International intercomparison for criticality dosimetry: the case of biological dosimetry

The Institute of Radiation Protection and Nuclear Safety (IRSN) organized a biological dosimetry international intercomparison with the purpose of comparing (i) dicentrics yield produced in human lymphocytes; (ii) the gamma and neutron dose estimate according to the corresponding laboratory calibration curve. The experimental reactor SILENE was used with different configurations: bare source 4 Gy, lead shield 1 and 2 Gy and a 60Co source 2 Gy. An increasing variation of dicentric yield per cell was observed between participants when there were more damages in the samples. Doses were derived from the observed dicentric rates according to the dose-effect relationship provided by each laboratory. Differences in dicentric rate values are more important than those in the corresponding dose values. The doses obtained by the participants were found to be in agreement with the given physical dose within 20%. The evaluation of the respective gamma and neutron dose was achieved only by four laboratories, with some small variations among them.     

A cytogenetic follow-up of some highly irradiated victims of the Chernobyl accident

A follow-up of 10 highly irradiated men, mostly reactor crew, from the Chernobyl accident is described. Their pre-accident medical conditions and relevant medical status approximately 10-13 y later are listed. A comparison is made between estimates of their average whole-body penetrating radiation doses derived from several biological parameters. First estimates were based on their presenting severity of prodromal sickness, early changes in blood cell counts and dicentric chromosome aberrations in lymphocytes. In three cases ESR measurements on tooth enamel were also made. Retrospective dosimetry using FISH translocations was attempted 10-13 y later. This showed good agreement for those patients with the lower earlier dose estimates, up to about 3 Gy. For the others, extending up to about 12 Gy, the translocations indicated lower values, suggesting that in these cases translocations had somewhat declined. Repeated chromosomal examinations during the follow-up period showed an expected decline in dicentric frequencies. The pattern of decline was bi-phasic with a more rapid first phase, with a half-life of approximately 4 months followed by a slower decline with half-lives around 2-4 y. The rapid phase persisted for a longer time in those patients who had received the highest doses. 10-13 y later dicentric levels were still above normal background, but well below the translocation frequencies.     

Bayesian methods for chromosome dosimetry following a criticality accident

Radiation doses received during a criticality accident will be from a combination of fission spectrum neutrons and gamma rays. It is desirable to estimate the total dose, as well as the neutron and gamma doses. Present methods for dose estimation with chromosome aberrations after a criticality accident use point estimates of the neutron to gamma dose ratio obtained from personnel dosemeters and/or accident reconstruction calculations. In this paper a Bayesian approach to dose estimation with chromosome aberrations is developed which allows the uncertainty of the dose ratio to be considered. Posterior probability densities for the total and the neutron and gamma doses were derived.     

Differential absorbed dose distributions in lineal energy for neutrons and gamma rays at the mono-energetic neutron calibration facility

Absorbed dose distributions in lineal energy for neutrons and gamma rays of mono-energetic neutron sources from 140 keV to 15 MeV were measured in the Fast Neutron Laboratory at Tohoku University. By using both a tissue-equivalent plastic walled counter and a graphite-walled low-pressure proportional counter, absorbed dose distributions in lineal energy for neutrons were obtained separately from those for gamma rays. This method needs no knowledge of energy spectra and dose distributions for gamma rays. The gamma-ray contribution in this neutron calibration field >1 MeV neutron was <3%, while for <550 keV it was >40%. The measured neutron absolute absorbed doses per unit neutron fluence agreed with the LA150 evaluated kerma factors. By using this method, absorbed dose distributions in lineal energy for neutrons and gamma rays in an unknown neutron field can be obtained separately.     

Alanine and TLD coupled detectors for fast neutron dose measurements in neutron capture therapy (NCT)

A method was investigated to measure gamma and fast neutron doses in phantoms exposed to an epithermal neutron beam designed for neutron capture therapy (NCT). The gamma dose component was measured by TLD-300 [CaF2:Tm] and the fast neutron dose, mainly due to elastic scattering with hydrogen nuclei, was measured by alanine dosemeters [CH3CH(NH2)COOH]. The gamma and fast neutron doses deposited in alanine dosemeters are very near to those released in tissue, because of the alanine tissue equivalence. Couples of TLD-300 and alanine dosemeters were irradiated in phantoms positioned in the epithermal column of the Tapiro reactor (ENEA-Casaccia RC). The dosemeter response depends on the linear energy transfer (LET) of radiation, hence the precision and reliability of the fast neutron dose values obtained with the proposed method have been investigated. Results showed that the combination of alanine and TLD detectors is a promising method to separate gamma dose and fast neutron dose in NCT.     

Bayesian estimation of dose thresholds

An example is described of Bayesian estimation of radiation absorbed dose thresholds (subsequently simply referred to as dose thresholds) using a specific parametric model applied to a data set on mice exposed to 60Co gamma rays and fission neutrons. A Weibull based relative risk model with a dose threshold parameter was used to analyse, as an example, lung cancer mortality and determine the posterior density for the threshold dose after single exposures to 60Co gamma rays or fission neutrons from the JANUS reactor at Argonne National Laboratory. The data consisted of survival, censoring times and cause of death information for male B6CF1 unexposed and exposed mice. The 60Co gamma whole-body doses for the two exposed groups were 0.86 and 1.37 Gy. The neutron whole-body doses were 0.19 and 0.38 Gy. Marginal posterior densities for the dose thresholds for neutron and gamma radiation were calculated with numerical integration and found to have quite different shapes. The density of the threshold for 60Co is unimodal with a mode at about 0.50 Gy. The threshold density for fission neutrons declines monotonically from a maximum value at zero with increasing doses. The posterior densities for all other parameters were similar for the two radiation types.     

Bayesian analysis of overdispersed chromosome aberration data with the negative binomial model

The usual assumption of a Poisson model for the number of chromosome aberrations in controlled calibration experiments implies variance equal to the mean. However, it is known that chromosome aberration data from experiments involving high linear energy transfer radiations can be overdispersed, i.e. the variance is greater than the mean. Present methods for dealing with overdispersed chromosome data rely on frequentist statistical techniques. In this paper. the problem of overdispersion is considered from a Bayesian standpoint. The Bayes Factor is used to compare Poisson and negative binomial models for two previously published calibration data sets describing the induction of dicentric chromosome aberrations by high doses of neutrons. Posterior densities for the model parameters, which characterise dose response and overdispersion are calculated and graphed. Calibrative densities are derived for unknown neutron doses from hypothetical radiation accident data to deterimine the impact of different model assumptions on dose estimates. The main conclusion is that an initial assumption of a negative binomial model is the conservative approach to chromosome dosimetry for high LET radiations.     

A molecular fraction method for measuring personnel radiation doses

This work represents a development in fast and albedo neutron and gamma ray dosimetry, using cellulose nitrate, as a tissue equivalent material, in which radiation damage was registered.The changes in molecular fractions of the polymer were measured after irradiation with neutron fluences from a ²⁵²Cf source in the range 10⁵−10¹⁰ n/cm² and gamma doses in the range 10⁻⁴–10⁻¹ Gy through the use of gel filtration chromatography.Effects of irradiation on phantom, phantom to dosimeter distance, phantom thickness and storage at extreme environmental conditions were studied on the detector response and readout.The results showed that main chain scission followed by formation of new molecular configurations is the predominant effect of radiation on the polymer. The method enables measurements of neutron fluences and gamma doses in mixed radiation fields. Empirical formulae for calculating the absorbed dose from the measured changes in molecular fraction intensities are given.     

Evaluation of gamma and neutron irradiation effects on the properties of mica film capacitors

We present an investigation of gamma and neutron radiation effects on mica film capacitors from an electrical point of view. We have studied quantitatively the effects of gamma and neutron irradiation on mica film capacitors of thickness, 20 and 40 μm (0.7874 and 1.5748 mil) with two different areas, 01 and 04 cm². The capacitance has been measured at room temperature in the frequency range 100 Hz-10 MHz. Negligible change in the capacitance due to high gamma dose of⁶⁰Co, 15 kGy at dose rate 0.25 kGy/h, has been observed. However, appreciable change in the capacitance has been observed due to low doses of fast neutrons (cumulative dose, 115 cGy) with flux ∼ 9.925 x 10⁷ neutrons/cm² h from²⁵²Cf neutron source of fluence, 2.5 × 10⁷ neutrons/s. We have also observed that the impact of gamma and neutron irradiation is more at frequencies higher than 10 kHz. These results show that the mica capacitors do not show any radiation response below 10 kHz. The study shows the radiation response of mica film capacitors to gamma and fast neutron radiations. Mica capacitors show low gamma radiation response in comparison to fast neutron radiation, because a total dose of kGy order has been given by gamma source and only few cGy dose has been given by fast neutron source.     

The performance evaluation of gamma- and neutron-sensitive superheated emulsion (bubble) detectors

The superheated emulsion (bubble) detectors have been developed at Defence Laboratory, Jodhpur (DLJ), India, for measurement of gamma doses. The developed detectors have been tested at Radiation Safety and System Division (RSSD), Bhabha Atomic Research Center (BARC), Mumbai (India) and DLJ having ISO-17025 accredited facility for testing and calibration of Radiation Monitors. A series of experiments were conducted to determine the gamma and neutron sensitivity of these detectors, i.e. batch homogeneity, reproducibility, dose equivalent rate effect, gamma/neutron dose equivalent response, gamma/neutron energy response and change in gamma sensitivity as a function of temperature. All the results were within +/- 20% of themselves. It is observed that the response of the detector is dependent upon temperature. The recommended ideal working temperature range of the detector is 20-28 degrees C, but a temperature correction is required beyond approximately 30 masculineC. The temperature compensation may be possible up to 45 degrees C in improved version using specially prepared reversible thermo-sensitive polymer gel. The detector may have applications in radio-diagnosis, R&D laboratories, and health physics as well as an indicator of gamma radiation for dirty bomb to be useful for first responder in any radiological emergency.     

Calibration curves for biological dosimetry by fluorescence in situ hybridisation

Dose-response curves were measured for the induction of chromosomal aberrations in peripheral blood lymphocytes after acute exposure in vitro to 60Co gamma rays. Blood was obtained from four different healthy donors, and chromosomes were either observed at metaphase, following colcemid accumulation, or prematurely condensed by calyculin A. Cells were analysed in three different Italian laboratories. Chromosomes 1, 2, and 4 were painted, and simple-type interchanges between painted and non-painted chromosomes were scored in cells exposed in the dose range 0.1-3.0 Gy. The chemical-induced premature chromosome condensation method was also used combined with chromosome painting (chromosome 4 only) to determine calibration curves for high dose exposures (up to 20 Gy X rays). Calibration curves described in this paper will be used in our laboratories for biological dosimetry by fluorescence in situ hybridisation.     

Method for determination of gamma and neutron dose components in mixed radiation fields using a high-pressure recombination chamber

A new method is proposed for the determination of dose components in mixed radiation fields (gamma + neutrons) using a recombination chamber. The method involves the determination of the ratio of ionisation currents measured at two different voltages applied to the chamber without the need of determining the saturation current, neither in the radiation field investigated nor during calibration. Therefore, the chamber can be filled with a gas under a pressure much higher than that used in presently available recombination chambers. This paper presents theoretically derived formulae supporting the method and the experimental results of dose component measurements using a high-pressure recombination chamber filled with methane. The method can be used for determining neutron and gamma dose components in the environment, especially in the vicinity of nuclear centres.     

Study and presentation of a fast neutron and photon dosemeter for area and criticality monitoring using radiophotoluminescent glass

This paper describes the results of a study performed on a mixed field neutron/gamma (n/gamma) area dosemeter incorporating radiophotoluminescent (RPL) glass detectors. RPL glass is known to be virtually insensitive to neutrons. The aim of the study was therefore to determine the neutron response of a dosemeter designed to combine n/gamma conversion with RPL detection capability. Monte Carlo calculations as well as measurements using monoenergetic beams and isotopic neutron sources showed this response to be constant, to within 30% in terms of H*(10), and independent of neutron energy from 250 keV to 10 MeV. For area monitoring, tests carried out in nuclear facilities (around PuO2 glove box and shipping casks containing PWR, MOX spent fuels or vitrified fission product) demonstrated that dosemeter response was accurate to within 15%, where the gamma component of the mixed n,gamma field remained below 1 MeV. When exposed in the Silene reactor simulating a criticality accident (10(17) fissions-liquid 235U--e.g. 1 Gy neutron and 1 Gy photon), the dosemeter exhibited good correlation with reference values and other measurement technologies (again to within 30%), for both neutron and gamma absorbed dose.     

On the reassessment of thermal neutron doses in TLD-100 by measuring the residual dose

By employing second readouts and the Phototransferred thermoluminescence (PTTL) method, high doses may be reassessed on the basis of residual dose information. It was shown in the past that for TLD-100, gamma doses can be reassessed by using a simple and efficient method, which consists of expanding the heating time to 30 s. In the present study, the 'extended time' method and the PTTL residual dose evaluations are used for reassessing thermal neutron doses when using TLD-100 crystals. Reassessment characteristics are presented for relatively low thermal neutron doses, in the range between approximately 1 and 18 mSv gamma dose equivalent.     

Thermoluminescence dosimetry of a thermal neutron field and comparison with Monte Carlo calculations

The characteristics of thermoluminescence dosemeters (TLDs) regarding the determination of photon and neutron absorbed doses were investigated in a thermal neutron beam. Harshaw TLD-100 (LiF:Mg,Ti) and TLD-700 (7LiF:Mg,Ti) were compared with similar materials from Solid Dosimetric Detector and Method Laboratory (People's Republic of China). Harshaw TLD-700H (7LiF:Mg,Cu,P) and aluminium oxide (Al2O3:Mg,Y) from Hungary were also considered for photon dose measurement. The neutron sensitivity of the investigated materials was measured and found to be consistent with values reported by other authors. A comparison was made between the TL dose measurements and results obtained via conventional methods. An agreement within 20% was obtained, which demonstrates the ability of TLD for measuring neutron and photon doses in a mixed field, using careful calibration procedures and determining the neutron sensitivity for the usage conditions.     

Dose measuring system for boron neutron capture therapy

Dose measuring systems for boron neutron capture therapy (BNCT) of brain tumors are presented. The systems are a real-time monitoring system, an integral measuring system and a ¹⁰B concentration measuring system. The real-time monitoring with a small PN junction silicon detector made it possible to simultaneously measure the thermal neutron flux and the gamma dose rate in a patient during neutron therapy. Another monitoring of dose equivalents of thermal neutrons and gamma rays was performed with a BGO scintillation detector connected to an optical fiber. The accurate neutron fluence and gamma dose were determined with the integral measurements of the foil activation method and thermoluminescent dosimeters (TLDs) after irradiation. Kerma doses of thermal neutrons and gamma-rays were also measured with the TLD at the same time. Preliminary measurements of ¹⁰B concentration in tissue and blood of a patient were carried out by prompt gamma-ray spectroscopy.     

Ambient gamma radiation levels (indoor and outdoor) in the villages around Jaduguda (India) using card-based CaSO₄: Dy TL dosemeters

A systematic study of the gamma radiation levels (indoor and outdoor) in the villages surrounding the uranium-enriched regions around Jaduguda, India has been undertaken by monitoring selected dwellings in six villages. Each dwelling unit was monitored for a total duration of 1 y. The gamma radiation measurements were carried out using card-based CaSO(4): Dy thermoluminescent dosemeters. The estimated average annual gamma dose values for indoor and outdoor were 980 and 924 (μGy y(-1)), respectively, for the entire region studied. The maximum indoor and outdoor gamma doses experienced in North Dungridih and South Dungridih villages were 1305 and 1223 (μGy y(-1)), respectively. The minimum indoor and outdoor gamma dose values observed in Chatikocha village were 624 and 696 (μGy y(-1)), respectively. Seasonal variation of the indoor gamma values was not observed during the year; however, a small variation was seen with the type of building materials used for construction purposes. A statistical analysis was attempted to characterise the distribution of terrestrial gamma radiation obtained in the study area. The average quarterly indoor gamma values for spring, summer, monsoon and winter seasons as prevalent in the regions were 267±71, 262±54, 213±91, 238±66 (μGy 90 d(-1)), respectively. The annual effective doses to the local population residing in the selected dwelling units were estimated to be 0.6 and 0.1 (mSv y(-1)) for indoor and outdoor, respectively, using an occupancy factor of 0.8 and 0.2.     

Visual and numerical methods to measure patient skin doses in interventional procedures using radiochromic XR-RV2 films

Radiochromic XR-RV2 films are considered as one of the best dosemeters to measure patient skin doses in fluoroscopy-guided interventional procedures. To fulfil this purpose, they need to be calibrated with diagnostic energies and doses beyond several Gray. The vendor provides a visual calibration strip to estimate the absorbed dose. Differences between visual dose estimation versus film digitisation were investigated. The influence of backscatter radiation on film sensitivity was also investigated and the sources of uncertainty were analysed when skin doses were measured with these films. When based on the visual comparison with the strip, the estimation of the dose resulted in an error of 50 % (2 Gy in the region around 4 Gy). However, when using numerical methods after film digitisation, the uncertainty in dose measurement fell to 7-14 % in the dose range of interest. Calibration under backscatter conditions demonstrates that the 'in air' calibration underestimates the doses by 7 %. When the dose was measured with a calibration method based on 16 bits grey digitisation, uncertainty was twice higher than when the red channel from red, green, blue digitised images was used.     

In-phantom spectra and dose distributions from a high-energy neutron therapy beam

In radiotherapy with external beams, healthy tissues surrounding the target volumes are inevitably irradiated. In the case of neutron therapy, the estimation of dose to the organs surrounding the target volume is particularly challenging, because of the varying contributions from primary and secondary neutrons and photons of different energies. The neutron doses to tissues surrounding the target volume at the Louvain-la-Neuve (LLN) facility were investigated in this work. At LLN, primary neutrons have a broad spectrum with a mean energy of about 30 MeV. The transport of a 10×10 cm² beam through a water phantom was simulated by means of the Monte Carlo code MCNPX. Distributions of energy-differential values of neutron fluence, kerma and kerma equivalent were estimated at different locations in a water phantom. The evolution of neutron dose and dose equivalent inside the phantom was deduced. Measurements of absorbed dose and of dose equivalent were then carried out in a water phantom using an ionization chamber and superheated drop detectors (SDDs). On the beam axis, the calculations agreed well with the ionization chamber data, but disagreed significantly from the SDD data due to the detector's under-response to neutrons above 20 MeV. Off the beam axis, the calculated absorbed doses were significantly lower than the ionization chamber readings, since gamma fields were not accounted for. The calculated data are doses from neutron-induced charge particles, and these agreed with the values measured by the photon-insensitive SDDs. When exposed to the degraded spectra off the beam axis, the SDD offered reliable estimates of the neutron dose equivalent.     

基于D-D中子源校准中子周围剂量当量仪的最小房间尺寸研究

在中子周围剂量当量仪校准过程中,使用D-D中子源代替传统同位素中子源会使校准过程具有更高的安全性。为推动D-D中子源在中子周围剂量当量仪校准过程中的应用,提出了应用蒙特卡罗(MC)法的能量截断法代替影锥法的散射中子研究方法;分析了在不同类型房间中,不同内部空间尺寸对散射中子的影响。计算结果表明随着房间内部空间的增大,入射探测器的散射中子所占比例会逐渐减小。若使用D-D中子源进行中子周围剂量当量仪校准,在探测距离75cm,探测器直径20cm的情况下,所需的最小正方体房间的内部空间棱长为332cm;所需最小长方体房间的内部空间的长、宽、高分别为410cm、410cm、205cm。