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.     

Development of the fast neutron standard using a Be({alpha},n) reaction at the National Metrology Institute of Japan

This paper describes the 8-MeV neutron field where the neutrons are generated in the (9)Be(alpha,n)(12)C reaction by bombardment of a beryllium target with a 2.4-MeV (4)He(+) beam from a Van de Graaff accelerator. The neutron field is being prepared for a new national standard on neutron fluence in Japan. Absolute measurement of the neutron fluence was taken using a proton recoil neutron detector, consisting of a silicon surface barrier detector with a polyethylene radiator. Neutron spectra were measured using a newly developed recoil proton spectrometer and a liquid organic scintillation detector. The gamma rays existing in the field were also characterised using a liquid organic scintillation detector. The ambient dose equivalents of the gamma rays were estimated to be <100 microSv at the neutron fluence of 10(7) neutrons cm(-2).     

Electronic neutron personal dosimetry with superheated drop detectors

The prototype of an electronic personal neutron dosemeter based on superheated drop detectors is presented. This battery operated device comprises a neutron sensor, bubble-counting electronics and a temperature controller ensuring an optimal dose equivalent response. The neutron sensor is a 12 ml detector vial containing an emulsion of about 50,000 halocarbon-12 droplets of 100 microns diameter. The temperature controller is a low-power, solid-state device stabilising the emulsion at 31.5 degrees C by means of an etched foil heater. The microprocessor controlled counting electronics relies on a double piezo-electric transducer configuration to record bubble formation acoustically via a comparative pulse-shape analysis of ambient noise and detector signals. The performance of the dosemeter was analysed in terms of the requirements presently developed for neutron personal dosemeters. The detection threshold is about 1 microSv, while the personal dose equivalent response to neutrons in the thermal to 62 MeV range falls within a factor 1.6 of 13 bubbles per microSv.     

Characterization of gamma rays existing in the NMIJ standard neutron field

Our laboratory provides national standards on fast neutron fluence. Neutron fields are always accompanied by gamma rays produced in neutron sources and surroundings. We have characterised these gamma rays in the 5.0 MeV standard neutron field. Gamma ray measurement was performed using an NE213 liquid scintillator. Pulse shape discrimination was incorporated to separate the events induced by gamma rays from those by neutrons. The measured gamma ray spectra were unfolded with the HEPRO program package to obtain the spectral fluences using the response matrix prepared with the EGS4 code. Corrections were made for the gamma rays produced by neutrons in the detector assembly using the MCNP4C code. The effective dose equivalents were estimated to be of the order of 25 microSv at the neutron fluence of 10(7) neutrons cm(-2).     

A directional dose equivalent monitor for neutrons

A directional dose equivalent monitor is introduced which consists of a 30 cm diameter spherical phantom hosting a superheated drop detector embedded at a depth of 10 mm. The device relies on the similarity between the fluence response of neutron superheated drop detectors based on halocarbon-12 and the quality-factor-weighted kerma factor. This implies that these detectors can be used for in-phantom dosimetry and provide a direct reading of dose equivalent at depth. The directional dose equivalent monitor was characterised experimentally with fast neutron calibrations and numerically with Monte Carlo simulations. The fluence response was determined at angles of 0, 45, 90, 135 and 180 degrees for thermal to 20 MeV neutrons. The response of the device is closely proportional to the fluence-to-directional dose equivalent conversion coefficient, h'phi (10; alpha, E). Therefore, our monitor is suitable for a direct measurement of neutron directional dose equivalent, H'(10), regardless of angle and energy distribution of the neutron fluence.     

Evaluation of solid state nuclear track detector stacks exposed on the international space station

The aim of the study was to investigate the contribution of secondary neutrons to the total dose inside the International Space Station (ISS). For this purpose solid-state nuclear track detector (SSNTD) stacks were used. Each stack consisted of three CR-39 sheets. The first and second sheets were separated by a Ti plate, and the second and third sheets sandwiched a Lexan polycarbonate foil. The neutron and proton responses of each sheet were studied through MC calculations and experimentally, utilising monoenergetic protons. Seven stacks were exposed in 2001 for 249 days at different locations of the Russian segment 'Zvezda'. The total storage time before and after the exposure onboard was estimated to be seven months. Another eight stacks were exposed at the CERF high-energy neutron field for calibration purposes. The CR-39 detectors were evaluated in four steps: after 2, 6, 12 and 20 h etching in 6 N NaOH at 70 degrees C (VB = 1.34 microm h(-1)). All the individual tracks were investigated and recorded using an image analyser. The stacks provided the averaged neutron ambient dose equivalent (H*) between 200 keV and 20 MeV, and the values varied from 39 to 73 microSv d(-1), depending on the location. The Lexan detectors were used to detect the dose originating from high-charge and high-energy (HZE) particles. These results will be published elsewhere.     

Introduction of a thermal response to the DSTL PADC personal neutron dosemeter

The response of the Defence Science and Technology Laboratory (DSTL) PADC personal neutron dosemeter is strongly dependent upon neutron energy, with a range of 300-500 tracks per cm2 per mSv for energies between 1 and 5 MeV. Below 1 MeV the response drops off sharply. This lack of sensitivity is undesirable when the dosemeter is employed with the softened fission spectra encountered in the workplace. In order to incorporate a thermal response, a polypropylene converter doped with LiF has been placed directly in front of the PADC elements. Tritons produced in the thermal neutron reaction 6Li (n,t)alpha at 2.7 MeV will then penetrate the PADC, leaving a trail of damage. The reaction rate within the converter has been calculated using MCNP for thermal neutrons and a range of higher energies, while transport of the tritons is modelled using the SRIM/TRIM package to determine the resultant track density and depth distribution. The modelling and experimental work have demonstrated that a concentration of 0.2% natural lithium by weight results in a track density in a thermal field comparable with that produced per unit personal dose equivalent by neutrons greater than 1 MeV in the standard dosemeter. Additional MCNP modelling has demonstrated that the dosemeters' albedo response to intermediate energy neutrons can be enhanced considerably by placing a boron-doped shield in front of the converter and increasing its lithium concentration.     

Measurements of neutron dose rates with a balloon in Japan

Measurements of cosmic-ray neutron dose rates with a balloon in Sanriku, Japan (geographic location: 39 degrees N, 142 degrees E; corresponding geomagnetic latitude: 30 degrees N) were conducted at an altitude from 0.2 to 25 km on 25-26 August 2004 when solar activity was at an average level. Neutron dose rates given as ambient dose equivalent rates (H*(10)) were measured with high-sensitive neutron dose equivalent counters and electronic silicon personal dosimeters (EPDs). The neutron dose rates increased with increasing altitude, but they were saturated around 15-20 km and decreased with increasing altitude beyond 20 km. The neutron ambient dose equivalent rate was 1.5 microSv/h(- 1) at 20 km. Measured values were corrected for the deviation of the energy response of the dose equivalent counter from the fluence-to-ambient dose equivalent conversion coefficient, and the corrected values were very close to the calculated values with EPCARD. On the other hand, neutron measurements by the EPDs gave about 10 times overestimation because of the high sensitivity to cosmic-ray protons.     

Space radiation measurements on-board ISS--the DOSMAP experiment

The experiment 'Dosimetric Mapping' conducted as part of the science program of NASA's Human Research Facility (HRF) between March and August 2001 was designed to measure integrated total absorbed doses (ionising radiation and neutrons), heavy ion fluxes and its energy, mass and linear energy transfer (LET) spectra, time-dependent count rates of charged particles and their corresponding dose rates at different locations inside the US Lab at the International Space Station. Owing to the variety of particles and energies, a dosimetry package consisting of thermoluminescence dosemeter (TLD) chips and nuclear track detectors with and without converters (NTDPs), a silicon dosimetry telescope (DOSTEL), four mobile silicon detector units (MDUs) and a TLD reader unit (PILLE) with 12 TLD bulbs as dosemeters was used. Dose rates of the ionising part of the radiation field measured with TLD bulbs applying the PILLE readout system at different locations varied between 153 and 231 microGy d(-1). The dose rate received by the active devices fits excellent to the TLD measurements and is significantly lower compared with measurements for the Shuttle (STS) to MIR missions. The comparison of the absorbed doses from passive and active devices showed an agreement within +/- 10%. The DOSTEL measurements in the HRF location yielded a mean dose equivalent rate of 535 microSv d(-1). DOSTEL measurements were also obtained during the Solar Particle Event on 15 April 2001.     

Improved neutron detection by gamma-ray spectroscopy

Most high-efficiency interactions for neutron detection are with thermal neutrons, and the detection of the prompt gamma radiation emitted as a result of thermal neutron absorption in specific materials was sometimes used. These materials are named converters, the most widespread being 10B, 113Cd or 73Ge (the last two materials are detector materials themselves for CdZnTe and HPGe detectors). However, for these materials, the main gamma-ray energies are relatively low (in the region 473-600 keV), and there are interferences from other gamma rays present in background. Different combinations of moderator-converter-detector systems were employed to optimise the system efficiency and signal-to-background ratio, using B, Cd, Gd and Cl converters and HPGe, NaI(Tl) and BGO detectors. When using a Cl converter, the 1164.7 keV peak seems to be a good choice when a HPGe spectrometer is used. A very practical material containing chlorine is PVC.     

Extremely high dose neutron dosimetry using CR-39 and atomic force microscopy

Atomic force microscopy (AFM) has been applied to the analysis of CR-39 nuclear track detectors for high dose neutron dosimetry. As a feasible study to extract the neutron dose, we have employed a (239)Pu-Be neutron source with the traditional track density measurement of recoil proton etch pits from a high density polyethylene (CH(2)) radiator. After very short etching ( approximately 1 microm), etch pit densities were measured as a function of neutron fluence (neutron dose) up to 1.4 x 10(10) cm(-2) (6.6 Sv). Neutron sensitivity was also measured to be 6.6 x 10(-4). Maximum measurable neutron dose was estimated to be approximately 200 Sv by measuring the fraction of the total image area occupied by the etch pits.     

A comprehensive study of natural gamma radioactivity levels and associated dose rates from surface soils in cyprus

A survey was carried out to determine activity concentration levels and associated dose rates from the naturally occurring radionuclides 232Th, 238U and 40K, in the various geological formations of Cyprus, by means of high-resolution gamma ray spectrometry. A total of 115 representative soil samples were collected from all over the bedrock surface of the island, based on the different lithological units of the study area, sieved through a fine mesh, sealed in 1000 ml plastic Marinelli beakers and measured in the laboratory with respect to gamma radioactivity for a counting time of 18-h each. From the measured spectra, activity concentrations were determined for 232Th (range from 1.0 x 10(-2) to 39.8 Bq kg(-1)), 238U (from 1.0 x 10(-2) to 39.3 Bq kg(-1)) and (40)K (from 4.0 x 10(-2) to 565.8 Bq kg(-1)). Gamma absorbed dose rates in air outdoors were calculated to be in the range from 1.1 x 10(-2) to 51.3 nGy h(-1), depending on the geological features, with an overall mean value of 8.7 nGy h(-1) and a standard deviation of 8.4 nGy h(-1). This value is by a factor of about seven below the corresponding population-weighted world-averaged value of 60 nGy h(-1) and one of the lowest that has been reported from similar investigations worldwide. Assuming a 20% occupancy factor, the corresponding effective dose rates outdoors equivalent to the population were calculated to be between 1.3 x 10(-2) and 62.9 microSv y(-1), with an arithmetic mean value of 10.7 microSv y(-1) and a standard deviation of 10.3 microSv y(-1).     

Evaluation of absorbed dose rate and annual effective dose equivalent due to terrestrial gamma radiation in rocks in a part of Southwestern Nigeria

The average outdoor absorbed dose rate in air and the average annual effective dose equivalent due to terrestrial gamma radiation from 40K, 238U and 232Th in rocks in Ondo and Ekiti States, Southwestern Nigeria have been evaluated from measurements of the concentrations of these radionuclides in this environmental material. The concentration measurements were obtained using a very sensitive gamma spectroscopic system consisting of a 7.6 cm x 7.6 cm NaI(Tl) scintillation detector coupled to a computerised ACCUSPEC installation. The average absorbed dose rate and average annual effective dose equivalent was found to be 8.33 +/- 2.76 nGy x h(-1) and 8.7 +/- 2.9 microSv x y(-1) respectively.     

Characterisation of neutron fields around high-energy x-ray radiotherapy machines

Photoneutron spectra around the treatment bed of a Varian Clinac 2100C machine were measured using a Bonner sphere spectrometer. To overcome problems with pulse pile-up and detection of non-neutron-induced events, the active detector of thermal neutrons normally used at the centre of the spheres was replaced by a sandwich of four CR-39 track detectors interleaved with 10B radiators. Track densities measured for the CR-39 detectors in Bonner spheres were used for the unfolding of neutron spectra. Neutron fluence and ambient dose equivalent for the whole energy range and partial energy intervals were derived from the neutron spectra.     

Induced radioactivity in the forward shielding and semiconductor tracker of the ATLAS detector

The radioactivity induced in the forward shielding, copper collimator and semiconductor tracker modules of the ATLAS detector has been studied. The ATLAS detector is a long-term experiment which, during operation, will require to have service and access to all of its parts and components. The radioactivity induced in the forward shielding was calculated by Monte Carlo methods based on GEANT3 software tool. The results show that the equivalent dose rates on the outer surface of the forward shielding are very low (at most 0.038 microSv h(-1)). On the other hand, the equivalent dose rates are significantly higher on the inner surface of the forward shielding (up to 661 microSv h(-1)) and, especially, at the copper collimator close to the beampipe (up to 60 mSv h(-1)). The radioactivity induced in the semiconductor tracker modules was studied experimentally. The module was activated by neutrons in a training nuclear reactor and the delayed gamma ray spectra were measured. From these measurements, the equivalent dose rate on the surface of the semiconductor tracker module was estimated to be < 100 microSv h(-1) after 100 d of Large Hadron Collider (LHC) operation and 10 d of cooling.     

Radionuclide neutron sources in calibration laboratory--neutron and gamma doses and their changes in time

The calibration laboratory, having standard neutron fields of radionuclide sources, should perform regular measurements of fields' parameters in order to check their stability and to get knowledge of any changes. Usually, accompanying gamma radiation is not of serious concern, but some personal dosemeters, old neutron dose equivalent meters with scintillation detectors and the dose meters of mixed radiation require the determination of this component. In the Laboratory of Radiation Protection Measurements in the Institute of Atomic Energy, Poland, the fields of radionuclide neutron sources (252)Cf, (241)Am-Be and (239)Pu-Be were examined for nearly 20 y. A number of detectors and methods have been applied for the determination of neutron ambient dose equivalent rate and for the determination of neutron and gamma dose components. This paper presents the recent results of measurements of gamma and neutron dose and dose equivalent, compared with the results accumulated in nearly 20 y.     

Active dosimetry on recent space flights

The radiation exposure inside the spacecraft in low earth orbit was investigated with a telescope based on two silicon planar detectors during three NASA shuttle-to-MIR missions (inclination 51.6 deg, altitude about 380 km). Count and dose rate profiles were measured, as well as separate linear energy transfer (LET) spectra, for the galactic cosmic rays (GCR) and the trapped radiation encountered in the South Atlantic Anomaly (SAA). Effective quality factors are deduced from the converted LET spectra (in water) in the range 0.1-120 keV micrometer-1 according to ICRP 60. Measured mission averaged dose rates in silicon are in the range 98-108 microGy d-1 and 137-178 microGy d-1 for the GCR and SAA contributions, respectively. The deduced effective quality factors are 2.95-3.29 (GCR) and 1.18-1.25 (SAA), resulting in mission averaged dose equivalent rates of 631-716 microSv d-1 for the comparable three missions.     

Investigations of doses on board commercial passenger aircraft using CR-39 and thermoluminescent detectors

Measurements of cosmic radiation dose rates (from the neutron and the non-neutron components) on board passenger aircraft were performed using environmental packages with thermoluminescent TL and CR-39 etched track detectors. The packages were calibrated at the CERN-EU high-energy Reference Field Facility and evaluated at the Institute of Nuclear Physics in Krakow (TL + CR-39) and at the German Aerospace Centre in Cologne (CR-39). Detector packages were exposed on board passenger aircraft operated by LOT Polish Airlines, flown between February and May 2001. The values of effective dose rate determined, averaged over the measuring period, ranged between 2.9 and 4.4 microSv h(-1). The results of environmental measurements agreed to within 10% with values calculated from the CARI-6 code.     

Radiation protection for an ultra-high intensity laser

Radiological characterisation of an experimental chamber and other areas of an ultra-high intensity laser facility (-terawatt) revealed significant levels of X ray, gamma and neutron radiation. Different techniques were used to detect and measure this radiation: TLD. photographic film, bubble detectors and germanium spectrometry. A test series of radiological measurements was made for 150 laser shots (300 femtoseconds) with energies in the 1 to 20 J range and a target illuminance of 10(19) W.cm2. Gamma dose equivalents in the vicinity of the chamber varied between 0.7 and 73 mSv. The dose equivalent due to the neutron component was evaluated to be 1% of the gamma dose equivalent. The amount of radiation generated depends on the laser energy and the nature of the target. No activation or contamination of the chamber or target holder were observed. Ultra-high intensity lasers are being extensively developed at the present time and the investigations performed demonstrate that it is necessary to take radiological risks into consideration in the design of ultra-high intensity laser facilities and to define personnel access conditions.     

Individual monitoring in mixed neutron/photon fields using a single silicon detector

A device based on a single silicon detector of special converter/detector design optimised for the determination of the neutron dose equivalent is also used for the determination of the photon dose equivalent. While the neutron dose is determined on the basis of signals corresponding to energy depositions above 1.5 MeV, depositions between 80 keV and 150 keV are used for the photon dose equivalent. In this way, a photon response is achieved which varies by less than 30% in the energy region from 80 keV to 7 MeV for irradiation at normal incidence and at 60 degrees to the normal. The lower limit of detection is of the order of 1 microSv. Neutrons contribute to the photon reading by less than 2% in mixed fields with a comparable dose equivalent from neutrons and photons.