Thermal Neutron Field Research Articles

Development and characterization of real-time personal neutron dosemeter with two silicon detectors

We have developed a new real-time personal neutron dosemeter containing two neutron sensors, a fast and a slow neutron sensor. The former sensor is a p-type silicon semiconductor detector. The slow neutron sensor is also a p-type silicon semiconductor detector and natural boron is doped on the aluminum electrode to produce 10B(n,α) reactions. A thin polyethylene radiator is contacted on the front surface of each sensor to produce recoil protons. The neutron detection efficiencies of these sensors were measured in a thermal neutron field and monoenergetic neutron fields from 8 keV to 22 MeV. By taking the weighted sum of counts given by the two sensors, the detection efficiency could be made to approach to the fluence-to-dose equivalent conversion factor given by ICRP 51. Field tests of the performance of this neutron dosemeter were performed in the radiation environments around several nuclear facilities, including reactor, accelerator, radioisotope and nuclear fuel handling facilities. Based on the results of these field tests, we conclude that our dosemeter is able to provide a reading of the neutron dose equivalent within a factor of 2 margin of accuracy.

An accelerator‐based epithermal photoneutron source for boron neutron capture therapy

Boron neutron capture therapy is an experimental cancer radiotherapy modality in which a boronated pharmaceutical that preferentially accumulates in malignant tissue is administered, followed by exposure of the tissue in the treatment volume to a thermal neutron field. Presently in the US and Europe, for treatment of deeper‐seated tumors, an epithermal beam is felt preferable in generating the necessary thermal neutron field. Boronated cells are selectively destroyed via energy deposition resulting from the interaction. Current usable beams are reactor‐based; a viable alternative is production of the epithermal neutron beam using an accelerator. Various proposed accelerator‐based designs exist, most based on proton beams with beryllium or lithium targets. This dissertation examines the efficacy of a novel approach incorporating an electron linear accelerator (4 to 8 MeV) in the production of a photoneutron source, which may help to resolve present concerns associated with some accelerator‐based sources, including target cooling. A conceptual design to produce epithermal photoneutrons, with high energy bremsstrahlung on deuterium targets, is presented along with computational and experimental neutron production data. A clinically acceptable epithermal neutron flux on the order of neutrons/second per milliampere of electron current is shown obtainable. Additionally, the neutron beam is modified and characterized, employing two unique moderating materials (an composite and a stacked Al/Teflon design).

Dependence of TLD thermoluminescence yield on absorbed dose in a thermal neutron field

The emission from 6LiF and 7LiF thermoluminescence dosimeters (TLDs) exposed to the mixed field of thermal neutrons and γ-rays of the thermal facility of a TRIGA MARK II nuclear reactor has been investigated for various thermal neutron fluences of the order of magnitude of those utilised in radiotherapy, with the purpose of investigating the reliability of TLD readouts in such radiation fields and of giving some information for better obtainment of the absorbed dose values. The emission after exposure in this mixed field is compared with the emission after γ-rays only. The glow curves have been deconvoluted into gaussian peaks, and the differences in the characteristics of the peaks observed for the two radiation fields, having different linear energy transfers, and for different doses are shown. Irreversible radiation damage in dosimeters having high sensitivity to thermal neutrons is also reported, showing a memory effect of the previous thermal neutron irradiation history which is not restored by anneal treatment.

Calibration study on the triton response of CR-39 track detector

Triton response of the CR-39 track detector has been studied at energy ranges below 2.7 MeV. Triton irradiation was made both on the top and the well polished side surface of the CR-39 sample sheets, using 6Li(n,t) 4He reaction in a thermal neutron field. At large residual ranges more than 10 μm, fine results of etch rate ratio were obtained by analysing the growth curves of etch-pit radius on the top surface. But the response at near the track end-point could not be obtained well in this method, as a result of the missing track effect. Clearly resolved response at short residual ranges was attained by means of the shape analysis of the etch-pit profile observed on the side surface.

Semi-insulating bulk GaAs as a semiconductor thermal-neutron imaging device

Thermal-neutron Schottky barrier detector arrays fabricated from semi-insulating bulk GaAs are presently being tested. The devices use a film of 10B to convert the incident thermal-neutron field into α particles and lithium ions, either of which interact in the detector. Bulk GaAs Schottky barrier detectors are relatively radiation hard to thermal neutrons and γ rays and have shown reasonably good energy resolution for α particles. Additionally, reverse biased radiation detectors fabricated from semi-insulating bulk GaAs have been shown to have truncated electric field distributions, resulting in the formation of a high field active region and a considerably lower field dead region. The device is sensitive only to electron-hole pairs excited in the high field region, thus the truncated field effect is advantageous as an inherent γ-ray discrimination feature for neutron detectors. Preliminary results show no indication of device degradation after over 2400 hr in a thermal-neutron beam from a reactor. Images have been formed of 1, 1.5, and 2 mm holes and crosses from 2 mm thick Cd templates.

Measurement of /spl gamma/-ray dose in a thermal neutron field by using a /sup 3/He-filtered GM counter

A new type of /sup 3/He-filtered GM counter has been used to measure the dose of /spl gamma/-rays existing in a thermal neutron field. The detector assembly consists of a commercially available GM counter (8.8 mm diameter and 31.8 mm height) surrounded by a cylindrical aluminum vessel (220.5 mm diameter and 163 mm height) filled with /sup 3/He gas in the pressure range up to 6 atmospheres. The contribution of secondary /spl gamma/-rays produced by thermal neutron captures in the GM counter itself can be excluded by increasing the gas pressure of /sup 3/He, because the thermal neutron capture reaction of /sup 3/He(n,p)T results in the emission of charged particles of a proton and a triton. In order to determine the response of the GM counter, the deposited energy distribution of the counter to monoenergetic /spl gamma/-rays was calculated by the EGS4 Monte Carlo code in the /spl gamma/-ray energy range from 60 keV to 10 MeV and then the response was derived through the calibration of the counter at /spl gamma/-ray energies of 60 keV, 662 keV and 1.17 MeV +1.33 MeV. On the other hand, the /spl gamma/-ray energy distribution in the thermal neutron field was calculated by the MCNP Monte Carlo calculation, assuming the three dimensional experimental assembly of a graphite pile including the concrete floor in the experimental room. As a result, the /spl gamma/-ray fluence and the effective dose equivalent in the thermal neutron field have been determined by using the GM counter response, the /spl gamma/-ray energy distribution in the field and the conversion factor from fluence to effective dose equivalent.

Transmutation of Minor Actinide in Well Thermalized Neutron Field and Application of Advanced Neutron Source(ANS).

Transmutation of Minor Actinide (MA) in a well thermalized neutron field was studied. Since MA nuclides have large effective cross sections in the well thermalized neutron field, the transmutation in the well thermalized neutron field has an advantage of high transmutation rate. However, the transmutation rate largely decreases by accumulation of 246Cm when MA is transmuted only in the well thermalized neutron field for a long period. An acceleration method of burn-up of 246Cm was studied. High transmutation rate can be obtained by providing a neutron field with high flux in the energy region between 1 and 100eV. Two stage transmutation using the well thermalized neutron field and this field can transmute MA rapidly. The applicability of the Advanced Neutron Source (ANS) to the transmutation of MA was examined for a typical MA with the composition in the high-level waste generated in the conventional PWR. If the ANS is applied without changing the fuel inventory, the amount of MA which corresponds to that...

Transmutation of Minor Actinide in Well Thermalized Neutron Field and Application of Advanced Neutron Source (ANS)

Transmutation of Minor Actinide (MA) in a well thermalized neutron field was studied. Since MA nuclides have large effective cross sections in the well thermalized neutron field, the transmutation in the well thermalized neutron field has an advantage of high transmutation rate. However, the transmutation rate largely decreases by accumulation of 246Cm when MA is transmuted only in the well thermalized neutron field for a long period. An acceleration method of burn-up of 246Cm was studied. High transmutation rate can be obtained by providing a neutron field with high flux in the energy region between 1 and 100eV. Two stage transmutation using the well thermalized neutron field and this field can transmute MA rapidly. The applicability of the Advanced Neutron Source (ANS) to the transmutation of MA was examined for a typical MA with the composition in the high-level waste generated in the conventional PWR. If the ANS is applied without changing the fuel inventory, the amount of MA which corresponds to that produced by a conventional 1,175 MWe PWR in one year can be transmuted by the ANS in one year. Furthermore, the amount of the residual can be reduced to about 1g (10−5 of the initial MA weight) by continuing the transmutation for 5 years owing to the two stage transmutation.

Neutron Economy of Transmutation of TRU in Thermal and Fast Neutron Fields

Neutron economy of the transmutation of TRU was examined in well thermalized, thermal and fast neutron fields. Burn-up chains of 237Np, 241Am and 243Am, which are the main TRU nuclides in the high level waste, were calculated in the flux region from 1014 to 1017 n/cm2.s. Numbers of neutrons absorbed and produced of each chain were calculated using JENDL-3. The net number of neutron produced n net, which was obtained by the difference of the two numbers, largely varied with the neutron fields, the nuclides and the flux levels. The n net value in the fast neutron field was positive (0.0–1.0) for 237Np, 241Am, 243Am and TRU with the nuclide composition in the high-level waste generated by the conventional PWR. The transmutation of TRU by fission can be performed with producing neutrons in the fast neutron field. On the other hand, the n net value was negative for the well thermalized and thermal neutron fields. For TRU in the high-level waste, the values in those fields were —1.0 at 1014 n/cm2.s and 0.0 at 1...

Improved apparatus for neutron capture therapy of rat brain tumors

Purpose: The assembly for irradiating tumors in the rat brain at the thermal neutron beam port of the Brookhaven Medical Research Reactor was redesigned to lower the average whole-body dose from different components of concomitant radiation without changing the thermal neutron fluence at the brain tumor. Methods and Materials: At present, the tumor-bearing rat is positioned in a rat holder that functions as a wholebody radiation shield. A 2.54 cm-thick collimator with a centered conical aperture, 6 cm diameter tapering to 2 cm diameter, is used to restrict the size of the thermal neutron field. Using the present holder and collimator as a baseline design, Monte Carlo calculations and mixed-field dosimetry were used to assess new designs. Results: The computations indicate that a 0.5 cm-thick plate, made of 6Li 2C0 3 dispersed in polyethylene (Li-poly), instead of the existing rat holder, will reduce the whole-body radiation dose. Other computations show that a 10.16 cm-thick (4 inches) Li-poly collimator, having a centered conical aperture of 12 cm diameter tapering to 2 cm diameter, would further reduce the whole-body dose. Conclusion: The proposed irradiation apparatus of tumors in the rat brain, although requiring a 2.3-fold longer irradiation time, would reduce the average whole-body dose to less than half of that from the existing irradiation assembly.

Neutron Economy of Transmutation of TRU in Thermal and Fast Neutron Fields.

Neutron economy of the transmutation of TRU was examined in well thermalized, thermal and fast neutron fields. Burn-up chains of 237Np, 241Am and 243Am, which are the main TRU nuclides in the high level waste, were calculated in the flux region from 1014 to 1017 n/cm2.s. Numbers of neutrons absorbed and produced of each chain were calculated using JENDL-3. The net number of neutron produced n net, which was obtained by the difference of the two numbers, largely varied with the neutron fields, the nuclides and the flux levels. The n net value in the fast neutron field was positive (0.0–1.0) for 237Np, 241Am, 243Am and TRU with the nuclide composition in the high-level waste generated by the conventional PWR. The transmutation of TRU by fission can be performed with producing neutrons in the fast neutron field. On the other hand, the n net value was negative for the well thermalized and thermal neutron fields. For TRU in the high-level waste, the values in those fields were —1.0 at 1014 n/cm2.s and 0.0 at 1016 n/cm2.s. In the high flux region of 1016 n/cm2.s, TRU in the high-level waste can be transmuted by fission without consuming additional neutrons. In the flux region about 1014 n/cm2.s, the transmutation of TRU in the high-level waste by fission requires about one neutron.

リアルタイム式広帯域中性子個人線量計の開発と特性評価

The authors developed a real-time personal neutron dosimeter which could give neutron dose equivalent over wide energy region from thermal to 10 odd MeV by using 2 silicon detectors, fast neutron sensor and slow neutron sensor. The energy response of this dosimeter was evaluated under thermal neutron field, monoenergetic neutron field between 200keV and 15MeV, and moderated 252Cf neutron field. The neutron dose equivalent was estimated by adding neutron dose equivalent below 1MeV given by slow neutron sensor and that above 1MeV by fast neutron sensor. It was verified from various field tests that this dosimeter is able to give neutron dose equivalent within a factor of 2 margin of accuracy in reactor, accelerator, fusion research and nuclear fuel handling facilities. This dosimeter has more than one order higher sensitivity than conventional personal neutron dosimeters and is insensitive to γ-rays up to about 500 mSv/h. This dosimeter will soon be commercially available as a personal dosimeter which gives neutron and γ-ray dose equivalents simultaneously by installing γ-ray silicon sensor.

Elemental assay of materials by NA and gamma-ray spectroscopy without standards and neutron flux monitoring

The theoretical expression of the induced radioactivity of a nuclide activated by neutron reactions is basically used to develop a mathematical model and an accurate method of evaluating concentrations of elements producing γ-ray emitting isotopes, without use of standards and of neutron flux monitoring. This initial “Without Standard” method is mainly applicable to materials obtained by special or industrial technologies, when precise values of the elemental concentrations, after the synthesis process, are required. A preliminary test of the method, using a Goodfellow Metals stainless steel material irradiated in a thermal neutron field is reported. The results confirm the applicability of the “Without Standard” method to the NAA experiments for the elemental analysis of materials.

Energy Response of Neutron Area Monitor with Silicon Semiconductor Detector.

A neutron area monitor should provide an energy response of a 1-cm dose equivalent according to International Commission on Radiological Protection recommendations. A prototype neutron area monitor has been developed using a silicon detector with a polyethylene moderator. The characteristics of neutron energy response to dose equivalent on this monitor were confirmed by the thermal neutron field and a monoenergetic fast neutron field using an accelerator. The neutron detection is based on the principle of the ionization of alpha particles generated by the nuclear reaction of thermal neutrons ({sup 10}B(n,{alpha}){sup 7}Li) and recoil protons generated from the interaction of fast neutrons with hydrogen compounds (proton radiator). One important feature of the monitor is its performance compensatory precision within {plus minus}50% in response to a 1-cm dose equivalent. Another is its performance for the energy range from {approximately}0.025 eV to 15 MeV with only one silicon detector.

A gas flow proportional detection system for the determination of boron in aerosol samples and other sample types

This work describes the design, fabrication, and testing of a lucite bodied proportional gas detection system for the analysis of boron in selected samples via detection of the charged particles produced in the 10B(n,α)7Li reaction induced by thermal neutrons. The detector was designed for internal placement of samples; the sample types of major interest were airborne aerosols collected on filters or particulate impaction plates. Samples were irradiated with the detector in the thermal neutron field produced in the graphite thermal column of the University of Lowell's one megawatt research reactor. Determined sensitivities for boron varied from 6.2·10−8 to 1.73·10−6 cpm·ng−1 (n·cm−2J·s−1)−1 depending on the physical characteristics of the samples. For a nominal counting time of ten minutes the lower limit of mass detection of natural boron was determined to be 12.1 nanograms. The analytical method was applied to the estimation of boron in fourteen samples of natural aerosols collected on membrane filters. Analysis of prepared samples and natural aerosol samples by ICP emission showed good agreement with analysis via the (n,α) reaction. Application of the method to other sample types was demonstrated by the determination of boron in samples of borosilicate glass and borated polyethylene.

Central nervous system radiation syndrome in mice from preferential 10B(n, alpha)7Li irradiation of brain vasculature.

Ionizing radiations were directed at the heads of anesthetized mice in doses that evoked the acute central nervous system (CNS) radiation syndrome. Irradiations were done using either a predominantly thermal neutron field at a nuclear reactor after intraperitoneal injection of 10B-enriched boric acid or 250-kilovolt-peak x-rays with and without previous intraperitoneal injection of equivalent unenriched boric acid. Since 10B concentrations were approximately equal to 3-fold higher in blood than in cerebral parenchyma during the reactor irradiations, more radiation from alpha and 7Li particles was absorbed by brain endothelial cells than by brain parenchymal cells. Comparison of the LD50 dose for CNS radiation lethality from the reactor experiments with the LD50 dose from the x-ray experiments gives results compatible with morphologic evidence that endothelial cell damage is a major determinant of acute lethality from the CNS radiation syndrome. It was also observed that boric acid is a low linear energy transfer radiation-enhancement agent in vivo.

An Examination of Neutron Induced Autoradiography and Gamma Spectroscopy for the Study of Paintings

Neutron-induced autoradiography is a further step in the development of techniques for the examination of paintings that complements x-ray radiography and often yields unique information that cannot be obtained by other means. After exposure to a field of thermal neutrons (<0.3eV), selected elements in the painting form radioactive nuclides that in turn decay at known half-lives with the emission of gamma rays and charged particles. The charged particles, mainly electrons, expose radiographic film that is held in close contact with the painting by a very slight vacuum. Unlike x-ray radiography, which involves detecting x-rays that have passed through the painting to expose a film, autoradiography uses one activation with neutrons followed by sequential film exposures to capture the images of activated elements that have different half-lives. The autoradiographs show the distribution patterns of the pigments in which the elements occur. These pigments may be on the surface and visible in the painting or they may be beneath the surface. Often autoradiography uncovers the preliminary sketch or underpainting preceding the rough blocking in of the subject and the subsequent paint layers that create the final image. X-rays reveal almost exclusively the presence of pigments containing heavy elements such as lead and barium; autoradiography identifies the presence of activated pigments and distinguishes the fine details of paint application as well as changes in composition.

Experimental determination of spectral indies in low enriched UO2-H2O lattices. On the basis of parallel irradiations.

To get information about the neutron spectrum in low enriched UO2-H2O lattices, the spectral indices SI(U8c/Dy) and SI(U8c/U5f) were measured on the basis of the parallel irradiation technique, which basically irradiates activation foils both in a neutron field to be investigated and in a reference field of thermal neutrons. In the present study, a fuel pellet of UO2 was used for the measurement of activities caused by the neutron capture of 238U and the fission of 235U. Besides the technical details of the measurements, the origins of experimental errors are listed with the method how to eliminate them. The measurements were carried out in lattices of different fuel enrichment to demonstrate the capability of the present method, and the experimental results were compared with the calculated ones. It was found that the results of the present measurements are useful to assess the validity of the cell calculations.

Detector asselby with a diamond detector for recording neutrons

The BDPN-1K detector assembly for neutron radiation comprises the detector and a preamplifier connected by a 10-m-long cable. The detector can be disassembled. Before the sensitive surface of the diamond there is placed a cover with a converter disk cemented to it; this disk is a U/sub 3/O/sub 8/ layer applied to an aluminum foil. Since the detector can be disassembled, the disk can be replaced and other materials can be used as the converter. In order to determine the optimal thickness of the converter, the sensitivity of recording thermal neutrons with a particular detector assembly was measured and converters with various surface densities of /sup 235/U were used in the detector. Measurements were made in a standard field of thermal neutrons from a plutonium-beryllium neutron source inserted in a paraffin moderator.

Lithium Fluoride (TLD-700) Response to a Mixed Thermal Neutron and Gamma Field

Irradiation studies of TLD-700 dosemeters in a mixed gamma and thermal neutron field show that TLD-700 responds to thermal neutron exposure. The results indicate that the error in estimating gamma dose in a mixed field by assuming that the TLD-700 response measures only the gamma dose can be significant in a mixed field with a high thermal neutron component.