Study of microdosimetric energy deposition patterns in tissue-equivalent medium due to low-energy neutron fields using a graphite-walled proportional counter

To improve radiation protection dosimetry for low-energy neutron fields encountered in nuclear power reactor environments, there is increasing interest in modeling neutron energy deposition in metrological instruments such as tissue-equivalent proportional counters (TEPCs). Along with these computational developments, there is also a need for experimental data with which to benchmark and test the results obtained from the modeling methods developed. The experimental work described in this paper is a study of the energy deposition in tissue-equivalent (TE) medium using an in-house built graphite-walled proportional counter (GPC) filled with TE gas. The GPC is a simple model of a standard TEPC because the response of the counter at these energies is almost entirely due to the neutron interactions in the sensitive volume of the counter. Energy deposition in tissue spheres of diameter 1, 2, 4 and 8 μm was measured in low-energy neutron fields below 500 keV. We have observed a continuously increasing trend in microdosimetric averages with an increase in neutron energy. The values of these averages decrease as we increase the simulated diameter at a given neutron energy. A similar trend for these microdosimetric averages has been observed for standard TEPCs and the Rossi-type, TE, spherical wall-less counter filled with propane-based TE gas in the same energy range. This implies that at the microdosimetric level, in the neutron energy range we employed in this study, the pattern of average energy deposited by starter and insider proton recoil events in the gas is similar to those generated cumulatively by crosser and stopper events originating from the counter wall plus starter and insider recoil events originating in the sensitive volume of a TEPC.     

The response of tissue-equivalent proportional counters to heavy ions

The paper presents a theoretical model for the response of a tissue-equivalent proportional counter (TEPC) irradiated with charged particles. Heavy ions and iron ions in particular constitute a significant part of radiation in space. TEPCs are used for all space shuttle and International Space Station (ISS) missions to estimate the dose and radiation quality (in terms of lineal energy) inside spacecraft. The response of the tissue-equivalent proportional counters shows distortions at the wall/cavity interface. In this paper, we present microdosimetric investigation using Monte Carlo track structure calculations to simulate the response of a TEPC to charged particles of various LET (1 MeV protons, 2.4 MeV alpha particles, 46 MeV/nucleon 20Ne, 55 MeV/nucleon 20Ne, 45 MeV/nucleon 40Ar, and 1.05 GeV/nucleon 56Fe). Data are presented for energy lost and energy absorbed in the counter cavity and wall. The model calculations are in good agreement with the results of Rademacher et al. (Radiat. Res. 149, 387-389, 1998), including the study of the interface between the wall and the sensitive region of the counter. It is shown that the anomalous response observed at large event sizes in the experiment is due to an enhanced entry of secondary electrons from the wall into the gas cavity.     

Microdosimetric measurements in the secondary radiation field produced in 12C-therapy irradiations

The ambient dose equivalent from the secondary radiation produced during irradiation of a cylindrical water phantom with 200 MeV/u ¹²C-ions was investigated at the biophysics cave at GSI Helmholtzzentrum für Schwerionenforschung in Darmstadt, Germany. Pencil-like ion beams were delivered by the heavy-ion synchrotron SIS18 using the slow extraction mode. Since the secondary radiation field outside the phantom is complex in its particle composition and particle energy distribution, microdosimetric methods developed for the dosimetry of the cosmic radiation field at flight altitudes, which is similar in terms of complexity, were applied. Lineal energy distributions and the ambient dose equivalent were measured with a tissue-equivalent proportional counter at different particle emission angles. An additional veto counter allowed the identification of the different contributions of charged and neutral particles. A significant increase in the mean quality factor was observed at large emission angles which could be attributed to the decreasing contributions of charged particles compared to the (relative) contributions from neutrons.     

Microdosimetry for boron neutron capture therapy.

Preclinical studies for boron neutron capture therapy (BNCT) using epithermal neutrons are ongoing at several laboratories. The absorbed dose in tumor cells is a function of the thermal neutron flux at depth, the microscopic boron concentration, and the size of the cell. Dosimetry is therefore complicated by the admixture of thermal, epithermal, and fast neutrons, plus gamma rays, and the array of secondary high-linear-energy-transfer particles produced within the patient from neutron interactions. Microdosimetry can be a viable technique for determining absorbed dose and radiation quality. A 2.5-cm-diameter tissue-equivalent gas proportional counter has been built with 50 parts per million (ppm) 10B incorporated into the walls and counting gas to simulate the boron uptake anticipated in tumors. Measurements of lineal energy (y) spectra for BNCT in simulated volumes of 1-10 microns diameter show a dose enhancement factor of 4.3 for 30 ppm boron, and a "y" of 250 keV/microns for the boron capture process. Chamber design plus details of experimental and calculated linear energy spectra will be presented.     

FLUKA capabilities for microdosimetric analysis

Delta-ray transport is important in microdosimetric studies, and how Monte Carlo models handle delta electrons using condensed histories is important for accurate simulation. The purpose of this study was to determine how well FLUKA can simulate energy deposition spectra in a tissue-equivalent proportional counter (TEPC) and produce a reliable estimate of delta-ray events produced when a TEPC is exposed to high-energy heavy ions (HZE) like those in the galactic cosmic-ray (GCR) environment. A 1.27-cm spherical TEPC with a low-pressure gas simulating a 1-μm site, typical of the one flown on the ISS, was constructed in FLUKA, and its response was compared to experimental data for an (56)Fe-ion beam at 360 MeV/nucleon. Several narrow beams at different impact parameters were used to explain the response of the same detector exposed to a uniform field of radiation. Additionally, the effect that wall thickness had on the response of the TEPC and the range of delta rays in the tissue-equivalent (TE) wall material was investigated, and FLUKA produced the expected wall effect for primary particles passing outside the sensitive volume. A final comparison to experimental data was made for the simulated TEPCs exposed to various broad beams in the energy range of 200-1000 MeV/nucleon. FLUKA overestimated energy deposition in the gas volume in all cases. The FLUKA results differed from the experimental data by an average of 25.2% for y(F) and 12.4% for y(D). It is suggested that this difference can be reduced by adjusting the FLUKA default ionization potential and density correction factors. Accurate transport codes are desirable because of the high cost of beam time for experimental evaluation of energy deposition spectra produced by HZE ions and the flexibility that calculations offer in the TEPC engineering and design process.     

Solid cancer risk coefficient for fast neutrons in terms of effective dose

Cancer mortality risk coefficients for neutrons have recently been assessed by a procedure that postulates for the neutrons a linear dose dependence, invokes the excess risk of the A-bomb survivors at a gamma-ray dose D(1) of 1 Gy, and assumes a neutron RBE as a function of D(1) between 20 and 50. The excess relative risk (ERR) of 0.008/mGy has been obtained for R(1) = 20 and 0.016/mGy for R(1) = 50. To compare these results to the current ICRP nominal risk coefficient for solid cancer mortality (0.045/Sv for a population of all ages; 0.036/Sv for a working population), the ERR is translated into lifetime attributable risk and is then related to effective dose. The conversion is not trivial, because the neutron effective dose has been defined by ICRP not as a weighted genuine neutron dose (neutron kerma), but as a weighted dose that includes the dose from gamma rays that are induced by neutrons in the body. If this is accounted for, the solid cancer mortality risk for a working population is found to agree with the ICRP nominal risk coefficient for neutrons in their most effective energy range, 0.2 MeV to 0.5 MeV. In radiation protection practice, there is an added level of safety, because the effective dose, E, is-for monitoring purposes-assessed in terms of the operational quantity H*, which overestimates E substantially for neutrons between 0.01 MeV and 2 MeV.     

Radiation-related posterior lenticular opacities in Hiroshima and Nagasaki atomic bomb survivors based on the DS86 dosimetry system

This paper investigates the quantitative relationship of ionizing radiation to the occurrence of posterior lenticular opacities among the survivors of the atomic bombings of Hiroshima and Nagasaki suggested by the DS86 dosimetry system. DS86 doses are available for 1983 (93.4%) of the 2124 atomic bomb survivors analyzed in 1982. The DS86 kerma neutron component for Hiroshima survivors is much smaller than its comparable T65DR component, but still 4.2-fold higher (0.38 Gy at 6 Gy) than that in Nagasaki (0.09 Gy at 6 Gy). Thus, if the eye is especially sensitive to neutrons, there may yet be some useful information on their effects, particularly in Hiroshima. The dose-response relationship has been evaluated as a function of the separately estimated gamma-ray and neutron doses. Among several different dose-response models without and with two thresholds, we have selected as the best model the one with the smallest x2 or the largest log likelihood value associated with the goodness of fit. The best fit is a linear gamma-linear neutron relationship which assumes different thresholds for the two types of radiation. Both gamma and neutron regression coefficients for the best fitting model are positive and highly significant for the estimated DS86 eye organ dose.     

The effect of changes in dosimetry on cancer mortality risk estimates in the atomic bomb survivors

In the spring of 1986 the Radiation Effects Research Foundation (RERF) received a new atomic bomb dosimetry system. This report presents the comparisons of leukemia and nonleukemia cancer mortality risk estimates under the old and new dosimetries. In terms of total kerma (essentially whole-body gamma plus neutron exposure), risk estimates for both classes of cancer are 75-85% higher with the new dosimetry. This and other summary comparisons allow for possible nonlinearity at high estimated doses. Changes are also considered in relation to organ doses and assumptions about the relative biological effectiveness (RBE) of neutrons. Without regard to RBE, the risk estimates for total organ dose are essentially unchanged by the dosimetry revision. However, with increasing assumed values of RBE, the estimated low-LET risk decreases much less rapidly under the new dosimetry, due to the smaller neutron component. Thus at an assumed constant RBE of 10, for example, the effect of the dosimetry revision is to increase organ dose risk estimates, relative to those based on the old dosimetry, by 30% for nonleukemia and 80% for leukemia. At an RBE of 20 these increases are 72 and 136%, respectively. A number of other issues are discussed. The city difference in dose is no longer statistically significant, even at an RBE of one. Estimation of RBE is even less feasible with new dosimetry. There is substantial question of the linearity in dose response, in the sense of a leveling off at higher doses. Finally, some indication is given of how risks estimated from this dosimetry and the current data may compare to widely used estimates based largely on the RERF data with the previous dosimetry.     

Neutron scattering and energy deposition spectra

Summary The influence of neutron scattering in the wall of a spherical proportional counter on the energy deposition spectra and the absorbed dose is investigated. Event probabilities, and frequency and dose-averaged deposited energies are calculated with and without scattering contribution and compared. The change of absorbed dose due to attenuation of the primary neutron flux is also evaluated.Contribution No. 1157 of the Biology Programme, Directorate General XII of the Commission of the European Communities.     

Calorimetric measurement of the carbon kerma factor for 14.6-MeV neutrons

A spherical graphite calorimeter was used to determine the ratio of kerma to influence (kerma factor) for neutrons whose mean energy was approximately 14.6 MeV. The calorimeter was used to measure carbon kerma, while activation foils of Al and Au were used to determine the neutron fluence. The calorimeter was constructed specifically to measure kerma in neutron fields. The amount of graphite and other materials was kept to a minimum to reduce absorption and scattering of the neutrons. Ionization chambers were used to measure A-150 plastic kerma and to monitor the intensity of the exposures. The value for the carbon kerma factor was determined to be 1.80 +/- 0.16 X 10(-11) Gy X cm2. The relationship of this value to other recent measurements and calculations at similar neutron energies is discussed.     

Derivation of radiation quality average parameters in neutron-gamma radiation fields with the high-pressure ionization chamber: theory and practice

The established radiation quality parameters in mixed neutron-gamma radiation fields may be measured by applying the initial (columnar) recombination of ions in tissue-equivalent (TE) high-pressure ionization chambers (recombination chambers). The mean quality factor can be determined to within 10-15% for mixed fields with neutrons ranging from thermal to 10 MeV, and the dose mean LET of the proton component can be determined to within 10-15% if the gamma-ray absorbed dose fraction is known. These average parameters are derived by measuring the ratio of the ionization currents collected at two high-field strengths and constant gas pressure applied to the ionization chamber. By utilizing approximate correlations between physical parameters in the neutron energy region from thermal to 10 MeV, the dose mean LET of the heavy ion component, the overall dose mean LET, and the microdosimetric parameter y0,D of the mixed field can also be derived. Experimental verification of the method is presented for various neutron-gamma radiation spectra in air and in water by comparison to theoretical calculations and results from low-pressure proportional counter measurements. Good agreement is shown. The TE high-pressure ionization chamber appears to have wide potential for use as a dose-equivalent meter in radiation protection or as a beam characterization device in radiobiology.     

Neutron versus Á-ray risk estimates

While it is recognized that neutrons contributed to the excess cancer incidence and mortality among the atomic bomb survivors in Hiroshima, there is no possibility to deduce the magnitude of this contribution from the data. This remains true even if the neutron doses in the dosimetry system DS86 are corrected upwards in line with recent neutron activation measurements. In spite of this fact, important information can be obtained in the form of an inverse relation of the risk coefficients for γ-rays and neutrons. Such an interrelation must apply because the observed excess incidence or mortality is made up of a γ-ray and a neutron component; increased attribution to neutrons decreases the attribution to photons. Computations with the uncorrected and the corrected DS86 are performed for the mortality and the incidence of solid tumors combined. They refer to doses up to 2 Gy and employ the constant relative risk model and a linear-quadratic dose dependence with variable ratio – the neutron relative biological effectiveness (RBE) at low doses – of the linear component for neutrons and γ-rays. In line with past analyses, no quadratic component is obtained with the uncorrected DS86, but it is seen, even in these calculations, that the assumption of increased neutron RBEs does not translate into proportional increases of the risk coefficients of neutrons, because it leads to substantially reduced risk estimates for γ-rays. Calculations with the corrected dosimetry bring out this reciprocity even more clearly. High values of the neutron RBE reduce – in line with recent suggestions by Rossi and Zaider – the risk estimates for γ-rays substantially. Even a purely quadratic dose relation for γ-rays is consistent with the data; it requires no major increase of the nominal risk coefficients for neutrons over the currently assumed values. The cancer data from Hiroshima can still provide `prudent' risk estimates for photons, but with the corrected DS86, they do not prove that there is a linear component in the dose dependence for photons. Received: 20 January 1997 / Accepted in revised form: 14 March 1997     

Micronucleus induction in Vicia faba roots. Part 2. Biological effects of neutrons below 1 cGy

A dose-effect relationship has been established for high-energy neutrons (maximum energy 600 MeV) within a dose range of 0.2 to 80 cGy and for low-energy neutrons produced by a 252Cf source (mean energy 2.35 MeV) for doses between 0.2 and 5 cGy. The frequency of micronuclei was found to increase linearly with dose. The relative biological effectiveness (r.b.e) values calculated using 60Co radiation as a reference were, in the high-dose region, 4.7 +/- 0.4 and 11.8 +/- 1.3 for the high- and low-energy neutrons, respectively. At doses below 1 cGy constant values of 25.4 +/- 4.4 and 63.7 +/- 12 were reached for the respective neutron energies.     

Influence of neutron collimation on the energy deposition in a tissue equivalent sphere

Summary The influence of neutron collimation on the shape of energy deposition spectra was investigated using a spherical walled proportional counter. Experimental dose averaged lineal energies were obtained and compared with theoretical values. The fractions of total absorbed dose corresponding to various intervals of lineal energies were also deduced from the measured distributions.Contribution No. 1289 of the Biology Programme, Directorate General XII of the Commission of the European Communities     

Recoil proton, alpha particle, and heavy ion impacts on microdosimetry and RBE of fast neutrons: analysis of kerma spectra calculated by Monte Carlo simulation

Fast neutrons (FN) have a higher radio-biological effectiveness (RBE) compared with photons, however the mechanism of this increase remains a controversial issue. RBE variations are seen among various FN facilities and at the same facility when different tissue depths or thicknesses of hardening filters are used. These variations lead to uncertainties in dose reporting as well as in the comparisons of clinical results. Besides radiobiology and microdosimetry, another powerful method for the characterization of FN beams is the calculation of total proton and heavy ion kerma spectra. FLUKA and MCNP Monte Carlo code were used to simulate these kerma spectra following a set of microdosimetry measurements performed at the National Accelerator Centre. The calculated spectra confirmed major classical statements: RBE increase is linked to both slow energy protons and alpha particles yielded by (n,alpha) reactions on carbon and oxygen nuclei. The slow energy protons are produced by neutrons having an energy between 10 keV and 10 MeV, while the alpha particles are produced by neutrons having an energy between 10 keV and 15 MeV. Looking at the heavy ion kerma from <15 MeV and the proton kerma from neutrons <10 MeV, it is possible to anticipate y* and RBE trends.     

Risk estimation for fast neutrons with regard to solid cancer.

In the absence of epidemiological information on the effects of neutrons, their cancer mortality risk coefficient is currently taken as the product of two low-dose extrapolations: the nominal risk coefficient for photons and the presumed maximum relative biological effectiveness of neutrons. This approach is unnecessary. Since linearity in dose is assumed for neutrons at low to moderate effect levels, the risk coefficient can be derived in terms of the excess risk from epidemiological observations at an intermediate dose of gamma rays and an assumed value, R(1), of the neutron RBE relative to this reference dose of gamma rays. Application of this procedure to the A-bomb data requires accounting for the effect of the neutron dose component, which, according to the current dosimetry system, DS86, amounts on average to 11 mGy in the two cities at a total dose of 1 Gy. With R(1) tentatively set to 20 or 50, it is concluded that the neutrons have caused 18% or 35%, respectively, of the total effect at 1 Gy. The excess relative risk (ERR) for neutrons then lies between 8 per Gy and 16 per Gy. Translating these values into risk coefficients in terms of the effective dose, E, requires accounting for the gamma-ray component produced by the neutron field in the human body, which will require a separate analysis. The risk estimate for neutrons will remain essentially unaffected by the current reassessment of the neutron doses in Hiroshima, because the doses are unlikely to change much at the reference dose of 1 Gy.     

Measurements of fast neutrons in Hiroshima by use of 39Ar

The survivors of the A-bomb explosions over Hiroshima and Nagasaki were exposed to a mixed neutron and gamma radiation field. To validate the high-energy portion of the neutron field and thus the neutron dose to the survivors, a method is described that allows retrospective assessment of the fast neutrons from the A-bombs. This is accomplished by the extraction of the noble gas argon from biotites separated from Hiroshima granite samples, and then the detection of the (39)Ar activity that was produced by the capture of the fast neutrons on potassium. Adjusted to the year 1945, activities measured in the first samples taken at distances of 94, 818, 992, and 1,173 m from the hypocenter were 6.9+/-0.2, 0.32+/-0.01, 0.14+/-0.02, and 0.09+/-0.01 mBq/g K, respectively. All signals were significantly above detector background and show low uncertainties. Considering their uncertainties they agree with the calculated (39)Ar activation in the samples, based on the most recent dosimetry system DS02. It is concluded that this method can be used to investigate samples obtained from large distances in Hiroshima, where previous data on fast neutrons are characterized by considerable uncertainties. Additionally, the method can be used to reconstruct the fast neutron fluence in Nagasaki, where no experimental data exist.     

Hiroshima-like neutrons from A-bomb replica: physical basis for their use in biological experiments

The lineal energy distribution and several other dosimetric parameters were measured for the neutrons emitted from a replica of the Hiroshima bomb to determine their usefulness in biological experiments designed to estimate the effectiveness of actual Hiroshima neutrons. The "Little-Boy" replica (LBR) was constructed at the Los Alamos National Laboratory in support of the recent atomic-bomb dose reevaluation and was made of identical materials and had nearly identical dimensions and geometry as the Hiroshima bomb. However, the LBR was operated as a steady-state nuclear reactor, which permitted measurements under controlled conditions. Detailed dosimetric measurements and calculations were made at distances of up to 2.1 m from the center of the LBR uranium core. At these distances, the in-air kerma was approximately 97% from neutrons and kerma rates were shown to be particularly useful for biological experiments (up to approximately 7 Gy/h was possible). Quantitative intercomparisons of neutron energy spectra, lineal energy distributions, and measured cytogenetic results for several fission-neutron sources indicate that Hiroshima and LBR neutrons should be of similar biological effectiveness. Based on these evaluations, and cytogenetic results for LBR neutrons reported in a companion paper (this issue), it is estimated that Hiroshima neutrons were 20 to 30% more effective than the fission neutrons commonly used in radiobiology.     

Lung carcinomas in Sprague-Dawley rats after exposure to low doses of radon daughters, fission neutrons, or gamma rays

The effectiveness of radon-daughter inhalation and irradiation with fission neutrons and gamma rays in the induction of lung carcinomas in Sprague-Dawley rats at low doses is compared. Earlier reports which compared radon-daughter inhalations and neutron irradiations over a wider range of doses were based on dosimetry for the radon-daughter inhalations which has recently been found to be faulty. In the present analysis, low-dose experiments were designed to derive revised equivalence ratios between radon-daughter exposures, and fission neutron or gamma irradiations. The equivalence is approximately 15 working level months (WLM) of radon daughters to 10 mGy of neutrons (the earlier value was 30 WLM to 10 mGy). The relative biological effectiveness (RBE) of neutrons is 50 or more at a gamma-ray dose of 1 Gy. In these experiments with low doses and exposures, the lifetime incidences can be estimated from the raw incidences, while the derivation of the time dependence of the prevalence is essential for the estimation of RBE values and equivalence ratios.     

Effects of cross-section structure on the dosimetric response functions for 0.4 to 10.0 MeV neutrons in the ICRU tissue sphere

The effect of the fluctuating cross-section structure in the energy range of 0.4 to 10.0 MeV on the dosimetric response functions of neutrons in the ICRU standard tissue sphere is analyzed. A Monte Carlo method with point-energy cross-section values, including coupled transport for neutrons and secondary charged particles, was used in the direct estimation of the absorbed dose and the dose equivalent. An approach was adopted in which source-energy band-average responses were calculated instead of the more usual approach involving monoenergetic source neutrons. Data were obtained for the newly defined term, ambient dose equivalent, at various depths, as well as the older index quantities. Such data generated were compared with information from other research workers. In general, good agreement was found, with due consideration to the differences engendered by the use of the source-energy band-average approach. Agreement was poorest for very shallow depths, corresponding to outer skin thickness, this being a most difficult depth to calculate accurately. The dosimetric data generated in this study should contribute to the ongoing efforts for the standardization of neutron protection dosimetry.