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.

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.

NEUTRON SPECTRUM UNFOLDING USING VARIOUS COMPUTER CODES: A COMPARISON

2021 ◽  
pp. 128-133
Author(s):  
Thiệm
Keyword(s):  
Neutron Spectrum ◽  
Neutron Fluence ◽  
Dose Equivalent ◽  
Spectrum Unfolding ◽  
Bonner Sphere ◽  
Ambient Dose Equivalent ◽  
Computer Codes ◽  
Neutron Spectrum Unfolding ◽  
Spectrometer System ◽  
Ambient Dose

Dosimetric quantities at various distances from a 30 cm diameter polyethylene sphere moderated 241Am-Be source were investigated using the Bonner sphere spectrometer system. The different international commercial unfolding codes were applied to unfold the neutron spectrum, and their shapes were compared to each other. Additionally, the integrated neutron fluence rates overall spectrum and fractional neutron fluence rates were deduced and compared between the results obtained from different unfolding codes. As an important quantity applying in radiation safety assessment, the neutron ambient dose equivalent rates were also calculated and compared to each other to verify the utility feasibilities of the codes.

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.

scholarly journals Analysis of neutron streaming through penetrations in the JET biological shielding

2019 ◽  
Vol 22 ◽  
pp. 34
Author(s):  
I. E. Stamatelatos ◽  
T. Vasilopoulou ◽  
P. Batistoni ◽  
S. Conroy ◽  
B. Obryk ◽  
...  
Keyword(s):  
Design Methodology ◽  
Neutron Fluence ◽  
Radiation Shielding ◽  
Mcnp Code ◽  
Dose Equivalent ◽  
Toroidal Plasma ◽  
Monte Carlo Calculations ◽  
Ambient Dose Equivalent ◽  
Shielding Design ◽  
Ambient Dose

In the present work neutron streaming through large ducts and labyrinths of the Joint European Torus (JET) biological shielding was evaluated. Neutron fluence and ambient dose equivalent were calculated along the total length of the ducts. Monte Carlo calculations using the MCNP code were performed for both Deuterium-Deuterium (D-D) and Deuterium-Tritium (D-T) toroidal plasma discharge sources. The results of the calculations were compared against measurements performed using thermoluminescence detectors. This work contributes to the operational radiation protection effort to minimize collective radiation dose to personnel at JET and, moreover, provides important information from JET experience that may assist in the optimization and validation of the radiation shielding design methodology used in future fusion plants, such as ITER and DEMO.

scholarly journals Development of a neutron detector for radiation protection monitoring

2021 ◽  
Vol 10 (4) ◽  
pp. 41-47
Author(s):  
Mai Van Dien ◽  
Nguyen Duc Tuan ◽  
Nguyen Ngoc Quynh ◽  
Vu Trung Tan ◽  
Le Ngoc Thiem ◽  
...  
Keyword(s):  
Radiation Protection ◽  
High Density Polyethylene ◽  
Neutron Detector ◽  
Neutron Fluence ◽  
Dose Equivalent ◽  
Nuclear Science ◽  
Ambient Dose Equivalent ◽  
Mcnp5 Code ◽  
Ambient Dose ◽  
Good Agreement

The paper presents the results of the development of a neutron detector for radiation protection purposes. Monte Carlo simulations, using MCNP5 code, were performed to optimize the configuration of the neutron detector. The developed detector consists of a 3He proportional counter embedded in a multi-layer moderator made of high-density polyethylene (HDPE) and Cadmium. The characteristics of the developed neutron detector including neutron fluence response and ambient dose equivalent response were calculated, analyzed and compared with those from other neutron survey meters. The simulation model and computed results were assessed through experimental measurements at the Secondary Standards Dosimetry Laboratory of the Institute for Nuclear Science and Technology (INST). A good agreement between the simulated and experimental results was observed within 9.3% for 241Am-Be source and four simulated workplace neutron fields.

scholarly journals Neutron streaming calculations in the JET entrance labyrinth

2019 ◽  
Vol 21 ◽  
pp. 133
Author(s):  
I. E. Stamatelatos ◽  
T. Vasilopoulou ◽  
P. Obryk ◽  
P. Bilski ◽  
S. Conroy ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Simulations ◽  
Radiation Protection ◽  
Safety Assessment ◽  
Neutron Fluence ◽  
Mcnp Code ◽  
Dose Equivalent ◽  
Ambient Dose Equivalent ◽  
Ambient Dose

Neutron streaming along the entrance labyrinth of the Joint European Torus (JET) was evaluated. Monte Carlo simulations using MCNP code were performed to calculate neutron fluence and ambient dose equivalent along the length of the labyrinth. The results of this work aim to assist operational radiation protection activities in the JET facility and contribute to the validation of the safety assessment calculations made for ITER.

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