Electrostatic precipitator (ESP) is high efficiency dust collector, which utilizes electric energy to remove particulate pollutant. The capital cost of ESP is relatively more expensive than general dust collectors. However, due to its small pressuredrop flow characteristics, the operational cost of ESP is usually lower than other particulate control devices (Lim et al., 1998). As a result, ESP is widely used in industry and in indoor air quality improvement. According to the principle of operation, air cleaners are sorted into filtration type, electrostatic type and mixing type. Offerman (1985) experimented with 11 commercial air cleaners, the results indicated that the electrostatic type had better efficiency at removing cigarette particles. Additionally, Don (2003) studied 4 kinds of commercial air cleaners (electrostatic type), the results indicated that collection efficiency, discharge current and ozone generation decreased over 18~26 days. A two-stage miniature ESP has been set up in this study (Figure 1). A high-voltage AC power supply was used to provide the energy for corona discharge and to produce high concentration single polarity ions. The alternating electric field discharge electrode range is 0 –kV to 11.2 –kV. The objectives of this research included studying the filtration characteristics of ESP using AC fields and the

This work presents the design of a CMOS quadrature voltage-controlled oscillator (QVCO) constructed in a totem-pole configuration in order to reuse the bias current and lower the power consumption. A new feature of using four RC delay lines connected at the output ports of an oscillator is adopted in order to obtain the desired quadrature phase shift of the oscillation signals. An experimental chip is designed and fabricated using 0.18-mum CMOS technology to verify the effectiveness of the design concept. The measurement result shows that the center oscillation frequency of the prototype is 6.3 GHz, associated with a 100-MHz tuning range and -108.7-dBc/Hz phase noise at 1-MHz offset. (C) 2004 Wiley Periodicals, Inc.

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 major objective of radiation protection is to determine the dose equivalent for routine radiation protection applications. As microdosimetry has developed over approximately three decades, its most important application has been in measuring radiation quality, especially in radiation fields of unknown or inadequately known energy spectra. In these radiation fields, determination of dose equivalent is not straightforward; however, the use of microdosimetric principles and techniques could solve this problem. In this paper, the authors discuss the measurement of lineal energy, a microscopic analog to linear energy transfer, and demonstrate the development and implementation of the variance-covariance method, a novel method in experimental microdosimetry. This method permits the determination of dose mean lineal energy, an essential parameter of radiation quality, in a radiation field of unknown spectrum, time-varying dose rate, and high dose rate. Real-time monitoring of changes in radiation quality can also be achieved by using microdosimetric techniques.

In June, 1995, Los Alamos National Laboratory hosted the 23rd U.S. Department of Energy sponsored Nuclear Accident Dosimetry Study at the Los Alamos Critical Experiments Facility. The participants tested their facilities accident dosimeters under a variety of neutrons fields produced by the Solution High Energy Burst Assembly (SHEBA) and the Godiva IV fast burst assembly. To provide useful information for the evaluation of the results, the neutron energy Spectrum was determined and the delivered absorbed dose to tissue. The measurement of the neutron energy spectrum on Godiva provides a unique problem in that the burst, which is nearly Gaussian in time, has a full width at half maximum of around 50 microseconds. The neutron spectrum was first determined at low-power while running at delayed critical using a standard set of Bonner spheres. At the same time, the response of a set of TLD dosimeters were measured. After that, measurements were conducted during a burst with another set of TLDs and with sulfur pellets.

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.

The responses of LiF and CaF{sub 2} TLD chips, calculated with the Monte Carlo code MCNP, show the following ratios: (1) for TLD chips free in the air exposed to a bare {sup 252}Cf neutron source; {sup 6}LiF / {sup 7}LiF / CaF{sub 2} = 1.0 / 0.43 / 0.21. (2) to a D{sub 2}O moderated {sup 252}Cf neutron source; {sup 6}LiF / {sup 7}LiF / CaF{sub 2} = 1.00 / 0.022 / 0.010. (3) for TLD chips placed inside a plastic holder and with a water phantom, exposed to a bare {sup 252}Cf source; {sup 6}LiF / {sup 7}LiF / CaF{sub 2} = 1.00 / 0.27 / 0.13 (4) to a D{sub 2}O moderated neutron source; {sup 6}LiF / {sup 7}LiF / CaF{sub 2} = 1.00 / 0.020 / 0.0070. (5) chips free in the air and to a {sup 137}CS photon source at energy of 661 keV, LiF / CaF{sub 2} = 1.00 / 1.05. (6) chips inside a plastic holder and with a water phantom, LiF / CaF{sub 2} = 1.00 / 1.10. Monte Carlo calculations for the responses tallied by energy deposition per unit mass (i.e., absorbed dose) did not include the difference in fluorescence mechanismmore » between two kinds of chips and the optical attenuations of the chip. The measured responses in terms of the photo-mulitiplier tube current may yield different ratios. We performed experimental measurements under the same given conditions used in the calculations. The results show reasonable agreements between calculations and measurements.« less

The Los Alamos Graphite Pile is used to measure the neutron emission rate of a variety of neutron sources against a secondary standard {sup 239}Pu/Be (Q{sub s} = 8.085 E+5 n s{sup -1}) neutron source in a reproducible geometry. The Graphite Pile is a National User Facility and is available to external customers for calibrating neutron sources. The cadmium-lined Pile has dimensions of 152.6 cm (L) x 152.6 cm (W) x 254.5 cm (H). Source-to-detector distances in the Pile can range from 21 cm (for weak source strengths) to 81 cm (for strong source strengths). Observations, benchmarked by Monte Carlo calculations, were made of relative count rates for several neutron sources at various source-to-detector distances. Results show that measurement of neutron emission rates can be determined within a few percent in comparison to the secondary standard source, and are also in close agreement in comparison to MCNP calculations. In the near future we intend to replace the {sup 239}Pu/Be neutron source with a NIST-traceable ANSI-type neutron source. This primary standard neutron source will eliminate concern for the increase in the rate of neutron emissions through {sup 241}Am ({alpha},n), which adds to the {sup 239}Pu/Be neutron source`s emission rate and maymore » lead to uncertainty in count rate measurements.« less

A Graphite Pile was built at Los Alamos National Laboratory in the early 1950`s to measure emission rates from neutron radiation sources using a BF{sub 3} proportional counter. The Pile was originally calibrated with a primary standard {sup 226}Ra/Be neutron source. The BF{sub 3} counter is inserted in the center of the Pile 138.5 cm above the ground surface, while neutron source position varies from 21 cm to 81 cm below the BF{sub 3} counter. In the past, it was assumed that neutrons emitted by a source would be totally thermalized by the time they reached the detector. Also, the detection efficiency was used for other neutron sources and the differences in energy spectra were not taken into account. Detailed Monte Carlo calculations were performed to determine Pile response functions for source-to-detector distances of 21 cm, 31 cm, 51 cm, and 81 cm. These response functions are then folded to energy spectra of ten different neutron sources: {sup 252}Cf, {sup 241}Am/F, {sup 241}Am/B, {sup 241}Am/Be, {sup 241}Am/Li, {sup 238}Pu/Li, {sup 239}Pu/Be, {sup 226}Ra/Be, Po/Be, and {sup 242}Cm/Be, whose neutron spectra are known. Results show that detection efficiencies are about 2.5 E-3 at the shortest distance (21 cm), and for differentmore » sources they can differ by 40% at that distance. Conversely, at the largest distance (81 cm), Pile response functions are relatively flat with respect to neutron energy. Moreover, detection efficiencies differ only by about 5%, but the absolute efficiency is about 2.0 E-4.« less

TLD material of CaF{sub 2} with various doping has been considered insensitive to neutron radiation for a long time. It is true for low energy neutrons. However neutron interaction cross section data of calcium and fluorine both show significant cross sections for (n, p), (n, {alpha}), and (n, {gamma}) reactions with neutron energy of 1 MeV and above. The Monte Carlo code MCNP was used to calculate CaF{sub 2} and LiF energy response functions for photons, electrons, and neutrons as incident radiations. The calculations were performed for a TLD free in the air and with a water phantom. In photon and electron cases, the energy deposition, referred to as the *F8 tally, was used. Detailed transport of both primary and secondary particles are included in the calculations. For neutrons, charged particle production and interaction with the TLD material, in form of a neutron heating tally, F6, is the main contribution of energy deposition. We also tallied the energy deposition due to capture gamma rays and secondary electrons. The results show that, for low energy neutrons, {sup 6}LiF, {sup 7}LiF, and CaF{sub 2}, responses differ in several order of magnitudes, but for neutron energy above 5 MeV, the responses are withinmore » a factor of two. CaF{sub 2} TLD chips show higher sensitivity to high energy neutrons than to high energy photons. LiF has higher sensitivity than CaF{sub 2} for electrons With energies less than 5 MeV and lower sensitivity for higher energy electrons. For photons, CaF{sub 2} has a higher sensitivity than LiF for energies above 25 keV.« less

In an improved Andersson-Braun (A-B) rem meter, Pb(n, xn) reactions at high neutron energies increase rem meter sensitivity when lead replaces a small part of the neutron-moderating material, polyethylene. Monte Carlo calculations were carried out using the-los Alamos High-Energy Transport code (LAHET) and Monte Carlo Neutron Transport (MCNP) code for both standard and lead-modified A-B rem meters with monoenergetic neutrons (up to 2.1 GeV). Experiments for medium and high energy tests (23.6 MeV-2.1 GeV at nine energy values) verified the results of theoretical calculations of response enhancement. Thus, Monte Carlo calculations that use the LAHET and MCNP codes prove to be powerful tools to simulate new neutron detectors.

An electron beam generated X-ray spectrum consists of characteristic X-rays of the target and continuous bremsstrahlung. The percentage of characteristic X-rays over the entire energy spectrum depends on the beam energy and the filter thickness. To determine the optimal electron beam energy and filter thickness, one can either conduct many experimental measurements, or perform a series of Monte Carlo simulations. Monte Carlo simulations are shown to be an efficient tool for determining the optimal target-filter system for electron beam generated X-ray spectra. Three of the most commonly used low-energy X-ray metal targets (Cu, Zn and Mo) are chosen for this study to illustrate the power of Monte Carlo simulations.

Cylindrical targets used for the production of radioactive beams by high-energy protons are analyzed in terms of target interaction length for some typical target systems. The analysis, using the Monte Carlo based LAHET code system, first determined applicable parameters of the beam/target interaction at 500 MeV incident proton energy. The analysis was then extended to other commonly-available proton energies up to 1000 MeV.

The investigators used Monte Carlo simulations to design a new high energy neutron dose equivalent meter or rem meter. This rem meter design uses new moderating materials and a He-3 detector that result in higher neutron detection efficiency, especially for higher energy neutrons. Response is also improved - the response curve of the new rem meter is a closer match to the fluence to dose curve.