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VALIDATION OF HANFORD PERSONNEL AND EXTREMITY DOSIMETERS IN PLUTONIUM ENVIRONMENTS

Description: A study was performed in the Plutonium Finishing Plant to assess the performance of Hanford personnel neutron dosimetry. The study was assessed whole body dosimetry and extremity dosimetry performance. For both parts of the study, the TEPC was used as the principle instrument for characterizing workplace neutron fields. In the whole body study, 12.7-cm-diameter TEPCs were used in ten different locations in the facility. TLD and TED personnel dosimeters were exposed on a water-filled phantom to enable a comparison of TEPC and dosimeter response. In the extremity study, 1.27-cm-diameter TEPCs were exposed inside the fingers of a gloveboxe glove. Extremity dosimeters were wrapped around the TEPCs. The glove was then exposed to six different cans of plutonium, simulating the exposure that a worker's fingers would receive in a glovebox. The comparison of TEPC-measured neutron dose equivalent to TLD-measured gamma dose equivalent provided neutron-to-gamma ratios that can be used to estimate the neutron dose equivalent received by a worker's finger based on the gamma readings of an extremity dosimeter. The study also utilized a Snoopy and detectors based on bubble technology for assessing neutron exposures, providing a comparison of the effectiveness of these instruments for workplace monitoring. The study concludes that the TLD component of the HCND performs adequately overall, with a positive bias of 30%, but exhibits excessive variability in individual results due to instabilities in the algorithm. The TED response was less variable but only 20% of the TEPC reference dose on average because of the low neutron energies involved. The neutron response of the HSD was more variable than the TLD component of the HCND and biased high by a factor of 8 overall due to its calibration to unmoderated 252Cf. The study recommends further work to correct instabilities in the HCND algorithm and to explore the potential ...
Date: February 10, 2000
Creator: Scherpelz, Robert I.; Fix, John J. & Rathbone, Bruce A.
Partner: UNT Libraries Government Documents Department

Comparisons of TORT and MCNP dose calculations for BNCT treatment planning

Description: The relative merit of using a deterministic code to calculate dose distributions for BNCT applications were examined. The TORT discrete deterministic ordinated code was used in comparison to MCNP4A to calculate dose distributions for BNCT applications
Date: December 31, 1996
Creator: Ingersol, D.T.; Slater, C.O.; Williams, L.R.; Redmond, E.L., II & Zamenhof, R.G.
Partner: UNT Libraries Government Documents Department

Neutron Exposure Parameters for the Dosimetry Capsule in the Heavy-Section Steel Irradiation Program Tenth Irradiation Series

Description: This report describes the computational methodology for the least-squares adjustment of the dosimetry data from the HSSI 10.OD dosimetry capsule with neutronics calculations. It presents exposure rates at each dosimetry location for the neutron fluence greater than 1.0 MeV, fluence greater than 0.1 MeV, and displacements per atom. Exposure parameter distributions are also described in terms of three- dimensional fitting functions. When fitting functions are used it is suggested that an uncertainty of 6% (1 o) should be associated with the exposure rate values. The specific activity of each dosimeter at the end of irradiation is listed in the Appendix.
Date: October 1, 1998
Creator: Baldwin, C.A.; Kam, F.B.K. & Remec, I.
Partner: UNT Libraries Government Documents Department

Microdosimetry of monoenergetic neutrons

Description: Tissue spheres 0.25, 0.5, 1.0, 2.0, 4.0, and 8.0 {mu}m in diameter were simulated using a wall-less spherical counter filled with a propane-based tissue-equivalent gas. Microdosimetric spectra corresponding to these site sizes were measured for five neutron energies (0.22, 0.44, 1.5, 6, and 14 MeV) and the related mean values {bar Y}{sub F} and {bar Y}{sub D} were calculated for several site sizes and neutron energies. An elaborate calibration technique combining soft x-rays, a {sup 55}Fe photon source, and a {sup 244}Cm collimated source of alpha particles was used throughout the measurement. The spectra and their mean values are compared with theoretically calculated values for ICRU tissue. The agreement between the calculated and the measured data is good in spite of a systematic discrepancy, which could be attributed, in part, to the difference in elemental composition between the tissue-equivalent gas and plastic used in the counter, and the ICRU standard tissue used in the calculations.
Date: December 31, 1993
Creator: Srdoc, D. & Marino, S. A.
Partner: UNT Libraries Government Documents Department

Fifth personnel dosimetry intercomparison study

Description: The fifth Personnel Dosimetry Intercomparison Study (PDIS) was conducted at the Oak Ridge National Laboratory's (ORNL) Dosimetry Applications Research (DOSAR) facility on March 20-22, 1979. This study is the latest PDIS in the continuing series started at the DOSAR facility in 1974. The PDIS is a three day study, typically in March, where personnel dosimeters are mailed to the DOSAR facility, exposed to a range of low-level neutron radiation doses (1 to 15 mSv or equivalently, 100 to 1500 mrem) and neutron-to-gamma ratios (1:1-10:1) using the Health Physics Research Reactor (HPRR) as the radiation source, and returned to the participants for evaluation. This report is a summary and analysis of the results reported by the various participants. The participants are able to intercompare their results with those of others who made dose measurements under identical experimental conditions.
Date: February 1, 1980
Creator: Sims, C.S.
Partner: UNT Libraries Government Documents Department

Neutron dosimetry of the HFIR hydraulic facility

Description: The total, fast, and thermal neutron fluxes at five axial positions in the High Flux Isotope Reactor (HFIR) hydraulic tube have been measured using bare and/or cadmium-covered activation, fission, and helium accumulation flux monitors. The spectrum-averaged, one-group cross sections over selected energy ranges for the reactions used in the measurements were obtained using cross sections from the ENDF/B-V file, and the target region volume-integrated spectrum was calculated with DORT, a two-dimensional discrete ordinates radiation transport code. The fluxes obtained from various monitors are in good agreement. The total and fast (>l MeV) neutron fluxes vary from 1.6 {times} 10{sup 19} n/m{sup 2} {center_dot} s and 1.6 {times} 10{sup 18} n/m{sup 2} {center_dot} s, respectively at the ends (HT-1 and -9) of the facility to 4.0 {times} 10{sup 19} n/m{sup 2} {center_dot} s and 4.6 {times} 10{sup 18} n/m{sup 2} {center_dot} s, respectively, at the center (HT-5) of the facility. The thermal-to-fast (>1 MeV) flux ratio varies from about 5.4 at the center to about 6.7 at the ends of the facility. The ratio of fast flux greater than 0.1 MeV to that greater than 1 MeV is 2.0 and stays almost constant along the length of the tube.
Date: February 1, 1995
Creator: Mahmood, S.T.; Mirzadeh, S.; Farrell, K.; Pace, J.V. III & Oliver, B.M.
Partner: UNT Libraries Government Documents Department

Moderation of neutron spectra

Description: Most of the accelerators that produce the various microenergetic neutron sources required for low-energy neutron dosimetry studies have been shut down. One alternative to accelerator-produced sources is the use of fission neutron or ({alpha},n) sources with unique neutron spectra. The problem with this solution is that maintenance of these sources is impractical. To help overcome this impracticality, the authors propose the use of moderating materials to produce a variety of spectra using a minimum number of sources. In the study, they performed Monte Carlo transport calculations under the following conditions: transporting neutrons from bare {sup 252}Cf or {sup 241}Am-Be sources from the center of various-sized spheres; tallying neutron spectra at 50 cm from the source. Of the twelve different moderating materials they studied, they found pure copper to be an ideal moderator. In this paper, they present flux-weighted energies, neutron spectra, and dose information for both {sup 252}Cf and {sup 241}Am-Be sources in bare and six-moderator configurations.
Date: May 1, 1997
Creator: Hsu, H.H. & Chen, J.
Partner: UNT Libraries Government Documents Department

Retrospective assessment of personnel neutron dosimetry for workers at the Hanford Site

Description: This report was prepared to examine the specific issue of the potential for unrecorded neutron dose for Hanford workers, particularly in comparison with the recorded whole body (neutron plus photon) dose. During the past several years, historical personnel dosimetry practices at Hanford have been documented in several technical reports. This documentation provides a detailed history of the technology, radiation fields, and administrative practices used to measure and record dose for Hanford workers. Importantly, documentation has been prepared by personnel whose collective experience spans nearly the entire history of Hanford operations beginning in the mid-1940s. Evaluations of selected Hanford radiation dose records have been conducted along with statistical profiles of the recorded dose data. The history of Hanford personnel dosimetry is complex, spanning substantial evolution in radiation protection technology, concepts, and standards. Epidemiologic assessments of Hanford worker mortality and radiation dose data were initiated in the early 1960s. In recent years, Hanford data have been included in combined analyses of worker cohorts from several Department of Energy (DOE) sites and from several countries through the International Agency for Research on Cancer (IARC). Hanford data have also been included in the DOE Comprehensive Epidemiologic Data Resource (CEDR). In the analysis of Hanford, and other site data, the question of comparability of recorded dose through time and across the respective sites has arisen. DOE formed a dosimetry working group composed of dosimetrists and epidemiologists to evaluate data and documentation requirements of CEDR. This working group included in its recommendations the high priority for documentation of site-specific radiation dosimetry practices used to measure and record worker dose by the respective DOE sites.
Date: September 1, 1996
Creator: Fix, J.J.; Wilson, R.H. & Baumgartner, W.B.
Partner: UNT Libraries Government Documents Department

Neutron interactions with biological tissue. Progress report, 1992--1993

Description: The first area of research focuses on track structure effects in neutron microdosimetry and nanodosimetry. This is an investigation of the effect of proton energy-loss straggling and the associated transport of energy by secondary electrons on neutron event-size distributions in small sites. Secondly, energy deposition spectra and their moments for fast neutrons are investigated. Using calculated charged particle initial spectra and slowing-down spectra the authors can calculate the energy deposition in spherical cavities. Lastly, an attempt was made to study the relation of neutron energy deposition calculations to biology and biophysical models.
Date: December 31, 1993
Partner: UNT Libraries Government Documents Department

Neutron dosimetry for low dose rate Cf-252 AT sources and adherence to recent clinical dosimetry protocol for brachytherapy

Description: In 1995, the American Association of Physicists in Medicine Task Group 43 (AAPM TG-43) published a protocol obsoleting all mixed-field radiation dosimetry for Cf-252. Recommendations for a new brachytherapy dosimetry formalism made by this Task Group favor quantification of source strength in terms of air kerma rather than apparent Curies or other radiation units. Additionally, representation of this dosimetry data in terms of radial dose functions, anisotropy functions, geometric factors, and dose rate constants are in an angular and radial (spherical) coordinate system as recommended, rather than the along-away dosimetry data (Cartesian coordinate system) currently available. This paper presents the initial results of calculated neutron dosimetry in a water phantom for a Cf-252 applicator tube (AT) type medical source soon available from Oak Ridge National Laboratory (ORNL).
Date: December 1, 1997
Creator: Rivard, M.J.; Wierzbicki, J.G.; Van den Heuvel, F. & Martin, R.C.
Partner: UNT Libraries Government Documents Department

Using response characteristics of neutron measurement devices to improve neutron dosimetry

Description: Recent administrative restrictions on personnel dose equivalent have resulted in increased pressure to more accurately report the neutron component without the traditional conservative added factors which sometimes inflate the reported values. Improvements include a new albedo neutron dosimeter which is capable of some limited energy discrimination. Also, additional emphasis has been placed on improving field measurements using traditional survey instrumentation and specialized spectroscopic techniques such as tissue equivalent proportional counters, Bonner spheres, and a modified 9 inches to 3 inches ratio technique. Improvements in these techniques along with a better understanding of the response of the TLD system have resulted in substantial reduction in the reported dose equivalent by improving the accuracy of the dosimeter system. The response characteristics of the TLD system and other instrumentation are obtained through modeling with the Monte Carlo code MCNP-4A. Neutron fields in work-areas are initially characterized with Bonner spheres. Routine updates are accomplished using a modified 9 inches to 3 inches ratio technique. These measurements are then used to predict the response of the TLD system when worn in that area. Correction curves are derived for the principal spectrum with various fractions of moderated or reflected neutrons. Work assignments are tracked through a database systems which is used to determine the principal spectrum that results in the neutron dose equivalent. The energy discrimination capability of the TLD system is used with the correction curve to derive an average correction appropriate to the readings of the dosimeter thus giving an energy corrected dose equivalent for the individual.
Date: December 1, 1995
Creator: Casson, W.H.; Hsu, H.H. & Hoffman, J.M.
Partner: UNT Libraries Government Documents Department