimated incident neutron interacts with a hydrogen atom in the methane gas to generate a recoil proton. The position information on the path of the recoil proton obtained from the three position-sensitive proportional counters gives the recoil angle. In the meanwhile, the energy of the recoil protons is measured with the three proportional counters and the silicon surface barrier detector. The characteristics of the spectrometer were evaluated with a monoenergetic neutron beam. The best energy resolution was 1.8% for 5.0 MeV neutrons.

We have developed a novel multi-pinhole-type gamma camera for imaging low-energy gamma-ray emitters. The gamma camera consists of Imaging Plates (IPs) and multi-pinhole collimators made of a lead plate with many pinholes in lattice arrangement. The three-dimensional spatial distribution of a gamma-ray source is reconstructed by using a stereoscopic back-projection technique, where the directions of gamma-ray incidence are determined with the digitized two-dimensional images on the IPs and the configuration of the multi-pinhole collimators. The horizontal and vertical spatial resolutions for a point-like 57Co source evaluated with simulations are 1.5 and 2.5 mm, respectively.

A thermal neutron reference field using a graphite pile of 190cm×190cm×250cm was developed at the High Energy Accelerator Research Organization. Inside the pile, the thermal neutron flux was measured by the gold foil activation technique using the Cadmium difference method, and outside the pile by the 3He proportional counter using the Cadmium difference method. Besides the gold foil, the optical fiber coupled to the ZnS(Ag) scintillator with the Li converter was used to measure the neutron flux inside the pile to verify its availability. The absolute values of thermal neutron fluxes measured inside the pile could be obtained with an accuracy of 7.5-7.8%. The measured thermal neutron fluxes inside the pile were compared with those calculated by use of the MCNP-4B Monte Carlo code. The calculated thermal neutron fluxes agreed very well with the measured ones. Outside the pile, the measured thermal neutron flux was 7.38×10-6 (n/[s·cm2·source neutron]), and the contribution of neutrons other than thermal neutrons was less than 1%. The thermal neutron fluxes measured by the optical fiber with the scintillator showed very good agreement with those measured with the gold foil activation method. The optical fiber detector was found to be quite useful to measure the thermal neutron flux distribution in the pile because of its much higher detection efficiency than that of gold foil activation.

A fast time response, wide dynamic range neutron flux monitor has been developed toward the LHD deuterium operation by using leading-edge signal processing technologies providing maxi- mum counting rate up to ?5 × 109 counts/s. Because a maximum total neutron emission rate over 1 × 1016 n/s is predicted in neutral beam-heated LHD plasmas, fast response and wide dy- namic range capabilities of the system are essential. Preliminary tests have demonstrated success- ful performance as a wide dynamic range monitor along the design.

Neutron monitoring is important to manage safety of fusion experiment facilities because neutrons are generated in fusion reactions. Monte Carlo simulations play an important role in evaluating the influence of neutron scattering from various structures and correcting differences between deuterium plasma experiments and in situ calibration experiments. We evaluated these influences based on differences between the both experiments at Large Helical Device using Monte Carlo simulation code MCNP5. A difference between the both experiments in absolute detection efficiency of the fission chamber between O-ports is estimated to be the biggest of all monitors. We additionally evaluated correction coefficients for some neutron monitors.

The purpose of this study is to develop a radiation distribution monitor using a normal plastic optical fiber. The monitor has a long operating length and can obtain continuous radiation distributions. A principle of;the position sensing is based on a time-of-flight technique. The monitor is sensitive to beta rays or charged particles, gamma rays, and fast neutrons. The spatial resolutions for beta rays (/sup 90/Sr-/sup 90/Y), gamma rays (/sup 137/Cs) and D-T neutrons are 30 cm, 37 cm and 13 cm, respectively. The detection efficiencies for the beta rays, the gamma rays and D-T neutrons are 0.11%, 1.6/spl times/10/sup -5/% and 1.2/spl times/10/sup -4/%, respectively. The effective attenuation length of the detection efficiency is 18 m. In this paper, we describe the basic characteristics of this monitor.

Rapid detection of fuel failures as well as failed fuel location are quite important for safety in fast reactor operation. As one of the failed fuel detection and location (FFDL) techniques, a tagging method has been adopted, where a tag gas with different isotopic abundance of krypton (Kr) and xenon (Xe) is loaded into each fuel pin, and analyzed when leaked into cover gas with fuel failure. On the other hand, with recent improvement of laser performance, resonance ionization mass spectrometry (RIMS), which can detect specific atoms with high sensitivity and elemental selectivity, has been practicable in engineering application. We propose a new FFDL technique for fast reactor safety instrumentation based on monitoring of a very small amount of the Kr/Xe tag gas leaked into the cover gas by RIMS. Here are discussed the basic performance necessary to the design of RIMS-FFDL such as the detection limit and the elemental selectivity of Kr and Xe through basic experiments and numerical analyses. The applicab...

The application of a ‘digital-analog hybrid signal processing system’ combined with a Euclidian distance method is presented. The proposed method is used to improve gamma-ray energy spectrum characteristics of a CdZnTe semiconductor detector. A digital and an analog system are simultaneously applied to the detector output after passing through a preamplifier, where the digital system digitizes the signal pulse shape for the Euclidian distance determination, while the analog system measures the signal pulse height with a conventional analog method. In this paper, the digital and the analog data are combined to improve the energy spectrum characteristics; giving the increase of the photopeak area, the peak-to-valley ratio, and the photopeak sharpness.

We present a study on pulse shapes of a CdZnTe semiconductor detector for beta particles and gamma-rays emitted by a 137Cs-90Sr mixed source. When beta and gamma radiations enter a semiconductor detector, two kinds of pulses are produced according to the type of radiations. The characteristics of the pulses are determined by their energy and depth of interaction. The energy deposited on the detector basically determines the pulse height and the depth of interaction affects the pulse shape. Because of the different characteristics of beta and gamma radiations, their pulse shapes are also different. The digital-analog hybrid signal processing method was applied to separation of the beta spectrum from the beta-gamma mixed one. The separation of beta spectrum was done based on a neural network algorithm, a leading edge area method, and a falling continuum tail value method. However, the perfect separation could not be obtained. The separation was also done by reconstructing the spectra on the basis of subtrac...

We present the application of a “fusion method” combined with a “similarity” method to improve γ-ray energy spectrum characteristics of a CdZnTe semiconductor detector. A digital and an analog system are simultaneously applied to the detector output after passing through a preamplifier, where the digital system digitizes the signal pulse shape for the similarity determination, while the analog measures the signal pulse heights with a conventional analog method. In this paper, the digital and the analog data are fused to improve the energy spectrum characteristics, giving an increase of the photopeak area, the peak-to-valley ratio, and the photopeak sharpness.

A new concept of compact neutron camera with directional neutron detector arrays is proposed for nuclear fusion experiment. The basic performance of the directional neutron detector is made clear through experiment for a prototype detector with 14MeV neutrons and design consideration of the detector specification to optimize its directionality. The results show that the angle resolution defined as a full width of neutron incident angles at a half maximum of detector counts would reach up to around 4 degrees at minimum around 45 degrees neutron incidence to the detector axis. The concept for a compact neutron camera is also given by making good use of an array of the optimized directional neutron detectors.

We have developed a thermal neutron calibration method using a reactor produced neutron beam in JRR-3M of the Japan Atomic Energy Agency. Neutron-induced prompt gamma ray analysis has usually been performed in this beam line. Neutron energy distributions with negligible contributions from epithermal neutrons were measured by a time-of-flight method with a chopper made of 6 LiF powder. The thermal neutron flux was determined by a gold foil activation method. We found that the thermal neutron beam in JRR-3M was well suited for calibration, neutron detector development or neutron dosimetry.

We present a new application of a `Fusion Method’ to improve gamma-ray energy spectrum characteristics of a CdZnTe semiconductor detector. A digital and an analog system are simultaneously applied for the detector output; The digital system digitizes the signal pulse shapes for pattern recognition based on a neural network algorithm. The analog measures the signal pulse heights with a conventional analog system. The digital and the analog output data are fused and then used for improvement of the energy spectrum characteristics, i.e. the photopeak area, the peak-to-valley ratio, and the FWHM of photopeaks.

The spatial dependence of thermal and fast neutron flux was measured axially in the core of a 1 MW research reactor. The measurements were made by a thin optical fibre detector with a neutron sensitive ZnS(Ag) scintillation tip. For thermal neutrons 6 Li was used, whereas for fast neutrons 232 Th was used as neutron converter. The spatial dependence was measured by moving the fibre axially with a uniform speed. The measurement takes a few minutes, compared to up to 10 h with the conventional wire activation method. Comparison with traditional measurements shows a good agreement.

On the tip of an optical fiber, a very small amount of ZnS(Ag) scintillator mixed with neutron converter and adhesive paste was painted, which can be inserted into a narrow space and moved by a fiber-driving unit controlled with a personal computer. The presented method made it possible to measure neutron flux distributions in critical assemblies, research reactors or other neutron facilities with position resolution of less than 1 mm and in 10 min or so over a meter. As neutron converter, 6Li and 235U are good for thermal neutrons and 232Th is useful for fast neutrons. When a neutron converter is not used, the method can be applied to the measurement of γ-ray intensity distribution.

Time-spatial neutron measurement was carried out by using a position-sensitive 3He proportional counter (PSPC) in the Kyoto University Critical Assembly. The data of the output pulses from the PSPC were acquired by a data acquisition system which was based on the LN-9000 system of Laboratory Equipment Corporation. In this system, the time and the position data of the neutron detection pulses were simultaneously accumulated by a personal computer. In the present study, the Rossi-α method, which was derived for time-domain neutron correlation analysis, was extended in order to process the measured two-dimensional data of time and space. The processed experimental results and a theoretical formula based on the one-group diffusion approximation neglecting the higher modes agreed well with each other, which indicates the present system is applicable for reactor physics experiments including reactor kinetics.

A novel neutron detector has been developed to measure energy spectra of space neutrons in a spaceship. The detector consists of a spherical polyethylene moderator with a diameter of 30 cm and 19 pairs of 6 Li- and 7 Li-glass scintillators arranged along the orthogonal x -, y -, and z -axes of the sphere. The pairs can distinctively detect thermal neutrons in some radiation fields, where other radiation exists, by subtracting the number of detected events of the 7 Li-glass scintillator from that of the 6 Li-glass one. Hence the detector can clearly obtain the thermal neutron flux distribution in the spherical moderator even in such fields. The detector was applied to the measurement of monoenergetic and 252 Cf neutrons. The characteristics of the detector were studied by comparison with the simulations.

A new type of fast neutron spectrometer composed of a CH 4 proportional counter and a Si semiconductor detector which is called the counter telescope type was constructed for plasma-temperature diagnostics. The counting gas CH 4 is used as a converter of neutrons to protons. Since the energy loss of the proton in the converter CH 4 can be fairly precisely measured with the proportional counter, the energy resolution for the proton is high, so that a good energy resolution for neutrons is obtained. The energy resolution of the spectrometer itself excluding the energy spread of the D–T 14 MeV neutrons was 1.7% which seems to be the best so far, and the detection efficiency was 1×10 −6 . If better energy resolution is preferable, even 0.5% is obtained and the efficiency in this case is 4×10 −8 .

A novel pulse-shape analysis method that uses similarity to recognize an individual pulse shape is presented in this paper. We obtain four pulse heights by using four linear amplifiers with different shaping time constants. We treat a combination of the four pulse heights as a pattern vector. A similarity of the pulse shape can be obtained by comparison between the pattern vector and a discriminant vector which was given in advance. Each pulse shape is analyzed by using the similarity. The method has been applied for the improvement of characteristics of a CdZnTe semiconductor detector. The characteristics of the energy spectrum of the CdZnTe detector such as the photopeak efficiency or the peak-to-valley ratio are improved after the correction procedure with the similarity.

In the absolute radioactivity determination of gaseous samples using the internal gas counting method with proportional counters, evaluation of counting losses of pulses whose pulse heights are lower than the discrimination level is very important to obtain high accuracy. Counting losses due to wall effect and to very low energy of a β-ray were evaluated by experiment and EGS4 Monte Carlo calculation. The calculated results agreed well with experimental results when parameters in the EGS4 code were properly chosen; which shows the calculation code is able to be applied to other radioactive nuclides and other measurement conditions for the evaluation of counting losses.

To obtain a high precision in radiation or radioactivity measurements with an imaging plate, the photo-stimulated luminescence intensity must be large. To get the highest intensity, we reread an imaging plate repeatedly and summed each intensity for the purpose of reducing the variation of the measured intensities. A great improvement of the precision in the measurement was achieved with the method of repeated readings when the time interval between each reading is more than about 1 h.

A novel pulse-shape-analysis method that uses the similarity of patterns to recognize individual pulse shapes is presented. The principle of this method is described. The method has been applied to improve the gamma-ray energy spectrum of a CdZnTe semiconductor detector. The energy spectrum has been improved with increased photopeak efficiency and better energy resolution.

Eigenvalue separation is an important index of power reactor stability in the view of neutronics. In recent years, a long, slender Position Sensitive Proportional Counter has been developed. By using this proportional counter, the spatial neutron flux distribution of a nuclear reactor can be measured rapidly. Using these characteristics, we intend to develop the experimental method evaluating eigenvalue separation as an index for reactor stability. A method based on spatial integrals of neutron flux measured at a subcritical system with an external source is proposed. To verify this method, experiments were carried out at Kyoto University Critical Assembly. It was confirmed that eigenvalue separation can be evaluated within a relative error of 10%.

We have studied the improvement of gamma-ray energy spectrum characteristics of a CdZnTe semiconductor detector by means of a neural network algorithm. The neural network recognizes the pulse shapes and determines the corrective magnification factors of pulse heights. After the corrective procedure with the network, the gamma-ray energy spectra of some radioisotope sources become more ideal with better energy resolution.

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.

We present a new idea to apply a parallel plate collimator to an electronically-collimated gamma camera, to reduce the complexity of image reconstruction. Back projected images obtained experimentally and using Monte Carlo simulations are shown. Using the collimator, backprojected images of single low-energy gamma-ray emitters can be obtained with better spatial resolution and larger signal-to-noise ratio than those obtained without the collimator. The results show that this method is useful in obtaining Tc-99m images.