Neutron rem meters are routinely used for real-time field measurements of neutron dose equivalent where neutron spectra are unknown or poorly characterized. These meters are designed so that their response per unit fluence approximates an appropriate fluence-to-dose conversion function. Typically, a polyethylene moderator assembly surrounds a thermal neutron detector, such as a BF3 counter tube. Internal absorbers may also be used to further fine-tune the detector response to the shape of the desired fluence conversion function. Historical designs suffer from a number of limitations. Accuracy for some designs is poor at intermediate energies (50 keV-250 keV) critical for nuclear power plant dosimetry. The well-known Andersson-Braun design suffers from angular dependence because of its lack of spherical symmetry. Furthermore, all models using a pure polyethylene moderator have no useful high-energy response, which makes them inaccurate around high-energy accelerator facilities. This paper describes two new neutron rem meter designs with improved accuracy over the energy range from thermal to 5 GeV. The Wide Energy Neutron Detection Instrument (WENDI) makes use of both neutron generation and absorption to contour the detector response function. Tungsten or tungsten carbide (WC) powder is added to a polyethylene moderator with the expressed purpose of generating spallation neutrons in tungsten nuclei and thus enhance the high-energy response of the meter beyond 8 MeV. Tungsten's absorption resonance structure below several keV was also found to be useful in contouring the meter's response function. The WENDI rem meters were designed and optimized using the Los Alamos Monte Carlo codes MCNP, MCNPX, and LAHET. A first generation prototype (WENDI-I) was built in 1995 and its testing was completed in 1996. This design placed a BF3 counter in the center of a spherical moderator assembly, whose outer shell consisted of 30% by weight WC in a matrix of polyethylene. A borated silicone rubber (5% boron by weight) absorber covered an inner polyethylene sphere to control the meter's response at intermediate energies. A second generation design (WENDI-II) was finalized and tested in 1999. It further extended the high-energy response beyond 20 MeV, increased sensitivity, and greatly facilitated the manufacturing process. A 3He counter tube is located in the center of a cylindrical polyethylene moderator assembly. Tungsten powder surrounds the counter tube at an inner radius of 4 cm and performs the double duty of neutron generation above 8 MeV and absorption below several keV. WENDI-II is suitable for field use as a portable rem meter in a variety of work place environments, and has been recently commercialized under license by Eberline Instruments, Inc. and Ludlum Measurements, Inc. Sensitivity is about a factor of 12 higher than that of the Hankins Modified Sphere (Eberline NRD meter) in a bare 252Cf field. Additionally, the energy response for WENDI-II closely follows the contour of the Ambient Dose Equivalent per unit fluenc...
The usual Bonner sphere set consists of six to eight high density polyethylene spheres with diameters varying from 3 to 12 inches. Either a BF, counter, LiI scintillator, or ,He detector is located at the center of each sphere to detect the moderated neutrons. The responses of these spheres for high energy neutrons are very low even for the largest 12 inch sphere. To increase the response for high energy neutrons, high Z material such as Pb, W, TI, or Au can be added to the sphere to utilize the (n,xn) reaction of these materials. Monte Carlo Simulations for the Pb-added case with a 3He detector as an example will be presented. The response at high energy for this case is enhanced by a factor of 5 to 8 depending upon the sphere diameter.The use of the Bonner sphere sets for neutron spectra measurements above 20 MeV has been limited because of small neutron interaction cross sections in the sphere materials. Recent calculations on the use of 3He detectors in Bonner spheres for neutrons up to 20 MeV are given in [3]. The addition of fissile materials such as 235U would be ideal because of their production of induced-fission neutrons by the higher energy neutrons.However, such materials were not used because of their inherent radioactivity. Enhanced response can be obtained through the use of other high Z materials, such as W, Au, T1, or Pb which have large (n,xn') cross sections. Lead was chosen for these calculations.This paper describes calculations done to investigate the use of lead to enhance the high energy response.
Criticality dosimeters were exposed to different degraded neutron and g arm a-ray energy spectra from the Los Alamos Solution High Energy Burst Assembly (SHEBA). The liquid critical test assembly was operated in the continuous mode to provide a mixed source of neutron and gamma-ray radiation for the evaluation of Los Alamos criticality detector systems.Different neutron and gamma-ray spectra were generated by operating the reactor (a) shielded by 12 cm of Lucite, (b) unshielded, (c) shielded by 20 cm of concrete, and (d) shielded by 15 cm of steel. This report summarizes the dosimetry measurements conducted f r these different configurations.In-air measurement, were conduc .ed with shielded and unshielded area and personnel dosimeters. Phantom measurements were made using personnel dos meters. Combined blood-sodiun and hair sulfur activation measurements of absorbed dose were also maH°. in addition, indiun foils placed on phantoms were evaluated fo*-the purpose of screening personnel for radiation expo? I.
The appropriateness of the Cu activation foil for determining the neutron dose in the energy region from 1 eV to 1 MeV has been investigated for spectra of seven different criticality accident configurations. A program was written for folding the published spectra with the Cu activation cross sections and with the fluence-to-dose or kerma conversion factors. It is shown that for these spectra the neutron dose and kerma result primarily from the energy region above 15 keV whereas the measured Cu activity is mainly determined by the fluence in the region between 1 eV and 15 keV. Uncertainties in the fluence spectrum in the low-energy region between 1 eV and 15 keV, which in reality do not affect the dose contribution, might 6k lead to large deviations in the measured Cu activity and hence to the derived dose in the 1 eV to 1 MeV range. Use of 10 B shielding for attenuating the fluence in the 1-eV to 15-keV region was evaluated, leading to the conclusion that the necessary amount of boron material is unacceptably large and would appreciably increase the cost of the dosimeter currently used at Los Alamos. The lower limit of neutron detectability would also be increased.
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