Air kerma rate constants for gamma emitters used most often in practice

It is often required to estimate the dose rate at a distance from radionuclides that are sources of X rays and gamma rays. Such calculations may be required for planning radiation protection measures in the vicinity of radioactive sources or patients containing radionuclides, calibrations of radiation instruments or for estimating the absorbed dose rate to patients receiving brachytherapy. The factor relating activity and air kerma rate is called air kerma rate constant--gamma(delta). In this paper, the results of recalculation of this quantity for unfiltered point sources of radionuclides in practice used most often are given. The calculations included corrections for internal conversion of X rays and gamma rays and detailed accounting of the generation of the K and L series X rays from internal conversion and electron capture. Particular air kerma rate constants were calculated for each discrete line in the photon spectrum of radionuclide with a yield per decay event >0.01% and the energy >20 keV. Since the energy structure of the photon spectra and accessible discrete numerical values of the mass energy-transfer coefficient for air are not the same, the cubic spline interpolation was used to obtained the coefficient, where the photon spectrum data are available. In the calculation, the latest gamma ray spectral data for all radionuclides and latest data for the mass energy-transfer coefficient for air are used. Air kerma rate constants for the following 35 radionuclides are calculated: 11C, 13N, 15O, 18F, 24Na, 42K, 43K, 51Cr, 52Fe, 59Fe, 57Co, 58Co, 60Co, 67Ga, 68Ga, 75Se, 99Mo, 99mTc, 111In, 113mIn, 123I, 125I, 131I, 127Xe, 133Xe, 137Cs, 152Eu, 154Eu, 170Tm, 182Ta, 192Ir, 197Hg, 198Au, 201Tl and 241Am.     

Materials and processing of vacuum components for the high-intensity proton beam accelerator, J-PARC

The Japan Proton Accelerator Research Complex (J-PARC) project of the proton beam accelerator, involving a 400-MeV linac, a 3-GeV rapid-cycling synchrotron and a 50-GeV synchrotron, started in 2002. Materials used as vacuum components, such as electroformed copper, titanium, stainless steel and alumina ceramics, were examined from the point of view outgassing and electrical breakdown. The mechanical characteristics of a hydro-formed titanium bellows and a quick-disconnect flange system were also investigated.     

Induced radioactivity in rare-earth permanent magnets at 2.5 GeV electron accelerator

Rare-earth permanent magnets such as Nd₂Fe₁₄B and Sm₂Co₁₇ are the main components of the insertion devices of synchrotron radiation facilities and are used at other particle accelerators. Due to radiation safety concern the induced radioactivity in permanent magnets at a 2.5 GeV electron accelerator were studied with Monte Carlo simulations and experiments. The saturated activity of each radioactive isotope was estimated with FLUKA code (2005 version) and compared with measurement results. Three models of NEOMAX Nd₂Fe₁₄B magnets and one trivial Sm₂Co₁₇ magnet were chosen as the test magnets. The remnant dose rate at 1 m from the magnet and its cooling time dependence are calculated with the induced activities from the view of radiation protection. These are investigated under unique radiation environments due to different target conditions. The transverse distributions of induced radioactivities of several representative isotopes, which were measured at different target conditions, agree well with the electromagnetic shower characteristics at different target conditions as well as the spatial distributions of photon and neutron fluences.     

Determination of natural radioactivity in building materials used in Tunisian dwellings by gamma ray spectrometry

The radioisotopic content of 17 samples of natural and manufactured building materials collected in Tunisia have been analysed by using gamma spectrometry. From the measured gamma ray spectra, activity concentrations are determined for (232)Th, (226)Ra, (235)U and (40)K. The total effective dose and the activity concentration index are calculated applying the dose criteria recommended by the European Union for building materials. The results of (226)Ra, (232)Th and (40)K found in Tunisian building materials indicate that radium and thorium concentrations do not exceed 40 Bq kg(-1), but potassium concentration varies between 50 and 1215 Bq kg(-1). The total effective dose rates per person indoors are determined to be between 0.07 and 0.86 mSv y(-1). Only two materials exceed the reference level of 0.3 mSv y(-1). The activity concentration index is <1.     

Natural radioactivity and external gamma radiation exposure at the coastal Red Sea in Egypt

Radionuclides which present in different beach sands are sources of external exposure that contribute to the total radiation exposure of human. In this work, superficial samples of beach sand were collected from the Red Sea coastline (Ras Gharib, Hurghada, Safaga, Qusier and Marsa Alam areas) and at 20 km on Qena-Safaga road. The distribution of natural radionuclides in sand beach samples was studied by gamma spectrometry. The activity concentrations of primordial and artificial radionuclides in samples that are collected from the coastal environment of the Red Sea were 19.2 +/- 3 Bq kg(-1) for (210)Pb, 21.1 +/- 1 Bq kg(-1) for (226)Ra, 22.7 +/- 2 Bq kg(-1) for (238)U, 1.0 +/- 0.1 Bq kg(-1) for (235)U, 11.6 +/- 1 Bq kg(-1) for (228)Ra, 13.0 +/- 1 Bq kg(-1) for (228)Th, 12.4 +/- 1 Bq kg(-1) for (232)Th, 930 +/- 32 Bq kg(-1) for (40)K and 1.2 +/- 0.3 Bq kg(-1) for (137)Cs. The mean external gamma-dose rate was 62.5 +/- 3.2 nSv h(-1), 54.4 +/- 2.8 nGy h(-1) Ra equivalent activity (Ra(eq)) was 107 +/- 5.8 Bq kg(-1), 0.86 +/- 0.04 Bq kg(-1) for representative level index (I(gamma)) and effective dose rate was 0.067 +/- 0.003 mSv y(-1) in beach sand red sea, in air due to naturally occurring radionuclides.     

Propagation characteristics of neutrons leaking from the accelerator facilities

In this study spatial and time distribution of neutrons leaking from Hokkaido University 45 MeV Electron Linac facility have measured and compared with the Monte Carlo simulations. The neutron transport processes inside and outside the facility building has been simulated using MCNP. The neutrons have measured by BF3 counters and 3He counters with polyethylene moderators up to the distance of 330 m from the facility. The spatial distribution of ambient dose equivalent converted from the counts has been compared with the simulation. The distribution estimated from the counts by the BF3 counter has been shown fairly good agreement with the calculation. The spatial distribution of counts obtained at the 45 MeV Electron Linac facility has been compared with that obtained at the Fusion Neutronics Source (FNS) facility of JAERI. The difference between the propagation characteristics of neutrons leaking from those facilities has been discussed.     

Natural radioactivity levels of granites used in Turkey

Thirty granite samples commonly used in Turkey were surveyed for natural radioactivity. Concentrations of natural radionuclides in all samples were determined by gamma-ray spectroscopy with hyper-pure germanium detector. The activity concentrations measured for (226)Ra and (232)Th ranged from 0.7±0.1 to 186±1 Bq kg(-1), and from 0.5±0.1 to 249±2 Bq kg(-1), respectively. The activity concentrations obtained for (40)K varied from minimum detectable activity (0.4 Bq kg(-1)) to 1935±11 Bq kg(-1). The radium equivalent activity (Ra(eq)), the absorbed dose rate (D), the external hazard index (H(ex)) and the annual effective dose equivalent were also calculated and compared with the international recommended values. Granite samples were also analysed mineralogically. It was observed that the presence of large amount orthoclase and radiogenic accessory minerals are the sources of high activity concentration levels.     

Evaluation of internal and external doses from 11C produced in the air in high energy proton accelerator tunnels

Air has been irradiated with high energy protons at the 12 GeV proton synchrotron to obtain the following parameters essential for the internal dose evaluation from airborne 11C produced through nuclear spallation reactions: the abundance of gaseous and particulate 11C, chemical forms, and particle size distribution. It was found that more than 98% of 11C is present as gas and the rest is aerosol. The gaseous components were only 11CO and 11CO2, and their proportions were approximately 80% and 20%, respectively. The particulate 11C was found to be sulphate and/or nitrate aerosols having a log-normal size distribution; the measurement using a diffusion battery showed a geometric mean radius of 0.035 micron and a geometric standard deviation of 1.8 at a beam intensity of 6.8 x 10(11) proton.pulse-1 and an irradiation time of 9.6 min. By taking the chemical composition and particle size into account, effective doses both from internal and from external exposures per unit concentration of 11C were calculated for various room sizes. The values can be used to evaluate the effective dose from the airborne 11C produced in the accelerator tunnels.     

Natural radioactivity content of granite tiles used in Greece

Measurements of (226)Ra, (232)Th and (40)K activity concentrations in commercial granite tiles imported in Greece were performed using gamma-ray spectrometry. The activity concentration of (226)Ra, (232)Th and (40)K ranged from 1 to 434, 2 to 239 and 71 to 1576 Bq kg(-1), respectively. The calculated activity concentration index (I) values for all granite samples examined were found to be within the EC limit values for superficial and other materials with restricted use.     

The problems associated with the monitoring of complex workplace radiation fields at European high-energy accelerators and thermonuclear fusion facilities

The European Commission is funding within its Sixth Framework Programme a three-year project (2005-2007) called CONRAD, COordinated Network for RAdiation Dosimetry. The organisational framework for this project is provided by the European Radiation Dosimetry Group EURADOS. One task within the CONRAD project, Work Package 6 (WP6), was to provide a report outlining research needs and research activities within Europe to develop new and improved methods and techniques for the characterisation of complex radiation fields at workplaces around high-energy accelerators, but also at the next generation of thermonuclear fusion facilities. The paper provides an overview of the report, which will be available as CERN Yellow Report.     

Analysis of induced radionuclides in low-activation concrete (limestone concrete) using the 12 GeV proton synchrotron accelerator facility at KEK

22Na is one of the long-lived radionuclides induced in shielding concrete of a beam-line tunnel of a high-energy particle accelerator facility and poses a problem of radiation wastes at the decommissioning of the facility. In order to estimate the 22Na concentration induced in shielding concrete, chemical reagents such as NaHCO3, MgO, Al203, SiO2 and CaCO3 were irradiated at several locations in the beam-line tunnel of the 12 GeV proton synchrotron accelerator at KEK, and the 22Na concentrations induced in those chemical reagents were measured. Low-activation concrete made up of limestone aggregates was also irradiated by secondary particles in the beam-line tunnel and the long-lived radionuclide, such as 22Na, concentrations induced in the concrete were measured. It was confirmed that 22Na concentrations induced in Mg, Al, Si and Ca were lower than that in Na, and that 22Na concentrations induced in the low-activation concrete was lower than those induced in ordinary concrete made up of sandstone aggregates.     

Induced radioactivity in CU targets produced by high-energy heavy ions and the corresponding estimated photon dose rates

Irradiation experiments were performed at the Heavy Ion Medical Accelerator in Chiba (HIMAC) facility, National Institute of Radiological Sciences. The radioactive spallation products in a thick Cu target were obtained for Ar(230, 400 MeV per nucleon), Si(800 MeV per nucleon), Ne(100, 230, 400 MeV per nucleon), C(100, 230, 400 MeV per nucleon), He(100, 230 MeV per nucleon), p(100, 230 MeV) ions. The gamma-ray spectra from irradiated Cu samples inserted into the composite Cu target were measured with a high-purity germanium (HPGe) detector. From the gamma-ray spectra, we obtained the spatial distribution of radioactive yields of spallation products of 40 nuclides in the Cu sample in the Cu target. From the spatial distribution of radioactive yields, we estimated the residual activity and photon dose induced in the Cu target. The residual activity and photon dose become larger with the increase in projectile energy per nucleon and the range of the projectile beam for the same projectile energy per nucleon.     

Radiation streaming experiment through a labyrinth of the 12 GeV proton accelerator facility at KEK(2)--TLD REM-counter method

The doses of radiation streaming through a labyrinth were measured using thermoluminescence dosemeters (TLDs) and neutron moderators for TLDs at the neutrino beam line of the 12 GeV proton accelerator facility of High Energy Accelerator Research Organization (KEK). The calculated doses using the Monte Carlo code, MCNPX basically agreed with the experimental results. However, unexpectedly, the calculated neutron doses were smaller than the measured ones along the upstream side of the labyrinth.     

Fatigue of materials used in microscopic components

This paper discusses the fatigue behaviour of materials used in microscopic components, which are defined as components in which the section size is between one and ten grain diameters. Experimental data were obtained on microscopic and conventional (macroscopic) notched specimens of 316L stainless steel, which is used in many biomedical components including the cardiovascular stent. Microscopic specimens showed unusual behaviour: low threshold values and very flat stress‐life(S‐N) curves. The low threshold values were attributed to the extremely small thickness, which may be preventing closure; a secondary contribution may be coming from a change in the length of non‐propagating cracks. Good predictions could be achieved using the Smith–Miller approach and the theory of critical distances (TCD) if this reduced threshold was taken into account or if, alternatively, the critical distance was made equal to the grain size.     

Evaluation of self-dose S values for positron emitters in voxel phantoms

In recent years, concern has been raised about patient-specific doses due to the increased use of PET imaging. It is of scientific and practical interest for the accurate quantification of the doses to patients administered a radiopharmaceutical for PET imaging. In the present study, Monte Carlo simulations have been carried out to evaluate S values--the mean absorbed dose to the target region per unit cumulated activity in the source region--for positron emitters in the brain, heart and the urinary bladder content. The positron emitters considered were four radionuclides that are frequently used for PET imaging: 11C, 13N, 15O and 18F. S values were evaluated for the Medical Internal Radiation Dose 5 type mathematical phantom and the adult voxel phantoms developed at the Japan Atomic Energy Research Institute (JAERI). Consequently, Monte Carlo simulation and voxel phantoms were found to be useful in the evaluation of the beta dose to organs, in particular to hollow organs such as the urinary bladder. It was also demonstrated that the self-dose S values largely depend on the beta-ray energy and the mass of each target region.     

Evaluation of the photoneutron field produced in a medical linear accelerator

The doses and spectra of photoneutrons produced in a medical linear accelerator with photon energies of 10 and 15 MV were evaluated. The Monte Carlo code, MCNPX, was used to simulate the transport of these photoneutrons around the head for 10 and 15 MV photons. The fully-described geometry of the accelerator head was used in this calculation. The photoneutron energy spectra and doses for various photon field sizes were calculated at each of 20 positions. The results indicate that the maximum dose equivalents are observed in 20 x 20 cm(2) case among photon fields. It was found the neutron average energy at isocenter for a 0 x 0 cm(2) field is 0.38 MeV for 10 MV and is 0.45 MeV for 15 MV. The neutron doses at 10 positions around the head in the treatment room of the operation facility at 10 and 15 MV were measured using the bubble detectors. Measurements were compared with the calculations under the same geometry in the experiment. It was found that the majority of the calculated results agreed to within the standard deviations of the measurements. These above results can be applied in the verification of maximum allowed neutron leakage percentage of treatment dose defined in the IEC. We have been employing them to derive the empirical formula for neutron dose equivalent level at the maze entrance of medical accelerator treatment rooms in a study that is still underway.     

Assessment of natural radioactivity concentrations and gamma dose rate levels in Kayseri, Turkey

This study assesses the level of background radiation for Kayseri province of Turkey. Natural radionuclide activity concentrations in soil samples were determined using high-resolution gamma spectroscopy. Outdoor gamma dose measurements in air at 1 m above ground level were determined using a portable gamma scintillation detector. The obtained results of uranium and thorium series as well as potassium ((40)K) are discussed. The present data were compared with the data obtained from different areas in Turkey. From the measured gamma-ray spectra, the average activity concentrations were determined for (238)U (range from 3.91 to 794.25 Bq kg(-1)), (232)Th (range from 0.68 to 245.55 Bq kg(-1)) and (40)K (range from 23.68 to 2718.71 Bq kg(-1)). The average activity concentrations of (238)U, (232)Th and (40)K were found to be 35.51, 37.27 and 429.66 Bq kg(-1), respectively, and 11.53 Bq kg(-1) for the fission product (137)Cs. The highest values of (238)U, (232)Th and (40)K concentrations (794.25 Bq kg(-1), 245.55 Bq kg(-1), 2718.71 Bq kg(-1), respectively) were observed in abnormal samples at Hayriye village. The average outdoor gamma dose rate in air at 1 m above the ground was determined as 114.43 nGy h(-1). Using the data obtained in this study, the average annual effective dose for a person living in Kayseri was found to be about 140.34 μSv.     

燃料电池用质子电解质的研究进展

目前,作为燃料电池的重要部件的质子电解质主要分为有机、有机一无机复合、无机三类.有机质子电解质接近商业化程度,但其价格昂贵,中温导电性差;复合质子电解质是以前者为基础掺杂增湿组分,使燃料电池性能有一定程度的增强;提高工作温度是解决燃料电池中催化荆CO中毒和提高燃料转化效率的有效办法,因此能在中高温环境中工作的无机质子导电材料已成为研究热点之一.