Photoneutron spectra around an 18 MV LINAC

Photoneutron spectra around an 18 MV LINAC were calculated in order to observe the effect produced by media around the accelerator. Calculations were carried out with MCNP 4C code, three different cases were analyzed: Head model, Head and phantom model, and Head, air, phantom and wall model. The spectra were calculated in five detectors located at the irradiation room at different distances from the isocentre. A sixth detector, located near the entrance door was included to analyze how the maze change the neutron spectrum. Neutrons are mainly produced in the LINAC head change the shape of evaporation neutrons from the source term, some of these neutrons leak out the head with lesser energy, another neutrons goes with the treatment beam. At any site near the isocentre neutron spectrum has evaporation and thermal neutrons joined by a set of epithermal neutrons. As the distance from the isocentre increases evaporation neutrons tend to decrease while, epithermal and thermal neutrons tend to remain constant regardless de distance due to room return produced by the walls. The maze contributes to reduce the neutron fluence, reducing the evaporation neutrons; resulting spectrum is mainly the contribution of thermal and epithermal neutrons. Near the door these neutrons can produce activation and prompt gamma rays.     

Measurement of photoneutrons in the output of 15 MV varian clinac 2100C LINAC using bubble detectors

Photoneutron contamination in the output of Varian Clinac 2100C medical linear accelerator (LINAC) operated at 15 MV photon beam energy has been investigated using bubble detectors. Photoneutrons are produced from the photo-disintegration reaction of photons with materials of the head components with threshold energy of approximately 8 MeV. Measurements were conducted in the patient plane at 100 cm source-to-detector distance on beam axis and at stipulated distances outside the irradiated field for 5×5 cm² to 40×40 cm² field sizes for in-air and 5×5 cm² to 20×20 cm² for water phantom measurements. Neutron dose equivalent of 1.57±0.10 mSv·Gy⁻¹ was measured for 10×10 cm² field size for in-air. For in phantom, neutron dose equivalent of 1.42 mSv·Gy⁻¹ was measured for 10×10 cm² field size on the beam axis at a depth of 1 cm but independent of field size at depth >5 cm.     

Neutron beam preparation with Am–Be source for analysis of biological samples with PGNAA method

Material analysis with prompt gamma neutron activation analysis (PGNAA) requires a proper geometrical arrangement for equipments in laboratory. Application of PGNAA in analysis of biological samples, due to small size of sample, needs attention to the dimension of neutron beam. In our work, neutron source has been made of ²⁴¹Am–Be type. Activity of ²⁴¹Am was 20 Ci which lead to neutron source strength of 4.4 × 10⁷ neutrons per second. Water has been considered as the basic shielding material for the neutron source. The effect of various concentration of boric acid in the reduction of intensity of fast and thermal components of the neutron beam and gamma ray has been investigated. Gamma ray is produced by (α, n) reaction in Am–Be source (4.483 MeV), neutron capture by hydrogen (2.224 MeV), and neutron capture by boron (0.483 MeV). Various types of neutron and gamma ray dosimeters have been employed including BF₃ and NE-213 detectors to detect fast and thermal neutrons. BGO scintillation detector has been used for gamma ray spectroscopy. It is shown that the gamma and neutron radiation dose due to direct beam is of the same magnitude as the dose due to radiation scattered in the laboratory ambient. It is concluded that 14 kg boric acid dissolved in 1,000 kg water is the optimum solution to surround the neutron source. The experimental results have been compared with Monte Carlo simulation.     

Assessment of radon concentration and external gamma radiation level in the environs of Narwapahar uranium mine,India and its radiological significance

In the environs of uranium mining, milling and processing facilities and in the uranium mineralized terrain, a little higher ambient radon concentration and gamma radiation level may be expected in comparison with natural background. The present study gives a brief account of atmospheric radon concentration, gamma absorbed dose rate and radiation dose received by the members of public in the vicinity of Narwapahar uranium mine. The ambient radon concentration in the air in the study area was found to vary from 5 to 107 Bq m⁻³ with geometric mean of 24 Bq m⁻³ and geometric standard deviation of 1.74 Bq m⁻³. The measured gamma absorbed dose rate in air at 1 m above the ground ranged from 87 to 220 nGy h⁻¹ with an overall arithmetic mean of 128 ± 18.5 nGy h⁻¹. The mean annual effective dose received by the members of public from inhalation of radon and its progeny and external gamma exposure was estimated to be 0.32 mSv year⁻¹, which is comparable to other reported values elsewhere.     

Natural <Superscript>3</Superscript>H radioactivity analysis in groundwater and estimation of committed effective dose due to groundwater ingestion in Varahi and Markandeya river basins,Karnataka State,India

The present study aimed at the assessment of natural tritium radioactivity in groundwater, being used for domestic and irrigation purposes in Varahi and Markandeya river basins. The study also intended to assess human health risk by estimating committed effective dose due to groundwater ingestion in the study area, taking into consideration the obtained tritium activity concentrations and annual water consumption. Tritium concentration of groundwater samples from the Varahi and Markandeya river basins were determined by liquid scintillation counting and the results laid in the range of 1.95 ± 0.25 to 11.35 ± 0.44 TU and 1.49 ± 0.75 to 9.17 ± 1.13 TU in Varahi and Markandeya river basins, respectively. Majority of the samples from Varahi (46.67%) and Markandeya (62.5%) river basins belong to modern water category aged between 5 and 10 years, while the remaining 53.33% and 37.5% of the samples from Varahi and Markandeya river basins respectively belong to sub-modern water with modern recharge, significantly influenced by precipitation and river inflowing/sea water intrusion. The effective committed dose for general public consumption considering the highest concentration value of 0.02 μSv year⁻¹, which is very negligible compared to EPA (0.04 mSv year⁻¹), WHO (0.1 mSv year⁻¹), ICRP (1.0 mSv year⁻¹) and UNSCEAR (2.4 mSv year⁻¹) recommended dose limits, should not mean any additional health risk for the population living nearby.     

Assessment of the doses due to natural radionuclide in the green leafy vegetables of Domiasiat,Meghalaya India

A radiation dose assessment exercise was carried out for the Ipomea batata, Allium sativum, Dacaus carota, and Solanum tuberosum due to naturally available radionuclide ⁴⁰K, ²³⁸U and ²³²Th in the Domiasiat area of Meghalaya. The concentration of radionuclides in biota as well as corresponding soil was measured by precipitation method using NaI detector for continuous 12 months. Transfer factor was calculated and was, for ⁴⁰K(3.96E−05, 3.40E−05, 3.40E−05, 2.70E−05), for ²³²Th(3.94E−05, 3.20E−05, 3.20E−054.93E−05), for ²³⁸U(3.60E−05, 3.89E−05, 3.85E−054.57E−05), respectively in each biota due to each radionuclide. The point source dose distribution (source ↔ target) hypotheses was applied for the consideration of absorbed fraction. The generated data were modeled using the FASSET method and obtained dose was 8.42E−03, 8.36E−03, 7.78E−03, 7.74E−03 μGy h⁻¹, respectively and finally compared with the IAEA and UNSCEAR dataset for screening level dose for terrestrial biota.     

Design and construction of a facility for neutron activation analysis using the 14 MeV neutron generator at HMS Sultan

The potential for using a small, sealed tube, DT neutron generator for neutron activation analysis has been well documented but not well demonstrated, except for 14 MeV activation analysis. This paper describes the design, construction and characterization of a neutron irradiation facility incorporating a small sealed tube DT neutron generator producing 14 MeV neutrons with fluence rates of 2·10⁸ s⁻¹ in 4π (steady state) and 10¹¹ s⁻¹ in 4π (pulsed). Monte Carlo modeling using MCNP4c and McBend9 has been used to optimize the design of this facility, including the location of a thermal irradiation facility for conventional neutron activation analysis. A significant factor in designing the facility has been the requirement to conform with Ionising Radiation Regulations and the design has been optimized to keep potential radiation doses to less that 1 μSv/h at the external walls of the facility. Activation of gold foils has been used for flux characterization and the experimental results agree well with the modeling.     

Assessment of natural radioactivity in silt samples from Moticher lake near Kakrapar Atomic Power Station,India

The naturally occurring radioisotopes such as ²³⁸U, ²³²Th and ⁴⁰K in the silt samples collected from Moticher lake, Gujarat were evaluated. The activity of ²³⁸U, ²³²Th and ⁴⁰K was found to be 4.4-9.7 Bq kg⁻¹ with a mean 6.4 ± 1.3 Bq kg⁻¹, 10.5–21.2 Bq kg⁻¹ with a mean 15.6 ± 2.5 Bq kg⁻¹ and 102–231 Bq kg⁻¹ with a mean 160 ± 40 Bq kg⁻¹, respectively. The depth profile study could not reveal any significant vertical correlation on radioactivity levels of ²³⁸U, ²³²Th and ⁴⁰K. Radium equivalent activity (Ra_eq), external hazard index (H _ex) and internal hazard index (H _int) were calculated by using the activity of ²³⁸U or ²²⁶Ra, ²³²Th and ⁴⁰K in silt samples. Annual Effective Dose Equivalent (AEDE) level in Moticher silt was found to be 24.8 ± 5.0 μSv year⁻¹, which is much lower than the worldwide average value. The relative contribution to dose due to ²³⁸U and ²³²Th series were found to be 14% and 53%, followed by the contribution of 33% due to ⁴⁰K.     

H*(10) and neutron spectra around linacs

Neutron spectrum and ambient dose equivalent has been measured around two 10 MV linear accelerators. Measurements were carried out in Mevatron and Primus model linacs, both made by Siemens. Main differences between those models are the beam collimator and the vault room. Here, Bonner sphere spectrometer with a passive thermal neutron detector has been utilized to measure the neutron spectrum inside the vault. Using an active detector the neutron spectrum was measured at the vault’s door of both accelerators. With a neutron area monitor the dose equivalent was measured by the door at both vault doors. Neutron strength, total fluence rate and ambient dose equivalent were compared, from this was found that shielding conditions are better in the Primus model.     

Evaluation of imaging plate technique coupled with activation detector as the passive neutron monitor

The spatial distribution of neutrons was measured at the muon science laboratory of KEK by the activation detector method using an imaging plate for the radioactivity measurement. It was confirmed that this method is highly sensitive to detect the average neutron dose of 10 μSv/h. The distribution of thermal and epithermal neutrons was also measured in the experimental room. The cadmium ratio inside the experimental room is one except for the neutron leakage point. The spatial distribution of neutrons inside the concrete shield of KENS was measured by the same method. Aluminum and gold foils were used for the measurement of fast and thermal neutrons, respectively. Two dimensional change of the reaction rate of the ²⁷Al(n,α)²⁴Na reaction shows a good agreement with the results calculated by the Monte Carlo simulation using MARS14 code. Thermal and epithermal neutron flux ratio on the beam axis was measured by the cadmium ratio method. The flux ratios were about 30 and almost constant for every slot except for the surface of the shield, because the cadmium ratio is 2. This method was very useful to measure the activity of many pieces of detector simultaneously without any efficiency and decay correction. Wide dynamic range and high sensitivity are also the merit of this method.     

Evaluation of stray neutron distribution in medical cyclotron vault room by neutron activation analysis approach

Stray neutron distribution in a medical cyclotron vault room was evaluated by neutron activation analysis (NAA). Neutrons were generated in the production of radioactive nuclides, such as ¹⁸F, ¹¹C, ¹³N and ¹⁵O, for diagnostic usage. Indium foil was adopted to evaluate the stray fast and thermal neutron intensity based on ¹¹⁵In(n_f, n′)^115mIn and ¹¹⁵In(n_th, γ)^116m1In reactions, respectively. The indium foils were weighed, sealed and placed at 62 points around the 6.7×8.2 m² cyclotron room. Additionally, each indium foil was exposed for over 80 minutes during cyclotron operation and γ-peaks were analyzed using an HPGe detector to evaluate the number of stray fast (Φ _f) or thermal (Φ _th) neutrons. The minimum to maximum numbers of fast and thermal neutrons were (3.47±0.11)×10³ to (1.06±0.21)×10⁴ n·cm⁻²·s⁻¹ and 9 to 965 n·cm⁻²·s⁻¹, respectively. The minimum detectable limit for stray neutrons was included herein to demonstrate the reliability. Accordingly, 60 and two points, respectively, the confidence level associated with the reported intensities of fast and thermal neutrons reached 95%. The low qualified ratio in the evaluation of stray thermal neutrons might have been caused by either the high Compton scattering plateau or the low intensity of the gamma-ray peak in the relevant spectrum.     

Cyclic neutron activation analysis for determination of selenium in food samples using <Superscript>77m</Superscript>Se

Cyclic neutron activation analysis method was conducted for determination of Se in food samples. High accuracy and good precision were proved by analyzing certified reference materials (CRMs) of chicken (GBW10018), rice (GBW10010) and cabbage (GBW10014). The detection limits for the three CRMs reached 0.16, 0.66 and 1.2 ng after 6 cycles at the 161.9 keV γ-peak from ^77mSe, under a neutron flux of 9.0 × 10¹¹ n cm⁻² s⁻¹ and the conditions of 30 s irradiation, 2 s decay, 30 s counting and 2 s waiting, significantly lower than those of conventional neutron activation analysis without any cycles, which were 0.94, 3.6 and 4.3 ng, respectively.     

Determination of boron in iron-based metallic glasses by neutron irradiation and alpha-particle track recording in cellulose nitrate

A method for determination of the boron concentration in Fe−Ni−B metallic glasses has been developed. A solution of the alloy is irradiated with neutrons from a 10 μg²⁵²Cf neutron source and the alpha-activity due to the reaction¹⁰B(n, α)⁷Li is recorded by a cellulose nitrate alpha-track detector immersed in the solution. After the exposure the alpha-tracks are made visible in a microscope by etching the cellulose nitrate film in hot NaOH and the track density is determined.     

Detection limits of Au using fast and thermal neutron activation analyses

The applicability of fast and thermal neutron activation analyses for the determination of gold in rock samples has been studied. Using a Ge/Li/ detector limit of 0.45 mg g⁻¹ was obtained for a fast neutron flux of 8.10⁷ n cm⁻².s⁻¹. With a thermal neutron flux of 6.10⁵ n cm⁻².s⁻¹ and the same detector a value of 35 μg g⁻¹ was obtained. Using a NaI/Tl/ crystal a sensitivity of 14 μg g⁻¹ was attained at the same thermal flux. This work was supported in part by the Hungarian Academy of Sciences.     

Levels of <Superscript>232</Superscript>Th activity in the South Adriatic Sea marine environment of Montenegro

²³²Th activities in the South Adriatic Sea-water, surface sediment, mud with detritus, seagrass (Posidonia oceanica) samples, and the mullet (Mugilidae) species Mugil cephalus, as well as soil and sand from the Montenegrin Coast, were measured using the six-crystal spectrometer PRIPYAT-2M, which has relatively high detection efficiency and a good sensitivity, and allows a short acquisition time, and measuring samples of any shape, without preliminary preparation and calibration measurements for different sample geometries. An average ²³²Th activity concentration in surface soil layer is found to be 40.33 Bq kg⁻¹, while in sand—4.7 Bq kg⁻¹. The absorbed dose rate in air due to ²³²Th gamma radiation from surface soil layer ranged from 11.76 to 63.39 nGy h⁻¹, with a mean of 24.06 nGy h⁻¹. Corresponding average annual effective dose rate has been found to be 0.03 mSv y⁻¹. The absorbed dose rates due to the thorium gamma radiation in air at 1 m above sand surface on the Montenegrin beaches have been found to be from 0.41 to 9.08 nGy h⁻¹, while annual effective dose rates ranged from 0.0005 to 0.011 mSv y⁻¹. ²³²Th activity concentration in seawater ranged from 0.06 to 0.22 Bq L⁻¹, as in the mullet (Mugil cephalus) whole individuals from 0.63 to 1.67 Bq kg⁻¹. Annual intake of ²³²Th by human consumers of this fish species has been estimated to provide an effective dose of about 0.003 mSv y⁻¹.     

Use of granulated modified zeolite Y for the removal of inorganic arsenic and selenium species

Column studies were performed to evaluate the performance of modified zeolite-Y with ion Fe (zeolite-FeY) in removing As(III), As(V), Se(IV) and Se(VI) from groundwater. The removal capacities for zeolite-FeY was carried out on arsenic and selenium species in aqueous solution by co-precipitation technique with DBDTC-Pp complex in the pH range of 1.5–2.5 followed by the neutron activation analysis (NAA) using a TRIGA-MkII reactor with an average flux of 2.1 × 10¹² neutrons cm⁻² s⁻¹ and inductively coupled plasma-mass spectrometry (ICP-MS) technique as comparison. The accuracy between the results obtained from both techniques were compared and evaluated.     

Radiological assessment of natural and fallout radioactivity in the soil of Chamba and Dharamshala areas of Himachal Pradesh,India

This paper presents the results of measurement of natural and fallout radioactivity in soil samples of Chamba and Dharamshala areas in Himachal Pradesh, India. Spatial distribution of ²³⁸U, ²²⁶Ra, ²³²Th, ⁴⁰K, ¹³⁷Cs was determined using High resolution gamma-ray spectrometry. The mean activity concentration in Chamba region due to ²³⁸U, ²³²Th, ⁴⁰K and ¹³⁷Cs was 32.3, 58.4, 588.3, and 10.9 Bq kg⁻¹, respectively, whereas in Dharamshala it was 35.7, 61.3, 594.9, 10.0 Bq kg⁻¹, respectively. Absorbed gamma dose rate (D) in air was calculated using appropriate dose conversion factors, which was varying from 45 to 103 nGy h⁻¹. To control the radiation exposure due to natural radioactivity in soil, if it is used as building materials, radium equivalent activity (Ra_eq) and activity index were also evaluated. Radium equivalent activity calculated for the soil ranged from 95.5 to 234.2 Bq kg⁻¹ with average of 171.0 Bq kg⁻¹.The calculated Activity concentration index was ranged from 0.34 to 0.85 with an average value of 0.64. The natural and fallout radioactivity in soil of this region is comparable with Indian average and other parts of the world. The percentage contribution of ²³⁸U, ²³²Th and ⁴⁰K and ¹³⁷Cs to the average external gamma dose rate was 22, 46, 32, 2%, respectively. This shows that the dose contribution due to fallout radioactivity is negligible as compared to the natural radioactivity.     

Rare-earth elemental analysis of banded iron-formations by instrumental neutron activation analysis

Representative banded iron-formations (BIFs) from various locations of the eastern Indian geological belt were investigated by instrumental neutron activation analysis (INAA). After pre-concentration, irradiation was carried out using a neutron flux of 5.1·10¹⁶ m⁻²·s⁻¹, 1.0·10¹⁵ m⁻²·s⁻¹ and 3.7·10¹⁵ m⁻²s⁻¹, with thermal, epi-thermal and fast neutrons, respectively. The activities in these samples were measured by a HPGe detector. Ten rare-earth elements, such as La, Ce, Nd, Sm, Eu, Tb, Ho, Tm, Yb and Lu, have been qualitatively identified and quantitatively estimated in these samples. The present investigation is an example of employing a pre-concentration method for high iron-containing ores prior to neutron activation analysis.     

Determination of iodine in biological materials by pseudo-cyclic epithermal INAA using anti-coincidence gamma-ray spectrometry and estimation of expanded uncertainties

Epithermal instrumental neutron activation analysis (EINAA) technique in conjunction with anti-coincidence gamma-ray spectrometry (AC) has been applied for the determination of ppm to ppb levels of iodine in biological materials containing high levels of Al, Br, Cl, K, Mn, and Na. Both conventional EINAA-AC and pseudo-cyclic EINAA-AC (PC-EINAA-AC) methods using a combination of Cd and B filters have been developed using Dalhousie University SLOWPOKE-2 reactor (DUSR) facility. The expanded uncertainties (EU), at about 95% confidence level, for iodine in biological materials by EINAA-AC varied between 6 and 10%. The advantages of the non-destructive PC-EINAA-AC method has been successfully demonstrated by analyzing the NIST Pine Needles (SRM 1575) containing a low amount of iodine in presence of high quantities of Mn and other interfering elements where an iodine content of 92.8 μg kg⁻¹ with an EU of 6.1 μg kg⁻¹ and a detection limit of 40 μg kg⁻¹ has been obtained at the end of fourth cycle.