Radiation shielding design of BNCT treatment room for D-T neutron source

Pouryavi, Mehdi; Masoudi, S. Farhad; Rahmani, Faezeh

doi:10.1016/j.apradiso.2015.02.016

Cited by 12 publications

(7 citation statements)

References 12 publications

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“…All five of these materials then simulated using MCNPX to shield the radiation which emitted from the collimator in thermal column with thickness variation of 10 cm, starting from 10 cm until 100 cm as shown in Figure 3. (Santoso, 2014) (Pouryavi, 2015) …”

Section: Methodsmentioning

confidence: 99%

“…This research was chosed as a reference to choose the shielding material and to designt the geometry of the treatment room. (Pouryavi, 2015) A research about the properties of some alloy materials for neutron and photon shielding was investigated as in (Singh & Badiger, 2014) . The alloy materials are CS-516, SS-403, SS-410, SS-316, SS-316L, SS-304L, Incoloy-600, Monel-400 and Cupero-Nickel.…”

Section: Methodsmentioning

confidence: 99%

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Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator

Dwiputra¹,

Harto²,

Sardjono

et al. 2016

ijpna

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Abstract-Studies were carried out to design a shielding for BNCT facility in the end of Kartini reactor's thermal column with predesigned collimator. The design consist of selecting the material and their thickness. The shielding is required to absorb the leaking radiation until the Dose Limit Value of 20 mSv/year for radiation worker is met. The material considered were paraffin, barite concrete, borated polyethylene, stainless steel 304 and lead. The calculation was done using MCNPX tally facility with converted dose limit value of 10.42 µSv/hour. Design number two were chosen as the best from three designs which surrounded a room with length, width and height of, respectively 200 cm, 200 cm and 166.4 cm. The first and main layer are borated polyethyelene and barite concrete of 20 and 30 cm, respectively. The additional layer are borated polyethyelene and barite concrete of 15 cm and 15 cm with less volume than the main layer to decrease the primary straight radiation from the thermal column. Maximum radiation dose rate is 7.0746 µ Sv/ hour in cell 227 with average dose rate of 2.58712 µSv/hour.

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Section: Methodsmentioning

confidence: 99%

Section: Methodsmentioning

confidence: 99%

Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator

Dwiputra¹,

Harto²,

Sardjono

et al. 2016

ijpna

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“…Also, there are documents dealing with the design and optimization of the BNCT treatment rooms considering appropriate radiation shielding owing to the fact that protecting the facility personnel and the general public from radiation exposure is a primary safety concern in radiation therapies. The shields may be designed for choosing appropriate materials and their dimensions and geometries, the monitoring window for the patient position, the maze, and the entrance door of the room 21 , 24 – 27 .…”

Section: Introductionmentioning

confidence: 99%

Using ANN for thermal neutron shield designing for BNCT treatment room

Rasouli,

Yahyaee,

Masoudi

2024

Sci Rep

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Occupational radiation protection should be applied to the design of treatment rooms for various radiation therapy techniques, including BNCT, where escaping particles from the beam port of the beam shaping assembly (BSA) may reach the walls or penetrate through the entrance door. The focus of the present study is to design an alternative shielding material, other than the conventional material of lead, that can be considered as the material used in the door and be able to effectively absorb the BSA neutrons which have slowed down to the thermal energy range of $$< 1$$ < 1 eV after passing through the walls and the maze of the room. To this aim, a thermal neutron shield, composed of polymer composite and polyethylene, has been simulated using the Geant4 Monte Carlo code. The neutron flux and dose values were predicted using an artificial neural network (ANN), eliminating the need for time-consuming Monte Carlo simulations in all possible suggestions. Additionally, this technique enables simultaneous optimization of the parameters involved, which is more effective than the traditional sequential and separate optimization process. The results indicated that the optimized shielding material, chosen through ANN calculations that determined the appropriate thickness and weight percent of its compositions, can decrease the dose behind the door to lower than the allowable limit for occupational exposure. The stability of ANN was tested by considering uncertainties with the Gaussian distributions of random numbers to the testing data. The results are promising as they indicate that ANNs could be used as a reliable tool for accurately predicting the dosimetric results, providing a drastically powerful alternative approach to the time-consuming Monte Carlo simulations.

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“…This study aims to find the geometry and thickness of each main wall, secondary wall, entrance, maze, and window monitor at BNCT treatment facility using MCNP-4C code [14]. It is such a possible way to build up a BNCT based on D-D Neutron Generator 2.4 MeV system for cancer treatment [15,16].…”

Section: Introductionmentioning

confidence: 99%

Modeling the Radiation Shielding of Boron Neutron Capture Therapy Based on 2.4 Mev D-D Neutron Generator Facility

Mu’Alim¹,

Sardjono²

2018

Tri Dasa Mega

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MODELING THE RADIATION SHIELDING OF BORON NEUTRON CAPTURE THERAPY BASED ON 2.4 MEV D-D NEUTRON GENERATOR FACILITY. Radiation shield at BoronNeutron Capture Therapy (BNCT) facility based on D-D Neutron Generator 2.4 MeV has been modified with pre-designed beam shaping assembly (BSA). Modeling includes the material and thickness used in the radiation shield. This radiation shield is expected to protect workers from radiation doses rate that is not exceed 20 mSv·year-1 of dose limit values. The selected materials are barite, paraffin, polyethylene and lead. Calculations were performed using the MCNPX program with tally F4 to determine the dose rate coming out of the radiation shield not exceeding the radiation dose rate of 10 µSv·hr-1. Design 3 was chosen as the recommended model of the four models that have been made. The 3rd shield design uses a 100 cm thickness of barite concrete as primamary layer to surrounding 100 cm x 100 cm x 166.4 cm room, and a 40 cm borated polyethylene surrounding the barite concrete material. Then 10 cm barite concrete and 10 cm of borated polyethylene are added to reduce the primary radiation straight from the BSA after leaving the main layer. The largest dose rate was 4.58 µSv·h-1 on cell 227 and average radiation dose rate 0.65 µSv·hr-1. The dose rates are lower than the lethal dose that is allowed by BAPETEN for radiation worker lethal dose.Keywords: Radiation shield, tally, radiation dose rate, BSA, BNCT ABSTRAK PEMODELAN PERISAI RADIASI PADA FASILITAS BORON NEUTRON CAPTURE THERAPY BERBASIS GENERATOR NEUTRON D-D 2,4 MeV. Telah dimodelkan perisai radiasi pada fasilitas Boron Neutron Capture Therapy (BNCT) berbasis reaksi D-D pada Neutron Generator 2,4 MeV dengan

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Radiation shielding design of BNCT treatment room for D-T neutron source

Cited by 12 publications

References 12 publications

Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator

Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator

Using ANN for thermal neutron shield designing for BNCT treatment room

Modeling the Radiation Shielding of Boron Neutron Capture Therapy Based on 2.4 Mev D-D Neutron Generator Facility

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