RADIATION SHIELDING AND DOSE ASSESSMENT OF THE CYCLOTRON FACILITY AT INRNE-BAS

The Institute for Nuclear Research and Nuclear Energy at the Bulgarian Academy of Sciences is working on the construction of a cyclotron centre. The facility is on a design level. At this stage of the project, an important task is the radiation shielding assessment of the facility. Nowadays, the Monte Carlo transport codes have become the tool of choice for solving this type of problem. In the current paper, the transport code FLUKA is used for the calculations. It is widely applied for shielding design and analysis of accelerators and their components. The distributions of the radiation fields inside and outside the cyclotron bunker are presented in this paper. Both different irradiation scenarios and bunker configurations are considered in the conducted Monte Carlo simulations. These results will be used as a guidance in site planning.

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Primary Shielding Design and Analysis for CLEAR-I

Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Radiation Protection and Nuclear Technology Applications ◽

10.1115/icone21-15995 ◽

2013 ◽

Author(s):

Qi Yang ◽

Bin Li ◽

Chao Chen ◽

Minghuang Wang ◽

Qin Zeng

Keyword(s):

Monte Carlo ◽

Monte Carlo Method ◽

Research Reactor ◽

Natural Circulation ◽

Radiation Shielding ◽

Lead Alloy ◽

Reactor Plant ◽

Calculation System ◽

Shielding Design ◽

Dual Models

The China Lead-Alloy Cooled Research Reactor (CLEAR-I) is critical/sub-critical dual-models natural circulation lead alloy cooled reactor. This study is to focus on the concern radiation shielding design and analysis for CLEAR-I. The modeling program MCAM and calculation system VisualBUS developed by FDS Team was used based on Monte Carlo method and other coupled methods. As indicated by the results, the dose rate in the reactor plant (outside the containing compartment above the reactor) was below 9 uSv/h during operation and less than 1 uSv/h during shutdown, meeting with the requirements of shielding.

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Monte Carlo Studies for the Radiation Shielding Design of LCLS-II

10.2172/1060215 ◽

2013 ◽

Author(s):

Shanjie Xiao ◽

Stan X. Mao ◽

Ludovic Nicolas ◽

Sayed H. Rokni ◽

Mario Santana Leitner

Keyword(s):

Monte Carlo ◽

Radiation Shielding ◽

Monte Carlo Studies ◽

Shielding Design

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Experimental investigation on radiation shielding of high performance concrete for nuclear and radiotherapy facilities

Polish Journal of Medical Physics and Engineering ◽

10.1515/pjmpe-2016-0008 ◽

2016 ◽

Vol 22 (2) ◽

pp. 41-47 ◽

Cited By ~ 5

Author(s):

Szymon Domański ◽

Michał A. Gryziński ◽

Maciej Maciak ◽

Łukasz Murawski ◽

Piotr Tulik ◽

...

Keyword(s):

High Performance ◽

High Performance Concrete ◽

Absorbed Dose ◽

Nuclear Research ◽

Neutron Radiation ◽

Radiation Shielding ◽

Transmitted Radiation ◽

Radiation Fields ◽

Shielding Properties ◽

Set Up

Abstract This paper presents the set of procedures developed in Radiation Protection Measurements Laboratory at National Centre for Nuclear Research for evaluation of shielding properties of high performance concrete. The purpose of such procedure is to characterize the material behaviour against gamma and neutron radiation. The range of the densities of the concrete specimens was from 2300 to 3900 kg/m3. The shielding properties against photons were evaluated using 137Cs and 60Co sources. The neutron radiation measurements have been performed by measuring the transmitted radiation from 239PuBe source. Scattered neutron radiation has been evaluated using the shadow cone technique. A set up of ionization chambers was used during all experiments. The gamma dose was measured using C-CO2 ionization chamber. The neutron dose was evaluated with recombination chamber of REM-2 type with appropriate recombination method applied. The method to distinguish gamma and neutron absorbed dose components in mixed radiation fields using twin detector method was presented. Also, recombination microdosimetric method was applied for the obtained results. Procedures to establish consecutive half value layers and tenth value layers (HVL and TVL) for gamma and neutron radiation were presented. Measured HVL and TVL values were linked with concrete density to highlight well known dependence. Also, influence of specific admixtures to concrete on neutron attenuation properties was studied. The results confirmed the feasibility of approach for the radiation shielding investigations.

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Yalina-thermal facility neutron characteristic computational study 129I, 237Np and 243Am transmutation reaction rates calculations

EPJ Web of Conferences ◽

10.1051/epjconf/202023922013 ◽

2020 ◽

Vol 239 ◽

pp. 22013

Author(s):

Tamara Korbut ◽

Maksim Kravchenko ◽

Ivan Edchik ◽

Sergey Korneev

Keyword(s):

Monte Carlo ◽

Computational Study ◽

Nuclear Research ◽

Nuclear Data ◽

Reaction Rates ◽

Neutron Field ◽

Minor Actinide ◽

National Academy Of Sciences ◽

Data Libraries ◽

Academy Of Sciences

Present work describes Monte-Carlo calculations of the neutron field and minor actinide transmutation reaction rates within the Yalina-Thermal sub-critical assembly of the Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus. The computer model of the facility was prepared for the corresponding calculations via MCU-PD and MCNP Monte-Carlo codes. The model neutron characteristics estimations were performed as well as the nuclear safety analysis. The up-to-date ENDF B/VIII, JEFF 3.3 and JENDL 4.0 nuclear data libraries were used during research.

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Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

EPJ Web of Conferences ◽

10.1051/epjconf/201817001019 ◽

2018 ◽

Vol 170 ◽

pp. 01019

Author(s):

Jérôme M. Verbeke ◽

Odile Petit ◽

Abdelhazize Chebboubi ◽

Olivier Litaize

Keyword(s):

Monte Carlo ◽

Reactor Core ◽

Radiation Shielding ◽

Model Parameters ◽

Monte Carlo Code ◽

Coincidence Counting ◽

Angular Correlations ◽

Fission Neutrons ◽

Nuclear Instrumentation ◽

Monte Carlo Transport

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using either FREYA or FIFRELIN, are compared to experimental results. For 240Pu(sf), the measured correlations were used to tune the model parameters.

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Radiation shielding design for neutron diffractometers assisted by Monte Carlo methods

Physica B Condensed Matter ◽

10.1016/j.physb.2006.06.064 ◽

2006 ◽

Vol 385-386 ◽

pp. 1321-1323 ◽

Cited By ~ 8

Author(s):

John C. Osborn ◽

Tunay Ersez ◽

George Braoudakis

Keyword(s):

Monte Carlo ◽

Monte Carlo Methods ◽

Radiation Shielding ◽

Shielding Design

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Neutron dose evaluation of Elekta Linac at two energies (10 & 18 MV) by MCNP code and comparison with experimental measurements

JOURNAL OF ADVANCES IN PHYSICS ◽

10.24297/jap.v6i1.1820 ◽

2014 ◽

Vol 6 (1) ◽

pp. 1006-1015

Author(s):

Negin Shagholi ◽

Hassan Ali ◽

Mahdi Sadeghi ◽

Arjang Shahvar ◽

Hoda Darestani ◽

...

Keyword(s):

Monte Carlo ◽

Measurement Data ◽

Calculated Data ◽

High Energy ◽

Neutron Dose ◽

Dose Assessment ◽

Photon Flux ◽

Dose Equivalent ◽

Monte Carlo Techniques ◽

Neutron Dose Equivalent

Medical linear accelerators, besides the clinically high energy electron and photon beams, produce other secondary particles such as neutrons which escalate the delivered dose. In this study the neutron dose at 10 and 18MV Elekta linac was obtained by using TLD600 and TLD700 as well as Monte Carlo simulation. For neutron dose assessment in 2020 cm2 field, TLDs were calibrated at first. Gamma calibration was performed with 10 and 18 MV linac and neutron calibration was done with 241Am-Be neutron source. For simulation, MCNPX code was used then calculated neutron dose equivalent was compared with measurement data. Neutron dose equivalent at 18 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 3.3, 4, 5 and 6 cm. Neutron dose at depths of less than 3.3cm was zero and maximized at the depth of 4 cm (44.39 mSvGy-1), whereas calculation resultedÂ in the maximum of 2.32 mSvGy-1 at the same depth. Neutron dose at 10 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 2.5, 3.3, 4 and 5 cm. No photoneutron dose was observed at depths of less than 3.3cm and the maximum was at 4cm equal to 5.44mSvGy-1, however, the calculated data showed the maximum of 0.077mSvGy-1 at the same depth. The comparison between measured photo neutron dose and calculated data along the beam axis in different depths, shows that the measurement data were much more than the calculated data, so it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry in linac central axis due to high photon flux, whereas MCNPX Monte Carlo techniques still remain a valuable tool for photonuclear dose studies.

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