The Design of an Ambient Neutron Dose-Equivalent Meter

2011 ◽  
Vol 133 (8) ◽  
Author(s):  
Xia Wenming ◽  
Jia Mingchun ◽  
Guo Zhirong
Keyword(s):  
Neutron Radiation ◽  
Monte Carlo Code ◽  
Energy Response ◽  
Dose Equivalent ◽  
Neutron Dosimeter ◽  
Radiation Fields ◽  
Ambient Dose Equivalent ◽  
Ambient Dose ◽  
Detector Design ◽  
Single Counter

At present, most of the developed neutron dosimeters used to measure the neutron ambient dose-equivalent that has a moderator with a single counter, applied in neutron radiation fields within large range energies from thermal to MeV neutrons, are not a satisfaction to energy response. The purpose of this article is to design a suitable neutron dosimeter for radiation protection purpose. In order to overcome the disadvantage of the energy response of the neutron dosimeters combining a single sphere with a single counter, three spheres and three H3e counters were combined for the detector design. The response function of moderators with different thicknesses combined with SP9 H3e counters were calculated with Monte Carlo code MCNP 4C. The selection of three different thicknesses of the moderating polyethylene sphere was done with a MATLAB program. A suitable combination of three different thicknesses was confirmed for the detector design. The electronic system of the neutron dosimeter was introduced. The results of ambient dose-equivalent per unit fluence in different radiation areas were calculated, analyzed, and compared with the values recommended in the ISO standard. The calculated result explains that it is very significant to this design of neutron dosimeter; it may be applied to the monitor of the ambient dose in the neutron radiation fields, improving at present the status of the energy response of neutron dosimeters.

The Design of a Neutron Dosimeter

2010 ◽  
Author(s):  
Wenming Xia ◽  
Mingchun Jia ◽  
Zhirong Guo
Keyword(s):  
Neutron Radiation ◽  
Electronic System ◽  
Energy Response ◽  
Neutron Dosimeter ◽  
Radiation Fields ◽  
The Status ◽  
Ambient Dose Equivalent ◽  
Ambient Dose ◽  
Detector Design ◽  
Single Counter

At present, most of the developed neutron dosimeters that have a moderator with a single counter, applied in neutron radiation fields within large range energies from thermal to MeV neutrons, are not a satisfaction to energy response. The purpose of the article is designing a suitable neutron dosimeter for the radiation protection purpose. In order to overcome the disadvantage of the energy response of the neutron dosimeters combined a single sphere with a single counter, three spheres and three 3He counters were combined for the detector design. The response function of moderators with different thicknesses combined with SP9 3He counters were calculated with MCNP program MCNP4C [1]. The selection of three different thicknesses of the moderating polyethylene sphere was done with a Matlab program [2]. A suitable combination of three different thicknesses was confirmed for the detector design. The electronic system of the neutron dosimeter was introduced. The fluence to ambient dose-equivalent conversion coefficient were calculated, analyzed and compared with the values recommended in the ICRP 74 Publication [3]. The calculated result explain that it is very significance to this design of neutron dosimeter, it may be applied to the monitor of the ambient dose in the neutron radiation fields, improving at present the status of the energy response of neutron dosimeters.

DEVELOPMENT OF A NEW SECONDARY STANDARD FOR H*(10)

2020 ◽  
Vol 188 (3) ◽  
pp. 378-382
Author(s):  
K Bairlein ◽  
B Behnke ◽  
O Hupe
Keyword(s):  
Ionization Chamber ◽  
Secondary Standard ◽  
Dose Equivalent ◽  
Air Kerma ◽  
Radiation Fields ◽  
Ambient Dose Equivalent ◽  
Conversion Coefficients ◽  
Aluminium Coating ◽  
Pulsed Fields ◽  
Ambient Dose

Abstract A secondary standard for ambient dose equivalent, H*(10), is necessary for the dissemination of the unit Sievert (Sv), but there is no such standard commercially available currently. Furthermore, the measurement of H*(10) instead of calculating H*(10) from air kerma and conversion coefficients is needed for unknown radiation fields. We developed a prototype of a new secondary standard for H*(10) based on a spherical 1 l ionization chamber for air kerma. This chamber was modified with copper wires at the inner surface to adjust the response of the chamber according to H*(10). Additionally, a Makrolon shell and an aluminium coating were added to optimize the response at energies below 50 keV. The prototype fulfils the requirements given in ISO 4037-2 in the energy range from 12 keV to 7 MeV. In combination with an electrometer, it can be used as area dosemeter, suitable for pulsed fields and for low energy radiation.

scholarly journals Development of Neutron Dosimeter Using CR-39 for Measurement of Ambient Dose Equivalent

2010 ◽  
Vol 45 (4) ◽  
pp. 306a-306a
Author(s):  
Daisuke MAKI ◽  
Wakako SHINOZAKI ◽  
Hiroyuki OHGUCHI ◽  
Takashi NAKAMURA ◽  
Takayoshi YAMAMOTO
Keyword(s):  
Dose Equivalent ◽  
Neutron Dosimeter ◽  
Ambient Dose Equivalent ◽  
Ambient Dose

scholarly journals Neutron dosimetry at workplaces of JC “Institute of Nuclear Materials”

2021 ◽  
Vol 14 (2) ◽  
pp. 89-99
Author(s):  
M. D. Pyshkina ◽  
A. V. Vasilyev ◽  
A. A. Ekidin ◽  
E. I. Nazarov ◽  
M. A. Romanova ◽  
...  
Keyword(s):  
Angular Distribution ◽  
Flux Density ◽  
Radiation Flux ◽  
Neutron Radiation ◽  
Correction Factors ◽  
Individual Dose ◽  
Dose Equivalent ◽  
Ambient Dose Equivalent ◽  
True Values ◽  
Ambient Dose

If the neutron fields at personnel workplaces differ from the neutron fields in which individual dosimeters are verified, there is a possibility of additional errors in the assessment of such dosimetric quantities as ambient dose equivalent, individual dose equivalent or effective dose. To take into account the energy distribution of the neutron radiation flux density and the geometry of the irradiation of workers, it is necessary to study the characteristics of the fields of neutron radiation at the workplaces of the personnel. In order to obtain conditionally true levels of personnel exposure to neutron radiation at nuclear facilities, studies of the energy and angular distribution of the neutron radiation flux density were carried out at the workplaces of the Institute of Reactor Materials JSC, Zarechny. The energy distribution of the neutron radiation flux density was obtained using an MKS-AT1117M multi-sphere dosimeter-radiometer with a BDKN-06 detection unit and a set of polyethylene spheres-moderators. The angular distribution of the neutron radiation flux density was estimated from the results of measurements of the accumulated dose of neutron radiation by individual thermoluminescent dosimeters placed on four vertical planes of a heterogeneous human phantom. The results of measurements of the energy and angular distribution of the neutron radiation flux density made it possible to estimate the conditionally true values of the ambient and individual dose equivalents. The calculated conventionally true values differ from the measured values from 0.7 to 8.9 times for the ambient dose equivalent and from 6 to 50 times for the individual dose equivalent. In order to reduce the error in assessing the effective dose of personnel using personal dosimeters, correction factors were determined. For different workplaces and types of personal dosimeters, correction factors are in the range of values from 0.02 to 0.16.

scholarly journals Dose Rate Measurements in Pulsed Radiation Fields by Means of an Organic Scintillator

2021 ◽  
Vol 253 ◽  
pp. 09002
Author(s):  
Theresa Werner ◽  
Roland Beyer ◽  
Richard Biedermann ◽  
Marko Gerber ◽  
Jürgen Götze ◽  
...  
Keyword(s):  
Digital Data ◽  
Laser Machining ◽  
Dose Equivalent ◽  
Organic Scintillator ◽  
Radiation Fields ◽  
Medical Diagnostic ◽  
Ambient Dose Equivalent ◽  
Short Time ◽  
Ambient Dose

A deficiency in the implementation of current radiation protection is the determination of the ambient dose equivalent H*(10) and the directional dose equivalent H´(0.07) in pulsed radiation fields. Conventional dosimeter systems are not suitable for measurements in photon fields comprising short radiation pulses, which consequently leads to high detector loads in short time periods. Nevertheless, due to the implementation of advanced medical accelerators for cancer therapy, new medical diagnostic devices as well as various laser machining systems, there is an urgent need for suitable dosimeter systems for real time dosimetry. In this paper, a detector concept based on an organic scintillator and a full digital data analysis with the aim of developing a portable, battery powered measurement system is presented.

scholarly journals Ambient dose equivalent measurements in secondary radiation fields at proton therapy facility CCB IFJ PAN in Krakow using recombination chambers

Nukleonika ◽  
2016 ◽  
Vol 61 (1) ◽  
pp. 23-28 ◽  
Author(s):  
Edyta A. Jakubowska ◽  
Michał A. Gryziński ◽  
Natalia Golnik ◽  
Piotr Tulik ◽  
Liliana Stolarczyk ◽  
...  
Keyword(s):  
Nuclear Physics ◽  
Gamma Probe ◽  
Secondary Radiation ◽  
Dose Equivalent ◽  
Radiation Fields ◽  
Tissue Equivalent ◽  
Ambient Dose Equivalent ◽  
Dosimetric System ◽  
Ambient Dose

AbstractThis work presents recombination methods used for secondary radiation measurements at the Facility for Proton Radiotherapy of Eye Cancer at the Institute for Nuclear Physics, IFJ, in Krakow (Poland). The measurements ofH*(10) were performed, with REM-2 tissue equivalent chamber in two halls of cyclotrons AIC-144 and Proteus C-235 and in the corridors close to treatment rooms. The measurements were completed by determination of gamma radiation component, using a hydrogen-free recombination chamber. The results were compared with the measurements using rem meter types FHT 762 (WENDI-II) and NM2 FHT 192 gamma probe and with stationary dosimetric system.

scholarly journals Development of neutron survey meter using prescila neutron probe

2021 ◽  
Vol 8 (4) ◽  
pp. 26-33
Author(s):  
Hong Luong Thi ◽  
Phong Nguyen Tien ◽  
Bich Pham Thi ◽  
Huyen Nguyen Du
Keyword(s):  
High Energy ◽  
Energy Response ◽  
Dose Equivalent ◽  
Directional Response ◽  
Neutron Probe ◽  
Ambient Dose Equivalent ◽  
The Relationship ◽  
Non Destructive ◽  
Proper Operation ◽  
Ambient Dose

This paper presents the design and validation of a neutron survey meter. The meter consists of a PRESCILA neutron probe (with good sensitivity, directional response, gamma rejection, and enhanced high-energy response to 20 MeV) and an electrometer developed at Non-Destructive Evaluation center. The homogeneity response of the PRESCILA neutron probe was investigated as a function of distances from the 241Am - 9Be source in order to obtain the appropriate distance for accurate count-rate measurements using the neutron survey meter. A system consists of the PRESCILA neutron probe and the Ludlum Model 2326 electrometer was then used for measuring neutron ambient dose equivalent rates in the range from 50 cm to 200 cm with the step of 25 cm. The relationship between the count-rates and neutron dose equivalent rates (in the distance ranged from 50 to 200 cm) were deduced to validate the proper operation of the neutron survey meter.

Study on On-Site Calibration Method of Stationary Neutron Dosimeter

2020 ◽  
Author(s):  
Jinxu Lv ◽  
Ning Lv ◽  
Huiping Guo ◽  
Mingyan Sun ◽  
Kuo Zhao ◽  
...  
Keyword(s):  
Radiation Field ◽  
Neutron Fluence ◽  
Neutron Radiation ◽  
Calibration Method ◽  
Calibration Coefficient ◽  
Correction Coefficient ◽  
Dose Equivalent ◽  
Energy Correction ◽  
Ambient Dose Equivalent ◽  
Ambient Dose

Abstract The on-site calibration system of stationary neutron ambient dose equivalent instrument is mainly comprised of a small controllable neutron source and a reference neutron ambient dose equivalent instrument. According to the principle of “relative calibration method”, a small controllable neutron source continuously emits neutrons at the calibration site to construct a neutron radiation field. The calibration factor (NB) can be obtained by comparing the response numbers from two instrument, the instrument to be calibrated and the reference instrument, which are symmetrically placed in the neutron radiation field. In order to complete the transfer of the ambient dose equivalent calibration coefficient of the reference instrument from National Metrology Center (252CfStandard radiation field) to nuclear facility site, the calibration coefficient needs to be corrected, that is, multiplied by the “energy correction coefficient”. Energy correction coefficient includes: (1) “Instrument Energy Response Correction Coefficient” kε for neutron fluence to neutron counting of the instrument, (2) “Conversion Correction Coefficient” kΦ for neutron fluence to neutron ambient dose equivalent. Tests have shown that the ambient dose equivalent rate measurement error of the instrument which have been calibrated but without energy correction was 16.5%, by contrast the measurement error was 2.9% with energy correction. It can be seen that the energy correction is necessary and effective for the measuring instrument of neutron ambient dose equivalent during the on-site calibration process.

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