Combined TL and 10B-alanine ESR dosimetry for BNCT

The dosimetric technique described in this paper is based on electron spin resonance (ESR) detectors using an alanine-boric compound acid enriched with (10)B, and beryllium oxide thermoluminescent (TL) detectors; with this combined dosimetry, it is possible to discriminate the doses due to thermal neutrons and gamma radiation in a mixed field. Irradiations were carried out inside the thermal column of a TRIGA MARK II water-pool-type research nuclear reactor, also used for Boron Neutron Capture therapy (BNCT) applications, with thermal neutron fluence from 10(9) to 10(14) nth cm(-2). The ESR dosemeters using the alanine-boron compound indicated ESR signals about 30-fold stronger than those using only alanine. Moreover, a negligible correction for the gamma contribution, measured with TL detectors, almost insensitive to thermal neutrons, was necessary. Therefore, a simultaneous analysis of our TL and ESR detectors allows discrimination between thermal neutron and gamma doses, as required in BNCT.     

Study of a method based on TLD detectors for in-phantom dosimetry in BNCT

A method has been developed, based on thermoluminescent dosemeters (TLD), aimed at measuring the absorbed dose in tissue-equivalent phantoms exposed to thermal or epithermal neutrons, separating the contributions of various secondary radiation generated by neutrons. The proposed method takes advantage of the very low sensitivity of CaF2:Tm (TLD-300) to low energy neutrons and to the different responses to thermal neutrons of LiF:Mg,Ti dosemeters with different 6Li percentage (TLD-100, TLD-700, TLD-600). The comparison of the results with those obtained by means of gel dosemeters and activation foils has confirmed the reliability of the method. The experimental modalities allowing reliable results have been studied. The glow curves of TLD-300 after gamma or neutron irradiation have been compared; moreover, both internal irradiation effect and energy dependence have been investigated. For TLD-600, TLD-100 and TLD-700, the suitable fluence limits have been determined in order to avoid radiation damage and loss of linearity.     

Measurement of the fluence response of the GSI neutron ball dosemeter in the energy range from thermal to 19 MeV

At high-energy particle accelerators, area monitoring needs to be performed in a wide range of neutron energies. In principle, neutrons occur from thermal energies up to the energy of the accelerated ions, which is for the present GSI (Gesellschaft für Schwerionenforschung) accelerator facility approximately 1-2 GeV per nucleon. There are no passive dosemeters available, which are designed for the use at high-energy accelerators. At GSI, a neutron dosemeter was developed, which is suitable for the measurement of high-energy neutron radiation by the insertion of a lead layer around Thermoluminescence (TL) detection elements (pairs of TL 600/700) at the centre of the dosemeter. The design of the sphere was derived from the construction of the extended range rem-counters for the measurement of ambient dose equivalent H(10). In this work, the dosemeter fluence response was measured in the quasi-monoenergetic neutron fields of the accelerator facility of the PTB in Braunschweig and in the thermal neutron field of the GKSS research reactor FRG-1 in Geesthacht. For the accelerator measurements, the reactions (7)Li(p,n)(7)Be, (3)H(p,n)(3)He and (2)H(d,n)(3)He were used to produce neutron fields with energy peaks between 144 keV and 19 MeV. The measured fluence responses are 27% too low for thermal energies and show an agreement with approximately 14% for the accelerator produced neutron fields related to the computed fluence responses (MCNP, FLUKA calculations). The measured as well as the computed fluence responses of the dosemeter are compared with the corresponding conversion coefficients.     

Advantages of passive detectors for the determination of the cosmic ray induced neutron environment

Due to the pronounced energy dependence of the neutron quality factor, accurate assessment of the biologically relevant dose requires knowledge of the spectral neutron fluence rate. Bonner sphere spectrometers (BSSs) are the only instruments which provide a sufficient response over practically the whole energy range of the cosmic ray induced neutron component. Measurements in a 62 MeV proton beam at Paul Scherrer Institute, Switzerland, and in the CERN-EU high-energy reference field led to the assumption that conventional active devices for the detection of thermal neutrons inside the BSS, e.g. 6Lil(Eu) scintillators, also respond to charged particles when used in high-energy mixed radiation fields. The effects of these particles cannot be suppressed by amplitude discrimination and are subsequently misinterpreted as neutron radiation. In contrast, paired TLD-600 and TLD-700 thermoluminescence dosemeters allow the determination of a net thermal neutron signal.     

Thermal neutron equivalent dose assessment around the KFUPM neutron source storage area using NTDs. King Fahd University of Petroleum and Minerals

Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li2B4O7) layer for thermal neutron detection via 10B(n,alpha)7Li and 6Li(n,alpha)3H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks x cm(-2) x microSv(-1) using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSvx h(-1) up to 11 microSv x h(-1). The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv x h(-1), which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems.     

A method for evaluating personal dosemeters in workplace with neutron fields

Passive detectors, as albedo or track-etch, still dominate the field of neutron personal dosimetry, mainly due to their low-cost, high-reliability and elevated throughput. However, the recent appearance in the market of electronic personal dosemeters for neutrons presents a new option for personal dosimetry. In addition to passive detectors, electronic personal dosemeters necessitate correction factors, concerning their energy and angular response dependencies. This paper reports on the results of a method to evaluate personal dosemeters for workplace where neutrons are present. The approach here uses few instruments and does not necessitate a large mathematical workload. Qualitative information on the neutron energy spectrum is acquired using a simple spectrometer (Nprobe), reference values for H*(10) are derived from measurements with ambient detectors (Studsvik, Berthold and Harwell) and angular information is measured using personal dosemeters (electronic and bubbles dosemeters) disposed in different orientations on a slab phantom.     

Monte Carlo simulation of the IRSN CANEL/T400 realistic mixed neutron-photon radiation field

The calibration of dosemeters and spectrometers in realistic neutron fields simulating those encountered at workplaces is of high necessity to provide true and reliable dosimetric information to the exposed nuclear workers. The CANEL assembly was set-up at IRSN to produce such neutron fields. It comprises a depleted uranium shell, to produce fission neutrons, then iron and water to moderate them and a polyethylene duct. The new presented CANEL facility is used with 3.3 MeV neutrons. Calculations were performed with the MCNP4C code to characterise this mixed neutron-photon expanded radiation field at the position where calibrations are usually performed. The neutron fluence energy and the direction distributions were calculated and the operational quantities were derived from these distributions. The photon fluence and corresponding ambient dose equivalent were also estimated. Comparison with experimental results showed an overall good agreement.     

Assessment of neutron dosemeters around standard sources and nuclear fissile objects

In order to evaluate the neutron doses around nuclear fissile objects, a comparative study has been made on several neutron dosemeters: bubble dosemeters, etched-track detectors (CR-39) and 3He-filled proportional counters used as dose-rate meters. The measurements were made on the ambient and the personal dose equivalents H*(10) and Hp(10). Results showed that several bubble dosemeters should have been used due to a low reproducibility in the measurements. A strong correlation with the neutron energy was also found, with about a 30% underestimation of Hp(10) for neutrons from the PuBe source, and about a 9% overestimation for neutrons from the 252Cf source. Measurements of the nuclear fissile objects were made using the CR-39 and the dose-rate meters. The CR-39 led to an underestimation of 30% with respect to the neutron dose-rate meter measurements. In addition, the MCNP calculation code was used in the different configurations.     

A prototype personal neutron dosemeter based on an ion chamber and direct ion storage

Ionisation chambers are sensitive to both neutrons and photons. In order to produce a neutron dosemeter based on an ion chamber a double-chamber system which allows for differential readings has to be built. The system consists of one chamber with high neutron sensitivity (e.g. A-150 or polyethylene with 10B or 6Li compounds) and one chamber with low neutron sensitivity (e.g. graphite or Teflon). Different combined dosemeter prototypes were produced and their responses for standard photon and neutron radiation fields, as well as various field spectra, were determined. The feasibility of neutron dosimetry with ion chambers and direct ion storage (DIS) electronics has been proved. The results obtained with prototype dosemeters indicate the system's promising potential for legal approval in the future. Apart from dosimetric properties, the advantages of the system are its small size and weight, easy readout and relatively low production cost.     

Present status of the personal neutron dosemeter based on direct ion storage

In this paper the present status of the Direct Ion Storage Neutron (DIS-N) prototype dosemeter (RADOS) is described. The separation of neutron from photon dose equivalent has been improved by adding tin shieldings. The neutron energy response has been changed by additional plastic covers containing 40% B4C in order to reduce the over-response to thermal neutrons. The responses of the dosemeters were determined for standard photon and neutron fields (monoenergetic neutrons, neutron sources and simulated workplace fields). Irradiations in real workplaces were also performed. The dependence of the neutron response on the angle of incidence was measured for different neutron sources.     

Conceptual design of spectrum changeable neutron calibration fields in JAERI/FRS

Spectrum changeable neutron calibration fields are planned to be established with an accelerator installed in Japan Atomic Energy Research Institute/Facility of Radiation Standards. The neutron fields are provided by bombarding a target surrounded by a moderator, with charged particles from the accelerator. In the fields, a wide variety of neutron spectra is provided with sufficient fluence rate for the calibration of dosemeters. In this study, necessity of the field was first discussed in view of relationship between readings of existing dosemeters and true dose equivalents where the dosemeters were used. Second, test simulation of neutron spectra was carried out with the Monte Carlo technique for some arrangements with a LiF target and quasi-cylindrical moderators with different materials. The simulated spectra were summarised in terms of fluence-average energy, fluence rate and calibration factor for the dosemeters.     

Calibration of the borated ion chamber at NIST reactor thermal column

In boron neutron capture therapy and boron neutron capture enhanced fast neutron therapy, the absorbed dose of tissue due to the boron neutron capture reaction is difficult to measure directly. This dose can be computed from the measured thermal neutron fluence rate and the (10)B concentration at the site of interest. A borated tissue-equivalent (TE) ion chamber can be used to directly measure the boron dose in a phantom under irradiation by a neutron beam. Fermilab has two Exradin 0.5 cm(3) Spokas thimble TE ion chambers, one loaded with boron, available for such measurements. At the Fermilab Neutron Therapy Facility, these ion chambers are generally used with air as the filling gas. Since alpha particles and lithium ions from the (10)B(n,alpha)(7)Li reactions have very short ranges in air, the Bragg-Gray principle may not be satisfied for the borated TE ion chamber. A calibration method is described in this paper for the determination of boron capture dose using paired ion chambers. The two TE ion chambers were calibrated in the thermal column of the National Institute of Standards and Technology (NIST) research reactor. The borated TE ion chamber is loaded with 1,000 ppm of natural boron (184 ppm of (10)B). The NIST thermal column has a cadmium ratio of greater than 400 as determined by gold activation. The thermal neutron fluence rate during the calibration was determined using a NIST fission chamber to an accuracy of 5.1%. The chambers were calibrated at two different thermal neutron fluence rates: 5.11 x 10(6) and 4.46 x 10(7)n cm(-2) s(-1). The non-borated ion chamber reading was used to subtract collected charge not due to boron neutron capture reactions. An optically thick lithium slab was used to attenuate the thermal neutrons from the neutron beam port so the responses of the chambers could be corrected for fast neutrons and gamma rays in the beam. The calibration factor of the borated ion chamber was determined to be 1.83 x 10(9) +/- 5.5% (+/- 1sigma) n cm(-2) per nC at standard temperature and pressure condition.     

Fricke gel dosimetry in boron neutron capture therapy

Gel dosimetry allows three-dimensional (3D) measurement of absorbed dose in tissue-equivalent dosemeter phantoms. Gel phantoms are imaged using optical techniques. In neutron capture therapy (NCT), properly designed gel dosemeters can give 3D dose distributions, due to the various components of the secondary radiation, in phantoms exposed in the thermal or epithermal column of a nuclear reactor. In addition to the therapeutic dose arising from the reaction 10B(n,alpha)7Li, the other dose components are also obtainable, i.e. the gamma dose (due to reactor background and to the reaction 1H(n,gamma)2H of thermal neutrons with hydrogen, the dose due to protons emitted in the reaction 14N(n,p)14C of thermal neutrons with nitrogen and the dose due to recoil protons resulting from elastic scattering of epithermal neutrons.     

Feasibility study on using imaging plates to estimate thermal neutron fluence in neutron-gamma mixed fields

In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field.     

Dose distributions in phantoms irradiated in thermal columns of two different nuclear reactors

In-phantom dosimetry studies have been carried out at the thermal columns of a thermal- and a fast-nuclear reactor for investigating: (a) the spatial distribution of the gamma dose and the thermal neutron fluence and (b) the accuracy at which the boron concentration should be estimated in an explanted organ of a boron neutron capture therapy patient. The phantom was a cylinder (11 cm in diameter and 12 cm in height) of tissue-equivalent gel. Dose images were acquired with gel dosemeters across the axial section of the phantom. The thermal neutron fluence rate was measured with activation foils in a few positions of this phantom. Dose and fluence rate profiles were also calculated with Monte Carlo simulations. The trend of these profiles do not show significant differences for the thermal columns considered in this work.     

Inter-comparison among different TLD-based techniques in a standard multisphere assembly for the characterisation of neutron fields

In the framework of collaboration among the ENEA Radiation Protection Institute (Bologna), the ENEA Fusion Department (Frascati) and the INFN-LNF-Radiation Protection Group (Frascati), an experimental campaign was organised on the usage of thermoluminescence dosemeters (TLDs) for the dosimetric and spectrometric characterisation of neutron fields. Commercially available TLDs of different material and different sensitivity to photons and thermal neutrons were selected, namely TLD600H and TLD700H from Harshaw, GR206 and GR207 from SSDML (China), MCP-6s from TLD Poland. The detectors were first calibrated in standard fields of photons ((60)Co) and thermal neutrons at the ENEA-IRP Secondary Standard Calibration Laboratory of Bologna, then exposed in fast neutron standard fields of different energy, using a standard multisphere moderating assembly. The paper compares the dosimetric characteristics of the studied TL detectors, underlining the (n-gamma) discrimination capability, and discusses their spectrometric performances addressed to radiation protection applications.     

Dose imaging with gel-dosemeter layers: optical analysis and dedicated software

In radiotherapy involving thermal and epithermal neutrons, the knowledge of dose distributions, with separation of the contribution of each secondary radiation component, is of utmost importance. Layers of Fricke-Xylenol-Orange-infused gel dosemeters give the possibility of achieving such requirements because, owing to the layer-geometry, enriching or depleting the gel matrix of suitable isotopes does not sensibly alter neutron transport. The dosimetry method has been critically re-examined with the aim of improving its suitability to boron neutron capture therapy (BNCT) requirements, as it applies to the protocol of measurement and analysis, the sensitivity of the method and the range of the linearity of the dosemeters. Software has been developed and studied to obtain automatically the images of the various dose components with the established separation procedure.     

Characterisation of laboratory-produced CANDU-like workplace neutron fields

Two neutron fields were produced in the Neutron Irradiation Facility (NIF) at the Chalk River Laboratories of the Atomic Energy of Canada Ltd. by directing (d,D) neutrons from a 150 kV neutron generator through a specially designed moderator assembly. Bonner sphere and proton recoil spectrometry systems were used to characterise these fields to determine whether they were CANDU-like*, i.e. whether they resembled neutron fields found in workplaces around pressurised heavy-water moderated power reactors such as CANDU reactors. Similarities were found between the distributions in energy of neutron fluence and ambient dose equivalent of the neutron fields produced in the NIF and those measured previously in power plants. In addition, there was agreement between theoretical (Monte Carlo) data and measured data, thereby validating continued use of Monte Carlo modelling for field characterisations in the NIF. The CANDU-like fields add to the repertoire of neutron fields available in the NIF and are expected to be useful for evaluating neutron dosemeters.     

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.     

Developing a thermal neutron irradiation system for the calibration of personal dosemeters in terms of Hp(10)

At the ENEA Radiation Protection Institute in Bologna a thermal neutron irradiation facility is available for the calibration of neutron dosemeters. It consists of a 1 m x 1 m x 1 m polyethylene cube containing three 241Am-Be sources of about 185 GBq. The cube contains three co-axial cylindrical calibration cavities of different dimension. Due to their limited dimensions, the cavities do not allow the calibration of thermal neutron personal dosemeters in terms of Personal Dose Equivalent Hp(d), that should be carried out on the 30 cm x 30 cm x 15 cm ISO phantom. The study herewith presented was addressed at adapting the facility for external irradiation of personal dosemeters on the ISO phantom. Extensive Monte Carlo studies were carried out to characterise the neutron fluence spatial distribution along the front face of the phantom. A satisfying neutron field homogeneity within the measurement area has been obtained by means of a pyramidal polyethylene fluence flattening filter and the selection of the proper cube to phantom distance. This new irradiation set-up was experimentally tested through measurements with activation foils, according to the spatial mapping array taken from the calculations.