Neutron spectra and dosimetric features around an 18 mv linac accelerator

Using the difference between responses to neutrons of TLD-600 and TLD-700, three experimental devices were constructed and arranged to measure thermal neutron fluences, neutron spectra, and neutron doses inside the treatment room of a radiotherapy 18 MV Linear electron accelerator (Linac). Thermal neutron fluences were measured with TLD-600/TLD-700 pairs arranged in both a bare and a cadmium (Cd) foil covered methacrylate box. Neutron spectra were measured in 26 energy bins by introducing pairs of TLD-600/TLD-700 in air and into the middle of five polyethylene spheres with diameters of 3, 5, 8, 10, and 12 inches. A PC version of the BUNKI code was used to unfold the six measurements in each sphere to obtain the 26 energy bins. Neutron and photon doses were measured by introducing pairs of TLD-600/TLD-700 into the middle of a single 25-cm-diameter paraffin sphere. The three required neutron calibrations were carried out at the Nuclear Technology Laboratory of the Polytechnique University of Madrid (UPM), using an 241Am-Be neutron source with an alpha activity of 111 GBq and a yield of 6.6 x 10(6) neutrons s(-1). Three devices were needed for the necessary calibrations: a BF3 counter for the thermal neutron fluence calibration, a LUDLUM 42-5 Bonner spectrometer with five 0.95 g cm(-3) polyethylene spheres with a LiI(Eu) 4 x 4 mm2 scintillation counter for the neutron spectrometer calibration and a NEMO 9140 remmeter for the paraffin remmeter calibration. The Monte Carlo code MCNP 4C has been used in two ways: to calculate the neutron kerma contribution to two TLDs (type 600 and 700) both in air and inside the paraffin sphere, and to determine the neutron spectra at those Linac room zones where the neutron spectra were measured. Thermal neutron fluences of 2.9 x 10(4) +/- 8.6 x 10(3) cm(-2) s(-1), measured around the Linac head plane, and 2.3 x 10(4) +/- 2.3 x 10(3) cm(-2) s(-1), measured at the patient couch plane, are in agreement with previous independent measurements from other authors. The calculated and measured neutron spectra obtained in the treatment room showed three distinct regions: a peak around 0.1 MeV, a flat epithermal region and a thermal region with values similar to those mentioned above. Patient dose equivalents of 0.5 mSv and 5 mSv from neutrons and photons, respectively, were obtained per treatment Gray.     

Observations on personnel dosimetry for radiotherapy personnel operating high-energy LINACs

A series of measurements were conducted to determine the cause of a sudden increase in personnel radiation exposures. One objective of the measurements was to determine if the increases were related to changing from film dosimeters exchanged monthly to TLD-100 dosimeters exchanged quarterly. While small increases were observed in the dose equivalents of most employees, the dose equivalents of personnel operating medical electron linear accelerators with energies greater than 20 MV doubled coincidentally with the change in the personnel dosimeter program. The measurements indicated a small thermal neutron radiation component around the accelerators operated by these personnel. This component caused the doses measured with the TLD-100 dosimeters to be overstated. Therefore, the increase in these personnel dose equivalents was not due to changes in work habits or radiation environments. Either film or TLD-700 dosimeters would be suitable for personnel monitoring around high-energy linear accelerators. The final choice would depend on economics and personal preference.     

Application of glow curve analysis methods to improve TLD-100 dose reassessment performance.

This paper presents a method of improving the TLD-100 dose reassessment performance. This method consists of applying numerical analysis techniques for evaluating the TLD-100 phototransferred thermoluminescence (PTTL) glow curve. From this analysis, a simple procedure for estimating the ultraviolet background components usually present in phototransferred thermoluminescence (TL) signals has been established. This procedure has been implemented in a computer program which performs the automatic evaluation of the glow curves and extracts the dose information contained in the PTTL curves. The use of this computer-aided evaluational method has enabled the extension of the working range of estimated absorbed dose down to 0.2-0.5 mGy with very adequate operational quality for doses even below the conventionally admitted lower reestimation limit (approximately 2 mGy). Because TL readout is a destructive process, the ability to reestimate doses can be important in any kind of dosimetric activity, such as operational dosimetry programs. The other commonly used dosimeter, film, uses a nondestructive readout and, therefore, presents some advantages over TLD when dose reassessment is necessary. With the reported improvements in the TLD-100 dose reassessment performance, the full range of absorbed doses covered by film dosimetry can now be reliably reassessed using TLD-100 dosimeters.     

A proposed four-element neutron-photon-beta thermoluminescence dosimeter.

It is common practice for a worker exposed to a mixed field with neutrons to wear both a photon-beta dosimeter and a neutron dosimeter. In this study, a thermoluminescence dosimeter has been designed and is proposed for use in mixed fields. The maximum applicable ranges of the mixed field can have photons with unknown energy from 20 keV to 2 MeV, betas with unknown energy from 147Pm to 90Sr-Y, and neutrons of known energy from thermal to 15 MeV. This proposed dosimeter (a combination of Harshaw beta-gamma thermoluminescence dosimeter and albedo neutron thermoluminescence dosimeter) has an advantage of using a minimum number of thermoluminescence dosimeter elements (therefore, making it less costly) to measure the dose equivalents in a mixed field of neutron, photon, and beta radiation. The basic dosimeter design consists of four thermoluminescence elements of TLD-600 and TLD-700 with different filtrations. Using the high-temperature peak methodology for TLD-600 and a filtration algorithm, the neutron, photon, and beta dose equivalents in a mixed field can be determined. The design, detection principle, and three dosimetric algorithms for three versions of the basic design of the four-element dosimeter are presented and discussed. The work that is required for the proposed dosimeter to be usable when it is made is also presented.     

Hiroshima neutron fluence on a glass button from near ground zero.

A decorative glass button that was uncovered at a location that is 190 +/- 15 m from directly beneath the atomic explosion at Hiroshima on 6 August 1945 has been scanned for induced fission tracks produced mostly by the thermal neutrons from the bomb due to interactions with the trace uranium that is normally present in silicate glasses. In surveying 4.14 cm2 at 500x magnification, 28 tracks were seen. From a calibration irradiation in a nuclear reactor we infer that the neutron fluence in 1945 was 5.7(+/-1.1) x 10(11) cm(-2); and, allowing for shielding by the structure in which the button was probably located, the free-air (i.e., outside) value is estimated as 1.5(+/-0.5) x 10(12) cm(-2). A limit has been placed on possible fading of the radiation-damage tracks that could increase the fluence by at most a factor of 1.27. The values bracket the calculated value of 9 x 10(11) given in DS86 but are higher than the 3.6 x 10(11) inferred from induced radionuclides for the distance given. The difference is, however, within the observed variability of the two types of results.     

Radiation does from medical in-vivo prompt gamma-ray activation using a mobile nuclear reactor

A method of medical diagnosis of toxic elements, using a neutron beam from a mobile nuclear reactor to perform partial-body in-vivo prompt gamma-ray activation technique, has been developed. Both neutron and gamma-ray dose equivalents in an irradiated phantom and around medical researchers were measured and evaluated. Neutron flux at various kinetic energies was measured using an activation foil technique, and the neutron dose equivalents at tissues of risk inside the irradiated phantom were calculated by neutron transport code. Gamma-ray dose equivalents inside the irradiated phantom and around the nuclear reactor were measured by thermoluminescent dosimeters. The risk associated with the neutron and gamma radiation dose equivalents received by both the irradiated phantom and medical researchers were evaluated in detail. The radiation safety of the in-vivo medical diagnosis using the mobile nuclear reactor, under the context of radiation protection guidelines, is discussed.     

Measuring the external exposure dose in the contaminated area near the Chernobyl nuclear power station using the thermoluminescence of quartz in bricks

We collected bricks from buildings in the heavily contaminated evacuated area of Belarus in a 30-km zone around the Chernobyl nuclear power station and the Gomel-Bryansk area of 150-250 km from Chernobyl and estimated the cumulative radiation dose caused by the reactor accident by measuring the thermoluminescence (TL) of the bricks. The annual dose at each location was measured using glass dosimeters and thermoluminescence dosimeters (TLD). The dose rate was measured using an energy-compensated NaI scintillation survey meter. The soil contamination near the location of each brick was measured using a germanium semiconductor detector. The main purpose of the project was to extrapolate the relation between the cumulative external dose and the present dose rate or contamination level to the lower contaminated areas. The results of the glass dosimeter, TLD, and survey meter determinations were almost identical. For a determination of the annual dose higher than 10 mGy y(-1), the cumulative dose by TL (TL dose) was roughly proportional to the annual dose and about 1.5 times larger than the cumulative dose calculated from the annual dose and 137Cs half life. The difference is expected due to the contribution of short-lived nuclides immediately after the accident or localized heavy contamination of the ground surface with 137Cs that migrated afterwards. For annual dose smaller than 10 mGy y(-1), the proportionality was not observed and most of the locations facing indoors showed TL doses very much larger than that expected from the proportionality. The cumulative dose outdoors by TL was also roughly proportional to the regional 137Cs contamination level and the proportional constant is about 10(-1) mGy per GBq km(-2), and is about 250 times larger than the present annual internal dose derived from published results. The correlation between the present dose rate where the brick was sampled and the average 137Cs concentration in the ground soil near the point is not clear, possibly because of the large spatial fluctuation in the 137Cs concentration in the soil.     

Boron neutron capture therapy (BNCT) for malignant melanoma with special reference to absorbed doses to the normal skin and tumor

Twenty-two patients with malignant melanoma were treated with boron neutron capture therapy (BNCT) using 10B-p-boronophenylalanine (BPA). The estimation of absorbed dose and optimization of treatment dose based on the pharmacokinetics of BPA in melanoma patients is described. The doses of gamma-rays were measured using small TLDs of Mg2SiO4 (Tb) and thermal neutron fluence was measured using gold foil and wire. The total absorbed dose to the tissue from BNCT was obtained by summing the primary and capture gamma-ray doses and the high LET radiation doses from 10B(n, alpha)7Li and 14N(n,p)14C reactions. The key point of the dose optimization is that the skin surrounding the tumour is always irradiated to 18 Gy-Eq, which is the maximum tolerable dose to the skin, regardless of the 10B-concentration in the tumor. The neutron fluence was optimized as follows. (1) The 10B concentration in the blood was measured 15-40 min after the start of neutron irradiation. (2) The 10B-concentration in the skin was estimated by multiplying the blood 10B value by a factor of 1.3. (3) The neutron fluence was calculated. Absorbed doses to the skin ranged from 15.7 to 37.1 Gy-Eq. Among the patients, 16 out of 22 patients exhibited tolerable skin damage. Although six patients showed skin damage that exceeded the tolerance level, three of them could be cured within a few months after BNCT and the remaining three developed severe skin damage requiring skin grafts. The absorbed doses to the tumor ranged from 15.7 to 68.5 Gy-Eq and the percentage of complete response was 73% (16/22). When BNCT is used in the treatment of malignant melanoma, based on the pharmacokinetics of BPA and radiobiological considerations, promising clinical results have been obtained, although many problems and issues remain to be solved.     

The enhancement of neutron irradiation of HeLa-S cervix carcinoma cells by cell-nucleus-addressed deca-p-boronophenylalanine

Boron neutron capture therapy (BNCT) is an experimental treatment modality which depends on a sufficient cellular uptake of Boron ((10)B) followed by an exposure to a thermal neutron beam from a nuclear reactor. High energetic particles (4He and 7Li) are created during the neutron capture reaction and produce DNA damages, which lead to cell killing. Regarding BNCT, the short radiation range of He- and Li-particles is decisive for the distribution of (10)B. Until now, BNCT has been lacking for therapeutically effective concentrations of (10)B. Twenty-four hours after the combined use of our 'Bioshuttle'-p-borono-phenylalanine(10)-constructs ('Bioshuttle'-p-BPA(10)) and neutron-irradiation, an obvious reduction of the radiation-resistant HeLa-S cells could be observed. No cells were alive 72 h after the incubation with 'Bioshuttle'-p-BPA(10) followed by neutron irradiation. A post-mitotic cell death could be assumed based on flow cytometrical data.     

An experimental study on radiation streaming through a labyrinth in a proton accelerator facility of intermediate energy

A radiation streaming experiment has been carried out at the Takasaki Ion Accelerator Facility for Advanced Radiation Application at the Japan Atomic Energy Research Institute in a room housing a Cu target irradiated with 68 MeV protons and in a labyrinth of three-legs having a total length of 29 m. In the experiment, neutron and gamma ray energy spectra, neutron reaction rates, and neutron and gamma ray dose equivalent rates were measured using various counters and dosimeters. The experimental data show the applicability of some empirical formulas for estimating the thermal neutron flux in a room and neutrons streaming in a labyrinth designed for a proton accelerator operating in the intermediate energy region. The data suggest that it is mandatory to estimate the gamma ray dose equivalent rate in a labyrinth, which is dominated by the secondary gamma rays due to the neutron capture reaction.     

Accelerator room photoneutron and photon background measurements using thermoluminescent dosimeters

Photoneutron dose equivalents and photon doses in the treatment room of a clinical linear accelerator were measured with sets of isotopically enriched LiF thermoluminescent dosimeters and a moderating sphere. Dosimeter neutron calibrations with 252Cf sources were repeated many times during the extended series of measurements because the 6LiF dosimeter sensitivity increased with successive neutron irradiations. Expressed as a fraction of the primary bremsstrahlung beam dose at maximum, the photoneutron background was 2.04 +/- 0.05 mrem/rad (10(-3) Sv/Gy) at 1 m lateral to beam center in the patient midplane at 25 MV. The fraction of this result due to thermal neutrons was found to be only about 2%. The photon background dose was 2.98 +/- 0.04 mrad/rad (10(-3) Gy/Gy). The photoneutron dose equivalent per unit primary dose was found to be nearly independent of the collimator size used but increased by 40% when the bremsstrahlung endpoint energy was increased from 20 to 35 MeV with no change in flattening filters.     

Use of glazes on porcelain from near ground zero to measure Hiroshima neutron fluence

Several porcelain samples from almost directly beneath the atomic explosion at Hiroshima on 6 August 1945, have been scanned for induced fission tracks, produced mostly by the thermal neutrons from the bomb due to interactions with trace uranium in their glass coatings. The ability to use porcelain opens a new and abundant material for retrospective dosimetry. Four different samples had thermal neutron fluences in 1945 of 1.0, 3.8, 4.1, and 8.9 x 10(12) cm(-2). The different values are not caused by track fading, but are likely to result from differing shielding at different nearby positions. Assuming that the three highest fluences, which have overlapping uncertainties, are at locations of minimum shielding, the statistically weighted thermal fluence in the air at ground level and ground zero was 4.8 x 10(12) cm(-2) with a statistical uncertainty of 15%. This value lies between the calculated value of 6.5 x 10(12) given in DS86 and the 3.7 x 10(12) inferred from induced radionuclides by Hoshi et al. (1998).     

Spectral and dosimetric characteristics of a D2O-moderated 252Cf calibration facility

Calculations have been made for a D2O-moderated 252Cf assembly like that being used for the calibration of neutron dosimeters at the U.S. National Bureau of Standards and being proposed by the International Standards Organization. Leakage spectra at various distances from the assembly are given along with variations in dose-equivalent rate, average neutron energy and 235U/237Np fission ratio. The spectral shape changes rapidly near the spherical assembly and the dose-equivalent rate changes more rapidly than would be expected on the basis of the inverse-square dependence. Calibration of neutron dosimeters should therefore be made at distances greater than 15 cm from the surface. At large distances from the source, the dose equivalent per unit fluence for neutrons above 1 eV is 9.3 X 10(-9) rem cm2. The effects of the structural material, recent revisions to nuclear data files and changes in the spectrum of the source neutrons on the external field were investigated. These changes produce only about a 5% change in the neutron-dose equivalent rate. The structural material introduces negligible anisotropy in the radiation field.     

Study of the radiation field characteristics of a 137Cs irradiator by Monte Carlo simulation

In this work a study of the energy fluence of the photon beam produced by a commercial irradiator that uses a single collimated 137Cs source is performed by employing the Monte Carlo code PENELOPE. A set of lead attenuators is placed at the exit window of the irradiator to vary the air kerma rate that is required to cover the instrument scales at a particular calibration distance. A possible variation in response due to this beam modification isalso investigated for LiF (TLD-100) dosimeters and for a secondary standard radiation protection level ionization chamber. The results show an important enhancement of beam mean energy from 633 to 642 keV as the lead attenuators increase in thicknesses. For this energy range, a maximum response change of 45% was found for LiF and 4.4% for the ionization chamber. These results reinforce the idea that a single source may very well be a practical solution for calibration laboratories without compromising the overall uncertainties acceptable for this application.     

An update on the discrepancy between calculated and measured neutron-induced radioactivity levels in Hiroshima.

The thermal neutron activation measurements carried out over many years in Hiroshima and Nagasaki have been the subject of ongoing debate in recent years because they indicate that current DS86 neutron doses may have been significantly underestimated in Hiroshima. Long-lived neutron activation products, 60Co, 152Eu, 154Eu and 36Cl, which are still detectable today using modern analytical techniques, appear to indicate that DS86 calculated thermal neutron activation products decrease with distance more rapidly than the measured values. The latest thermal neutron activation measurements have been collated and a new relationship for the measured to calculated (M/C) ratio of induced activity has been derived as a function of slant range. This indicates a stronger dependence of M/C on slant range than previously derived by Straume et al (1992 Health Phys. 63 421-6) and emphasises even more the discrepancy between measured and calculated (DS86) neutron doses at distances beyond 1 km. While the main body of thermal neutron activation data appears to support a significant increase in the DS86 neutron dose component in Hiroshima, there are some thermal neutron activation measurements and some very recent fast neutron activation measurements which suggest that the discrepancy may not be so great. The extent of the required revision to the neutron component of the DS86 dosimetry remains the subject of ongoing new neutron activation measurements and re-analysis of existing published measurements. A companion paper considers the impact on radiation risk estimates of possible modifications to the DS86 dosimetry system on the basis of a broad range of interpretations of the neutron activation data.     

Glow-peak stability in 6LiF:Mg,Ti (TLD-600) exposed to a Fe-ion beam

The stability of glow peaks in 6LiF:Mg,Ti (TLD-600) exposed to a high-energy Fe-ion beam was examined in comparison to 137Cs gamma-ray irradiation under changing annealing conditions. The peak areas induced by the Fe ions were much smaller than those by gamma-rays. The sizes and positions of peaks 3-5 in Fe-ion irradiated samples were hardly changed after post-annealing at 100 degrees C x 30 min, regardless of the pre-annealing conditions (fast quenching or subsequent pre-annealing at 100 degrees C x 2 h). Whereas, the peaks in gamma-ray irradiated samples were notably affected by post-annealing; the peak positions and peak-area sizes changed in different ways depending on the pre-annealing conditions. The effects of post-annealing on peak 6 were identical for Fe ions and gamma-rays. These facts suggest that peaks 3-5 in TLD-600 comprised both stable and unstable luminescent centers, and that the latter part would be easily depleted in highly dense ionization.     

Preliminary studies to develop a personal dosemeter for use by aircraft crew.

This paper describes preliminary work to develop a cosmic-radiation dosemeter for use by military aircraft crew. The dosemeter is based on a combination of CR-39 etched-track detectors and TLD-700 thermoluminescent detectors. It is intended that the CR-39 be used to assess the neutron dose, while the TLD-700 is used to assess the photon and charged particle dose. The sensitivity of CR-39 to the neutron component of cosmic radiation was estimated by irradiating samples of the plastic at the CERN-CEC High Energy Reference Field Facility. This facility produced a radiation field with a neutron spectrum resembling that of the neutron component of cosmic radiation at typical airflight altitudes. The response of the CR-39 was linear over the range of doses studied (0.2-6.0 mSv) and there was no significant fading in the 6-month period after irradiation. The TLD-700 component of the dosemeter was calibrated using 137Cs gamma rays. The response of the TLD-700 was linear over the range of doses studied (0.01-5.0 mSv) with no significant fading in the 6-month period after irradiation. It was concluded that a combination of CR-39 and TLD-700 detectors would provide an effective cosmic-radiation dosemeter for use by military aircraft crew.     

Neutron dosimetry in the containment of a pressurized water reactor using a neutron-sensitive beta/gamma dosimetry system

In this study the Panasonic UD-802 dosimeter was evaluated as a potential neutron dosimeter for use in the containment of a pressurized water reactor by comparing the results from the UD-802 with remmeter readings. The Panasonic UD-802 dosimeter is used routinely as a beta and gamma dosimeter but due to the natural Li and B in the thermoluminescent materials, it is also sensitive to neutrons. Since a dosimeter's response to neutrons is energy-dependent, proper calibration of the UD-802 in the environment for which it is to be used was an important consideration of the study. To calibrate the system, UD-802 dosimeters were mounted on polyethylene phantoms and irradiated to reference doses at selected locations in containment. The reference doses were determined based on remmeter dose-rate measurements and stay times. The thermoluminescent response of the dosimeters and the reference measurements were used to obtain a response ratio at each location. The average response ratio (unit of dosimeter response per millirem) was 3.7 and all response ratios were within +/-30% of this mean value. Specific characteristics of the UD-802 were also investigated, that is, the effects that dosimeter distance from the phantom and a person's movement through containment have on response. The dosimeter distance from the phantom was found to have a minimal effect on response, but the system was found to be dependent upon the angle of the phantom relative to the reactor core, necessitating a correction in the calibration factor. The overall conclusion of this study was that the Panasonic UD-802 dosimeter can be used for neutron dosimetry in containment of a pressurized water reactor.     

Nuclear accident dosimetry intercomparison studies

Twenty-two nuclear accident dosimetry intercomparison studies utilizing the fast-pulse Health Physics Research Reactor at the Oak Ridge National Laboratory have been conducted since 1965. These studies have provided a total of 62 different organizations a forum for discussion of criticality accident dosimetry, an opportunity to test their neutron and gamma-ray dosimetry systems under a variety of simulated criticality accident conditions, and the experience of comparing results with reference dose values as well as with the measured results obtained by others making measurements under identical conditions. Sixty-nine nuclear accidents (27 with unmoderated neutron energy spectra and 42 with eight different shielded spectra) have been simulated in the studies. Neutron doses were in the 0.2-8.5 Gy range and gamma doses in the 0.1-2.0 Gy range. A total of 2,289 dose measurements (1,311 neutron, 978 gamma) were made during the intercomparisons. The primary methods of neutron dosimetry were activation foils, thermoluminescent dosimeters, and blood sodium activation. The main methods of gamma dose measurement were thermoluminescent dosimeters, radiophotoluminescent glass, and film. About 68% of the neutron measurements met the accuracy guidelines (+/- 25%) and about 52% of the gamma measurements met the accuracy criterion (+/- 20%) for accident dosimetry.     

Locating tritium sources in a research reactor building

Despite renovation of the D2O facility, tritium concentrations in the condensates of reactor room air showed tens of Bq mL before venting resumption on July 1997. This suggested the presence of tritium sources in the research reactor-containment building. An investigation was therefore initiated to locate the source and determine the distribution of tritium in the containment building. Air monitoring in the working area using a dish of water placed in the building suggested that the source of tritium was near the reactor core. Monitoring exhaust air from the two facilities (a cold neutron source and a D(2)O tank) showed high specific activity on the order of 10 Bq mL(-1), suggesting the presence of tritium in condensates near the reactor core. The major concern was whether the leakage of liquid deuterium (4 L) and heavy water (2 x 10(3) L) used as a moderator had occurred. The concentration of tritium in condensates has not increased over the past few years in either the exhaust line or working area, and the deuterium itself has not been found in the surrounding environment. The concentration of tritium measured using an ionization chamber after Ar decay was dependent on the thermal output of the research reactor, indicating that the tritium was produced by the irradiation process within shielding/moderator materials or cover gas with neutrons.