Monte Carlo simulations were performed to extend existing neutron personal dose equivalent fluence-to-dose conversion coefficients to an energy of 250 MeV. Presently, conversion coefficients, H(p,slab)(10,alpha)/Phi, are given by ICRP-74 and ICRU-57 for a range of angles of radiation incidence (alpha = 0, 15, 30, 45, 60 and 75 degrees ) in the energy range from thermal to 20 MeV. Standard practice has been to base operational dose quantity calculations <20 MeV on the kerma approximation, which assumes that charged particle secondaries are locally deposited, or at least that charged particle equilibrium exists within the tally cell volume. However, with increasing neutron energy the kerma approximation may no longer be valid for some energetic secondaries such as protons. The Los Alamos Monte Carlo radiation transport code MCNPX was used for all absorbed dose calculations. Transport models and collision-based energy deposition tallies were used for neutron energies >20 MeV. Both light and heavy ions (HIs) (carbon, nitrogen and oxygen recoil nuclei) were transported down to a lower energy limit (1 keV for light ions and 5 MeV for HIs). Track energy below the limit was assumed to be locally deposited. For neutron tracks <20 MeV, kerma factors were used to obtain absorbed dose. Results are presented for a discrete set of angles of incidence on an ICRU tissue slab phantom.

The high-energy neutron response of three passive dosemeters in use at the Los Alamos National Laboratory (LANL) has been investigated using metrology-grade fields. The dosemeters include the LANL Model 8823 TLD badge and the LANL PN3 track etch device. Both are dosemeters of record at LANL. The third device was the Personal Neutron Dosemeter (PND), a superheated emulsion device, manufactured by Bubble Technology Industries, Inc. (BTI). The response of the three dosemeters at neutron energies exceeding 10 MeV was assessed with monoenergetic neutrons at the Physikalisch-Technische Bundesanstalt facility (14.8 and 19 MeV). For the sake of completeness, data collected at lower energies are also included in this study. High-energy quasi-monoenergetic beams produced by the cyclotron facilities at the Université Catholique de Louvain (UCL) and the The Svedberg Laboratory (TSL) were also utilised as part of this study. These measurements were made to better understand and help interpret dosemeter readings obtained by workers at high-energy accelerators, such as the 800 MeV spallation neutron source facility located at the Los Alamos Neutron Science Center (LANSCE).

CHELSI is a CsI-based portable spectrometer being developed at Los Alamos National Laboratory for use in high-energy neutron fields. Based on the inherent pulse shape discrimination properties of CsI(Tl), the instrument flags charged particle events produced via neutron-induced spallation events. Scintillation events are processed in real time using digital signal processing and a conservative estimate of neutron dose rate is made based on the charged particle energy distribution. A more accurate dose estimate can be made by unfolding the 2D charged particle versus pulse height distribution to reveal the incident neutron spectrum from which dose is readily obtained. A prototype probe has been assembled and data collected in quasi-monoenergetic fields at The Svedberg Laboratory (TSL) in Uppsala as well as at the Los Alamos Neutron Science Center (LANSCE). Preliminary efforts at deconvoluting the shape/energy data using empirical response functions derived from time-of-flight measurements are described.

The Health Physics Measurements Group at the Los Alamos National Laboratory (LANL) has initiated a study of neutron reference fields at selected US Department of Energy (DOE) calibration facilities. To date, field characterisation has been completed at five facilities. These fields are traceable to the National Institute for Standards and Technology (NIST) through either a primary calibration of the source emission rate or through the use of a secondary standard. However, neutron spectral variation is caused by factors such as room return, scatter from positioning tables and fixtures, source anisotropy and spectral degradation due to source rabbits and guide tubes. Perturbations from the ideal isotropic point source field may impact the accuracy of instrument calibrations. In particular, the thermal neutron component of the spectrum, while contributing only a small fraction of the conventionally true dose, can contribute a significant fraction of a dosemeter's response with the result that the calibration becomes facility-specific. A protocol has been developed to characterise neutron fields that relies primarily on spectral measurements with the Bubble Technology Industries (BTI) rotating neutron spectrometer (ROSPEC) and the LANL Bonner sphere spectrometer. The ROSPEC measurements were supplemented at several sites by the BTI Simple Scintillation Spectrometer probe, which is designed to extend the ROSPEC upper energy range from 5 to 15 MeV In addition, measurements were performed with several rem meters and neutron dosemeters. Detailed simulations were performed using the LANL MCNPX Monte Carlo code to calculate the magnitude of source anisotropy and scatter factors.

The personnel dosimetry operations team at the Los Alamos National Laboratory (LANL) has accepted the laser illuminated track etch scattering (LITES) dosemeter reader into its suite of radiation dose measurement instruments. The LITES instrument transmits coherent light from a He-Ne laser through the pertinent track etch foil and a photodiode measures the amount of light scattered by the etched tracks. A small beam stop blocks the main laser light, while a lens refocuses the scattered light into the photodiode. Three stepper motors in the current LITES system are used to position a carousel that holds 36 track etch dosemeters (TEDs). Preliminary work with the LITES system demonstrated the device had a linear response in counting foils subjected to exposures up to 50 mSv (5.0 rem). The United States Department of Energy requires that the annual general employee dose not exceed 50 mSv (5.0 rem). On a regular basis, LANL uses the Autoscan-60 reader system (Thermo Electron Corp.) for counting track etch dosemeters. However, LANL uses a 15 h etch process for CR-39 dosemeters, and this produces more and larger track etch pits than the 6 h etch used by many institutions. Therefore, LANL only uses the Autoscan-60 for measuring neutron dose equivalent up to exposure levels of approximately 3 mSv (300 mrem). The LITES system has a measured lower limit of detection of approximately 0.6 mSv (60 mrem), and it has a correlation coefficient of R (2) = 0.99 over an exposure range up to 500 mSv (50.0 rem). A series of blind studies were done using three methods: the Autoscan-60 system, manual counting by optical microscope and the LITES instrument. A collection of track etch dosemeters of unknown neutron dose equivalent (NDE) were analysed using the three methods, and the performance coefficient (PC) was calculated when the NDE became known. The Autoscan-60 and optical microscope methods had a combined PC = 0.171, and the LITES instrument had a PC = 0.194, where a PC less than or equal to 0.300 is considered satisfactory.

The intrinsic pulse shape discrimination properties of CsI(Tl) form the basis of a high-energy neutron (> 20MeV) spectrometer (CHELSI) currently being developed at LANL that shows promise in satisfying the requirements of an ideal survey meter; lightweight, portable and real time display of dose.Charged particle spallation products generated in the scintillator via neutron interactions are identified on the basis of pulse shape using digital pulse processing. Conservative estimates of dose rate can be given in real time based on count rates and pulse height distributions. More accurate dose measurements can be done off-line using unfolding methods to analyze stored pulse shape versus energy data.As a precursor to the development of a portable instrument, data has been obtained using a 1" x 1" CsI-based probe and a digital spectrometer. This system has been used to collect data on the 90 m flight path at the LANSCE/WNR facility at an average neutron energy of 335 MeV. The spectrometer has the capability, in addition to storing individual waveforms for later analysis, of recording time-of-flight data and calculating a pulse shape parameter and pulse height for each scintillation event in real time. Combining these data with traditional multichannel analyzer data has yielded a set of empirical response functions with respect to neutron energy. Analysis of the charged particle spectra has yielded an overall average count rate of 0.12 +/- 0.02 cps/mu Sv h for a 1" x 1" CsI(Tl) scintillator in this neutron field. (c) 2006 Elsevier B.V. All rights reserved.

A simple dosemeter made of a sulphur tablet, bare and cadmium-covered indium foils and a cadmium-covered copper foil has been modelled using MCNP5. Studies of the model without phantom or other confounding factors have shown that the cross sections and fluence-to-dose factors generated by the Monte Carlo method agree with those generated by analytic expressions for the high energy component. In this study, a comparison of the effect of location on phantoms and an extension to low and intermediate energies is done. The activities expected from exposure to four critical assemblies on phantom is calculated and compared with observations.

Analysis of accident dosemeters usually involves the use of laboratory-based counting equipment. Gamma spectrometers are used for indium, copper and gold, and alpha-beta detectors for sulphur. This equipment is usually not easily transported due to the shielding required and the weight and delicacy of the counters. For intercomparison studies that require reading the dosemeters on site, a transportable system is required unless the site operating the study can count samples for all the participants. In the case of an actual accident these systems would have a difficulty in counting a large number of accident dosemeters. In an accident, personnel are usually subdivided according to their level of exposure. Those exposed to higher doses are treated immediately. An alternate system should be made available to handle the dosemeters worn by those personnel are likely to receive lower doses. Improvements in portable operational equipment for gamma and beta monitoring allow their use as spectrometers. Such a system was used for the SILENE intercomparison conducted at IRSN Valduc on 12 June and 19, 2002, and the preliminary results compared well with the other participants.

A simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. Studies of the model without phantom or other confounding factors have shown that the cross sections and fluence-to-dose factors generated by the Monte Carlo method agree with those generated by analytic expressions for the high energy component. The threshold cross sections for the detectors on a phantom were calculated. The resulting doses assigned agree well with exposures made to three critical assemblies. In this study the angular dependence on a phantom is studied and compared with measurements taken on the GODIVA reactor. The dosimeter positions on the phantom are facing the source, on the back and the side. In previous papers the modeling of a simple dosimeter made of a sulfur tablet, bare and cadmium covered indium foils and a cadmium covered copper foil has been modeled using MCNP5. The conclusion made was that most of the neutron dose from criticality assemblies results from the high energy neutron fluences determined by the sulfur and indium detectors. The results using doses measured from the GODIVA, SHEBA, and bare and lead shielded SILENE reactors confirmed this. Themore » angular dependence of an accident dosemeter is of interest in evaluating the exposure of personnel. To investigate this effect accident dosemeters were placed on a phantom and exposed to the GODIVA reactor at phantom orientations of 0{sup o}, 45{sup o}, 90{sup o}, 135{sup o}, and 180{sup o} to the assembly center line.« less

Initial calibration of a multisphere spectroscopy system has been completed at Los Alamos National Laboratory using four standard calibration scenarios. Spectrum unfolding was performed using three methods of constructing the default spectrum: simple parameter models, Monte Carlo calculations and physical measurement. Comparisons of the resulting spectra for each solution method are presented. Implications of the spectral solutions upon dosemeter characterisation are addressed.

In the application of criticality accident dosemeters the cross sections and fluence-to-dose conversion factors have to be computed. The cross section and fluence-to-dose conversion factor for the thermal and epi-thermal contributions to neutron dose are well documented; for higher energy regions (>100 keV) these depend on the spectrum assumed. Fluence is determined using threshold detectors. The cross sections require the folding of an expected spectrum with the reaction cross sections. The fluence-to-dose conversion factors also require a similar computation. The true and effective thresholds are used to include the information on the expected spectrum. The spectra can either be taken from compendia or measured at the facility at which the exposures are to be expected. The cross sections can be taken from data computations or analytic representations and the fluence-to-dose conversion factors are determined by various standards making bodies. The problem remaining is the method of computation. The purpose of this paper is to compare two methods for computing these factors: analytic and Monte Carlo.

The design of a portable high-energy (20-800 MeV) neutron spectrometer based on CsI or BaF2 is described. The particle discrimination properties of these scintillators allow the light-ion spallation products (p, d, t and alpha) from neutron interactions to be identified uniquely. One or more of the resulting pulse-height spectra can be unfolded to reveal the incident neutron spectrum. Dosimetric quantities can then be calculated based on the unfolded spectrum. Due to the high stopping power of these scintillators, modest-sized crystals are suitable for this application. Combined with advances in electronics, a lightweight instrument capable of on-line particle discrimination with a real-time display of neutron-induced count rate is feasible. Preliminary experimental data are presented, and the importance of validating MCNPX-generated response functions is discussed. A brief discussion on future work follows.

Several neutron spectrometers manufactured by Bubble Technology Industries (BTI) were tested and evaluated in a variety of neutron fields. Findings and conclusions are presented for the following BTI instruments: a modification of the Rotational Spectrometer (ROSPEC) that includes a thermal and epithermal capability, the Simple Scintillation Spectrometer that is used in conjunction with the ROSPEC to extend its high-energy range, and the MICROSPEC N-Probe which is capable of providing a crude spectrum over the energy range from thermal to 18MeV. The main objective of these measurements was to determine the accuracy of both the energy spectrum and dose equivalent information generated by these devices. In addition, the dose response of the Wide-Energy Neutron Detection Instrument (WENDI-II) was measured in all neutron fields relative to a bare 252Cf calibration. The performance of the WENDI-II rem meter was compared to the dose information generated by the neutron spectrometers. The instruments were irradiated to bare 252Cf and 241AmBe sources, and in a series of moderated 252Cf fields using a standard D2O sphere and a set of polyethylene spheres. The measured spectra were benchmarked with a set of detailed Monte Carlo calculations with the same energy bin structure as that of the instruments under test. These calculations allowed an absolute comparison to be made with the measurements on a bin by bin basis. The simulations included the effects of room return and source anisotropy.

Los Alamos National Labs (LANL) has developed an etched track foil (CR-39) reader for neutron dose between 0 and 50.0 mSv. Currently, the US Department of Energy mandates general employee annual exposure not to exceed 50.0 mSv (5 rem). At LANL, due to a non-linear response at higher exposures, accepted practice only uses an Autoscan 60 system up to 3 mSv. The LITES system, however, has demonstrated linear response to 50 mSv, where the proprietary design measures the amount of laser light scattered by the etched tracks, proportional to dose. A collection of calibrated foils was counted by an Autoscan 60 and the LITES prototype, and the Autoscan 60 showed good linearity when counting exposure up to about 15 mSv, but not for higher exposures. From 0 to 50 mSv, the Autoscan 60 had a correlation coefficient of R-2 = 0.941 and the LITES system had R-2 = 0.991.

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 fluence function [H'(10)/phi] above 0.1 MeV. Its energy response at 500 MeV is approximately 15 times higher than that of the Hankins and Andersson-Braun meters. Measurements of the energy and directional response of the improved meter are presented and the measured response function is shown to agree closely with the predictions of the Monte Carlo simulations in the range from 0.144 MeV to 19 MeV.

Track etch detectors were exposed to neutrons produced by a spallation target struck by a beam of 800 MeV protons. The fields were filtered by 0, 10, and 40 centimeters of polyethylene. The track etch dosimeters were exposed on a polyethylene phantom. The dosimeters were exposed bare and behind lead filters of 0.25, 0.50, 0.75, 1.00, 1.25 and 1.50 cm of lead with the face of the dosimeter perpendicular to the beam and bare and behind lead filters of 0.50, 1.0, and 1.5 cm of lead with angle of incidence 45{degree} and 75{degree}. Monte Carlo calculations of these experimental configurations were done using MCNP and LAHET with input from the calculated spectra. These results are compared with the experimental results to understand the basic processes involved in the production of tracks with high energy neutrons and develop a high energy neutron area monitor.

A problem in using Bonner spheres for the spectroscopy of neutron of energy greater than 20 MeV is that the response functions for differing polyethylene thicknesses become similar in shape, making the spectral unfolding quite difficult. The possibility that use of a neutron detector such as CR-39 that is sensitive to high-energy neutrons combined with a lead radiator that has a significant variation of cross section due to the Pb(n,xn) reaction above 20 MeV, led to the theoretical investigation of the response of a CR-39 detector with lead in front of it. The authors have calculated the response of CR-39 (polyallyl-diglycol carbonate) in terms of charged particles produced per-unit-incident neutron for a foil of 2-cm diameter and 0.15-cm thick. They also calculated another set, using cylindrical slugs of lead of 2-cm diameter and 1-cm and 2-cm thick, with the foil positioned on the end of the slug so that the neutron beam had to pass through the lead. MCNP was used to calculate the responses in all three cases for energies from 0.1 MeV to 20 MeV. LEHET was applied to calculations from 0.1 MeV to 1,000 MeV. The MCNP calculation yields H(n,p), C(n,{alpha}), and O(n,{alpha}) reactions; the LAHET computation produces the sum of proton recoil and deuteron, triton, and alpha-particle production. The LAHET calculation indicates that the enhanced response due to Pb(n,xn) production may prove useful in neutron spectroscopy.

Electrochemical and chemical etched track dosemeters were obtained from commercial suppliers (American Acrylics and NE Technology) and exposed to neutrons produced at the LAMPF WNR white neutron source with no shielding and filtered by four polyethylene blocks of 2.5, 5.1, 10.2 and 20.3 cm thickness. The neutron spectrum was determined using a U-235 fission chamber and time of flight measurements. Mean energies from 14 to 100 MeV were produced. Dose was calculated from the NCRP 38 flux-to-dose conversion. Response relative to bare californium for the electrochemical etch varied from 0.43 to 0.16, for the chemical etch from 0.56 to 0.33. The results are compared with NTA film which was exposed in the same configuration.

Electro-chemical and chemical track-etch dosemeters were obtained from commercial suppliers (American Acrylics and NE Technology) and exposed to neutrons produced at the LAMPF WNR white neutron source at 15 degrees with no shielding and filtered by polyethylene blocks of 2.5, 5.1, 10.2, 20.3 and 40.6 cm thickness. The neutron spectrum was determined using calculations. Mean energies from 28 to 300 MeV were produced. Dose was calculated from the NCRP-38 flux-to-dose conversion. The results are compared with NTA film which was exposed in the same configuration. The response of track etch dosimeters was found to reach a miniumum and then rise as the the average neutron energy increased. The response of the NTA film increased as the neutron energy increased.