W.D. Reece

A study of the angular dependence problem in effective dose equivalent assessment.

The newly revised American National Standard N13.11 (1993) includes measurements of angular response as part of personnel dosimeter performance testing. However, data on effective dose equivalent (HE), the principle limiting quantity defined in International Commission on Radiological Protection (ICRP) Publication 26 and later adopted by U.S. Nuclear Regulatory Commission (NRC), for radiation incident on the body from off-normal angles are little seen in the literature. The absence of scientific data has led to unnecessarily conservative approaches in radiation protection practices. This paper presents a new set of fluence-to-HE conversion factors as a function of radiation angles and sex for monoenergetic photon beams of 0.08, 0.3, and 1.0 MeV. A Monte Carlo transport code (MCNP) and sex-specific anthropomorphic phantoms were used in this study. Results indicate that Anterior-posterior (AP) exposure produces the highest HE per unit photon fluence in all cases. Posterior-anterior (PA) exposure produces the highest HE among beams incident from the rear half-plane of the body. HE decreases dramatically as one departs from the AP and PA orientations. The results also indicate that overestimations caused by using isotropic dosimeters in assessing effective dose equivalent from near-overhead and near-underfoot exposures are 550%, 390%, and 254% for 0.08, 0.3, and 1.0 MeV, respectively. Comparisons of the angular dependence of HE with those based on the secondary quantities defined in International Commission on Radiation Units and Measurements (ICRU) Reports 39, 43, and 47 show significant differences. This paper discusses why more accurate assessments of HE are necessary and possible. An empirical equation is proposed which can be used as the optimum dosimeter angular response function for radiation angles ranging from 0 degrees to 90 degrees for dosimeter calibration, performance testing, and design.

Monte Carlo calculations of the dose backscatter factor for monoenergetic electrons.

Recently, there has been growing interest in beta emitters for therapeutic uses, especially in connection with so-called endovascular (or intravascular) brachytherapy. Since accurate dose estimation is necessary for the success of such applications, some problems in beta-ray dosimetry need further study. Among these problems, we have investigated the effect of electron backscattering on dose, which has significance not only for accurate dose estimation but also for new source design. In this study, an empirical measure of electron backscattering, known as the dose backscatter factor, was calculated using EGS4 Monte Carlo calculations for monoenergetic electrons and various scattering materials. Electron energies were 0.1, 0.5, 1.0, 2.0 and 3.0 MeV in combination with Al (Z = 13), Ti (Z = 22), Sr (Z = 38), Ag (Z = 47) and Pt (Z = 78) scatterers. The dose backscatter factor ranged from 10% to 60%, depending on electron energy and material, and was found to increase with the atomic number Z by a log(Z + 1) relationship. A method is presented for calculating the beta-ray dose backscatter factors using the results of this study. To demonstrate the efficacy of this method, a dose backscatter factor depth profile for 32P near a water/aluminium interface was calculated and these calculated results were found to generally reproduce the depth profile obtained from direct EGS4 calculations using the 32P spectrum. The data presented in this study can be used to calculate dose backscatter factors for any combination of beta emitter/scatterer whose atomic number ranges from 13 to 78.

Sex-specific tissue weighting factors for effective dose equivalent calculations

The effective dose equivalent was defined in the International Commission on Radiological Protection Publication 26 in 1977 and later adopted by the U.S. Nuclear Regulatory Commission. To calculate organ doses and effective dose equivalent for external exposures using Monte Carlo simulations, sex-specific anthropomorphic phantoms and sex-specific weighting factors are always employed. This paper presents detailed mathematical derivation of a set of sex-specific tissue weighting factors and the conditions which the weighting factors must satisfy. Results of effective dose equivalent calculations using female and male phantoms exposed to monoenergetic photon beams of 0.08, 0.3, and 1.0 MeV are provided and compared with results published by other authors using different sex-specific weighting factors and phantoms. The results indicate that females always receive higher effective dose equivalent than males for the photon energies and geometries considered and that some published data may be wrong due to mistakes in deriving the sex-specific weighting factors.The effective dose equivalent was defined in the International Commission on Radiological Protection Publication 26 in 1977 and later adopted by the U.S. Nuclear Regulatory Commission. To calculate organ doses and effective dose equivalent for external exposures using Monte Carlo simulations, sex-specific anthropomorphic phantoms and sex-specific weighting factors are always employed. This paper presents detailed mathematical derivation of a set of sex-specific tissue weighting factors and the conditions which the weighting factors must satisfy. Results of effective dose equivalent calculations using female and male phantoms exposed to monoenergetic photon beams of 0.08, 0.3, and 1.0 MeV are provided and compared with results published by other authors using different sex-specific weighting factors and phantoms. The results indicate that females always receive higher effective dose equivalent than males for the photon energies and geometries considered and that some published data may be wrong due to mistakes in deriving the sex-specific weighting factors.

Calculation of effective doses for broad parallel photon beams.

Values of effective dose (E) were calculated for the entire range of incident directions of broad parallel photon beams for selected photon energies using the Monte Carlo N-Particle (MCNP) transport code with a hermaphroditic phantom. The calculated results are presented in terms of conversion coefficients transforming air kerma to effective dose. This study also compared the numerical values of E and H(E) over the entire range of incident beam directions. E was always less than H(E) considering all beam directions and photon energies, but the differences were not significant except when a photon beam approaches some specific directions (overhead and underfoot). This result suggests that the current H(E) values can be directly interpreted as E or, at least, as a conservative value of E without knowing the details of irradiation geometries. Finally, based on the distributions of H(E) and E over the beam directions, this study proposes ideal angular response factors for personal dosimeters that can be used to improve the angular response properties of personal dosimeters for off-normal incident photons.

Development of new two-dosimeter algorithm for effective dose in ICRP Publication 103.

The two-dosimeter method, which employs one dosimeter on the chest and the other on the back, determines the effective dose with sufficient accuracy for complex or unknown irradiation geometries. The two-dosimeter method, with a suitable algorithm, neither significantly overestimates (in most cases) nor seriously underestimates the effective dose, not even for extreme exposure geometries. Recently, however, the definition of the effective dose itself was changed in ICRP Publication 103; that is, the organ and tissue configuration employed in calculations of effective dose, along with the related tissue weighting factors, was significantly modified. In the present study, therefore, a two-dosimeter algorithm was developed for the new ICRP 103 definition of effective dose. To that end, first, effective doses and personal dosimeter responses were calculated using the ICRP reference phantoms and the MCNPX code for many incident beam directions. Next, a systematic analysis of the calculated values was performed to determine an optimal algorithm. Finally, the developed algorithm was tested by applying it to beam irradiation geometries specifically selected as extreme exposure geometries, and the results were compared with those for the previous algorithm that had been developed for the effective dose given in ICRP Publication 60.

MONTE CARLO MODELING OF THE TEXAS A&M UNIVERSITY RESEARCH REACTOR

Abstract The Monte Carlo N-Particle (MCNP) code and a set of high-temperature neutron cross-section data were used to develop an accurate three-dimensional computational model of the Texas A&M University Nuclear Science Center Reactor (NSCR) at full power. The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements and control rods. The most significant approximation was made for entrained fission products because of the lack of knowledge of fission product inventory in the current reactor core. This study used the concept of “average fission product” to model the fission product in the reactor core and determined the concentration of the average fission product by repeating criticality calculations to make the reactor critical for a given critical condition. Finally, the developed model was tested by comparing the calculated results with those of other approaches, i.e., (a) an in-house three-dimensional diffusion code and (b) foil activation measurement. The developed reactor model showed a good agreement with these approaches. The developed model predicted the thermal neutron flux in samples within 11% of difference when compared with the results from the diffusion code and predicted the production of 198Au and 60Co within ~20% of difference when compared with the values measured with foils. The developed model also calculated the neutron energy spectrum very consistently with the other approaches for the entire energy range considered in this study.

Improved estimates of effective dose equivalent using two optimal anisotropic responding dosimeters.

Although the use of two dosimeters, one on the chest and the other on the back, successfully solved the underestimation problem for posterior incident photon beams, the two-dosimeter approach still has some problems-significant overestimations for lateral, overhead, and underfoot beam directions when isotropic-responding dosimeters are used for measurement. A solution to this problem is to intentionally construct the dosimeters to under-respond as the beam direction departs from normal incidence and approaches lateral, overhead, or underfoot beam directions. The objective of this study is to develop a dosimeter that does not significantly overestimate effective dose equivalent (H(E)) for lateral, overhead, and underfoot beam directions, while maintaining good performance for anterior and posterior beam directions. Several dosimeter geometries were investigated using Monte Carlo simulation to find the best geometry using aluminum oxide (Al2O3) as dosimeter attenuator material. Then, the developed Optimal Anisotropic Responding dosimeters were tested for 0.08, 0.30 and 1.00 MeV photon beams of various beam directions. The dosimeters did not overestimate H(E) by more than 80% considering all photon energies and beam directions, which is much less than the overestimation of isotropic-responding dosimeters (202%). The dosimeters also showed similar performance compared to isotropic-responding dosimeters for anterior and posterior beam directions. Finally, the dosimeters were applied to effective dose (E) and the results are compared with those of H(E).

Monte Carlo simulation of single and two-dosimeter approaches in a steam generator channel head.

In a steam generator channel head, it was not unusual to see radiation workers wearing as many as twelve dosimeters over the surface of the body to avoid a possible underestimation of effective dose equivalent (HE) or effective dose (E). This study shows that only one or two dosimeters can be used to estimate HE and E without a significant underestimation. MCNP and a point-kernel approach were used to model various exposure situations in a steam generator channel head. The single-dosimeter approach (on the chest) was found to underestimate HE and E significantly for a few exposure situations, i.e., when the major portion of radiation source is located in the backside of a radiation worker. In this case, the photons from the source pass through the body and are attenuated before reaching the dosimeter on the chest. To assure that a single dosimeter provides a good estimate of worker dose, these few exposure situations cannot dominate a worker’s exposure. On the other hand, the two-dosimeter approach (on the chest and back) predicts HE and E very well, hardly ever underestimating these quantities by more than 4% considering all worker positions and contamination situations in a steam generator channel head. This study shows that two dosimeters are adequate for an accurate estimation of HE and E in a steam generator channel head.

Calculation of the dose distribution in water from 71Ge K-shell x-rays

The dose distribution in water from 71Ge K-shell x-rays (Eave = 9.44 keV) was calculated for various source configurations using both analytic and EGS4 Monte Carlo calculations. The point source kernel and the buildup factor are presented. The buildup factor for a point source in water has been found to increase up to about 1.1 as radial distance approaches 1 cm. Comparison between 71Ge and 90Sr/Y shows a similarity between their relative dose distribution in water. The dose distribution from a disc source was calculated using the EGS4 code and compared with the results from analytic calculation. Excellent agreement was observed, confirming the validity of analytic calculations. The dose rate at 0.01 cm from a 71Ge disc source was calculated to be about 1.3 x 10(-5) Gy MBq-1 s-1. Based on the results from this study, 71Ge activity of the order of 3.7 x 10(10) Bq (approximately 1 Ci) might be necessary to obtain dose rates typical of 90Sr/Y ophthalmic applicators. The possibility of using 71Ge as a source of radioactive stents was also investigated. A 71Ge stent was modelled as a cylindrical shell source and the dose rates were determined by Monte Carlo calculations. Some calculated results are compared with published values for a 32P-coated stent. The dose rate at 0.01 cm from a 71Ge stent has been calculated to be about 6.5 x 10(-3) Gy MBq-1 h-1, which is much lower than the reported dose rate at the same distance from a 32P-coated stent. However, an initial source activity of the order of 3.7 x 10(7) Bq (approximately 1 mCi) would easily result in a typical target dose (approximately 24 Gy) needed for intravascular stent applications. In conclusion, 71Ge sources could be used as alternatives to beta sources and, unlike high-energy (approximately MeV) beta sources, may provide easily predictable dose distributions in heterogeneous media and low dose rates, which might be beneficial for some clinical applications.