Effect of electrode separation and electrode backscatter thickness of a parallel plate ionization chamber on the measurement of dose to tissue in beta radiation fields.

Plane parallel plate ionization chamber (PPC) is used in dosimetry especially for beta particles and low energy electron beam. The response of the PPC is affected by the electrode separation, thickness and material of backscatterer. The effect of electrode separation arises due to the well-known inscattering effect of electron which causes fluence perturbation inside the chamber cavity. The perturbation is caused due to reduced electron scattering in the chamber cavity compared to an ideal phantom material and it depends on the thickness of the cavity. Furthermore, the response of PPC also gets affected by the material behind the air cavity. Variation in the response depends on the thickness and the atomic number of the material behind collecting electrode of the PPC. The theoretical studies on the effect of collecting electrode backscatter thickness and electrode separation on the response of a PPC used as a transfer standard in ISO 6980 reference beta radiation field from 85Kr, 90Sr-90Y and 106Ru-106Rh radionuclides are presented. Multi-particle transport code FLUKA is used for the studies. The Polymethyl Methacrylate (PMMA) plate having 3.5 mm thickness is found to provide full backscatter for the above beta radionuclides. It is also observed that even for a well-guarded PPC, as the chamber electrode separation increases, the measured depth dose curve deviates from the ideal depth dose curve and the effective point of measurement (EPOM) of the PPC shifts towards downward direction from chamber reference point. It is also observed that the deviation between ideal and measured depth dose curve (related to EPOM shift) depends on the cavity thickness of the PPC. In the present work, optimization of design parameters of a PPC is carried out to establish it as a transfer standard in compliance with ISO 6980 for the standardization of reference beta radiation fields from 85Kr, 90Sr-90Y and 106Ru-106Rh radionuclide.

Activity measurement of mixed complex radionuclide like 152Eu with different methods.

152Eu has been standardized by three independent 4π ß-γ coincidence counting systems with beta detectors as proportional counter, plastic scintillator and liquid scintillator along with the CIEMAT/NIST method. The average activity concentration by primary methods was linked to key comparison reference value (KCRV) by comparing it with that of 4π γ ionization chamber (GIC) whose calibration factor was determined from the KCRV (BIPM.RI(II)-K1.Eu-152 and CCRI(II)-K2.Eu-152) and deviates from GIC by ± 0.16% indicating good agreement within standard uncertainties.

Evaluation of influential correction factors applicable for absolute beta dosimetry of 85Kr, 90Sr-90Y and 106Ru-106Rh beta emitters.

Extrapolation ionization chamber is normally used for absolute measurement of absorbed dose to tissue and in standardization of beta particle emitting sources in particular. However, in this technique, several correction factors are applied to the measured ionization current to evaluate the true value of the dose to the tissue from beta radiation. These correction factors are described in ISO 6980-2:2004. Out of these, ion recombination correction factor (ksat) and ambient air density correction factor (kad) are evaluated experimentally during measurement of ionization current. Other significant correction factors are backscatter correction factor (kba), correction factor for perturbation in electron spectrum due to ionization chamber side wall scatter (kpe), correction factor for difference in attenuation of beta spectrum due to change in ambience condition (kabs) and correction factor for radial non-uniformity of dose inside detector volume (kra). Experimental evaluation of correction factors kba, kpe and kabs are not easily achievable. In the present work four significant correction factors kba, kpe, kabs and kra are derived theoretically by Monte Carlo particle transport simulation. These correction factors are evaluated for beta emission spectrum of 85Kr, 90Sr-90Y and 106Ru-106Rh radioactive sources.

Standardisation of 133Ba by efficiency extrapolation method and calibration of ionisation chamber.

133Ba has been standardised by direct measurements for the first time in the laboratory using two counting systems: (i) the 4πß (plastic scintillator) -γ coincidence, (ii) the 4πß (proportional counter) -γ coincidence. Furthermore, this standardisation experiment demonstrates the performance and applicability of the recently developed 4πß (plastic scintillator)-γ coincidence system for radionuclides decaying with complex decay schemes as well as for e, X-γ emitters. Additionally, 133Ba solution standards were prepared to calibrate the pressurized 4π γ ionisation chamber and determination of the calibration coefficient. The En score is a statistical indicator of the agreement between two independent estimations. Thus, the performance of the PS system was compared to the result obtained with the PC system using the En score as specified in the ISO13528:2015. The results of measurements are acceptable if En â‰¦ 1.0. An En score of 0.2 was obtained which indicates that, the 133Ba activity concentration obtained by the 4πß (plastic scintillator) -γ coincidence and 4πß (proportional counter) -γ coincidence systems are in agreement. This paper presents the standardisation procedure, the results obtained by the measurements and their comparison.

ABSOLUTE MEASUREMENT OF 14.57 MeV NEUTRON FLUENCE RATE USING PROTON RECOIL NEUTRON TELESCOPE.

A single stage vacuum-type proton recoil neutron telescope (PRT) was used for accurate measurement of 14.57 MeV neutron fluence rate from an indigenously developed D-T neutron generator at Purnima, BARC. The telescope consists of a polyethylene radiator having 4 cm diameter and CsI (Tl) scintillation crystal having thickness 1.5 mm and 4 cm diameter separated by 20.5 cm kept in a vacuum chamber. The neutron detection efficiency of the telescope for 14.57 MeV neutrons was calculated analytically using n-p scattering cross section data from Evaluated Nuclear Data File VII and also evaluated using fluka simulation. The relativistic transformation of n-p differential scattering cross section from centre-of-mass to laboratory system was used for calculating the efficiency of PRT. The 14.57 MeV neutron fluence rate was also measured using copper foils. The comparison of fluence rate measured using PRT and copper foil activation techniques is presented in this paper. The total uncertainty in measurement using PRT and copper foil activation technique is found to be 3.93 and 6.97%, respectively.

Development of plastic scintillator based primary standard for activity measurements and its performance evaluation.

4πß-γ coincidence technique is a powerful tool and widely recognised method to determine the absolute activity concentration of radioactive solutions. A new plastic scintillator based coincidence system has been developed and established as a primary standard for radioactivity measurements. The performance of the system was evaluated by the standardisation of 60Co radioactive solution due to its simple decay scheme. The activity concentration results obtained by the new system were compared with the existing proportional counter and liquid scintillation based 4πß-γ coincidence systems. This paper discusses the design details of the new system and its performance evaluation.

Neutron radiation shielding properties of polymer incorporated self compacting concrete mixes.

In this work, the neutron radiation shielding characteristics of a class of novel polymer-incorporated self-compacting concrete (PISCC) mixes are evaluated. Pulverized high density polyethylene (HDPE) material was used, at three different reference volumes, as a partial replacement to river sand in conventional concrete mixes. By such partial replacement of sand with polymer, additional hydrogen contents are incorporated in these concrete mixes and their effect on the neutron radiation shielding properties are studied. It has been observed from the initial set of experiments that there is a definite trend of reductions in the neutron flux and dose transmission factor values in these PISCC mixes vis-à-vis ordinary concrete mix. Also, the fact that quite similar enhanced shielding results are recorded even when reprocessed HDPE material is used in lieu of the virgin HDPE attracts further attention.

Study of neutron spectra emitted by moderated 241Am-Be source for various moderating materials by using simulation technique.

A study was done on the tailored neutron energy spectra of (241)Am-Be neutron source due to the effect of moderators. The (241)Am-Be laboratory neutron source was used as the basic source and the emitted spectrum was modified using various neutron moderators. The various moderators used are high-density polythene, light water, heavy water, graphite, (56)Fe, BeO, Be, (6)Li and (7)Li. The absolute energy spectra and fluences in each case are calculated by using the Monte Carlo code FLUKA. This paper describes the simulation work done to design a moderated (241)Am-Be neutron source to produce various energy neutron spectra.

Standardisation of water-moderated 241Am-Be neutron source using De Pangher neutron long counter: experimental and Monte Carlo modelling.

A convenient neutron source is made for calibration of neutron survey instruments and personal dosimeters that are used in various nuclear installations such as fuel reprocessing, waste management, fuel fabrication and oil and well logging facilities, etc. This source consists of a bare (241)Am-Be neutron source placed at the centre of a 15-cm radius stainless steel spherical shell filled with distilled water. This paper describes the standardisation of the source at Bhabha Atomic Research Centre, using De Pangher neutron long counter both experimentally and using the Monte Carlo simulation. The ratio of neutron yield of water moderated to the bare (241)Am-Be neutron source was found to be 0.573. From the simulation, the neutron-fluence-weighted average energy of water-moderated (241)Am-Be source (fluence-weighted average energy of 2.25 MeV, dose-weighted average energy of 3.55 MeV) was found to be nearly the same as that of a (252)Cf source (fluence-weighted average energy of 2.1 MeV, dose-weighted average energy of 2.3 MeV). This source can be used for calibration in addition to (252)Cf, to study the variation in response of neutron monitoring instruments.

Study on the response of indigenously developed CaSO4:Dy phosphor-based neutron dosemeter.

In the present paper, we report the indigenous development of a neutron-sensitive thermoluminescent dosemeter (TLD) based on CaSO4:Dy Teflon TL disc. The study includes indigenous development of a neutron dosemeter and its response in terms of operational quantity to different energies of neutrons under various irradiation conditions. It was found that the thermal neutron sensitivity of the CaSO4:Dy Teflon neutron disc is about one-third of TLD-600. However, the thermal neutron sensitivity with respect to CaSO4:Dy Teflon gamma disc is about 42 times for in-air irradiation and about 84 times for on-phantom irradiation condition. This newly developed neutron dosemeter can be used as a routine TLD with a slight change in the design of the TLD badge system in the mixed fields of gamma and neutrons of energy up to 500 keV for radiation workers engaged in nuclear fuel cycle operation and other applications.