Hybrid halide perovskite neutron detectors

Interest in fast and easy detection of high-energy radiation (x-, γ-rays and neutrons) is closely related to numerous practical applications ranging from biomedicine and industry to homeland security issues. In this regard, crystals of hybrid halide perovskite have proven to be excellent detectors of x- and γ-rays, offering exceptionally high sensitivities in parallel to the ease of design and handling. Here, we demonstrate that by assembling a methylammonium lead tri-bromide perovskite single crystal (CH3NH3PbBr3 SC) with a Gadolinium (Gd) foil, one can very efficiently detect a flux of thermal neutrons. The neutrons absorbed by the Gd foil turn into γ-rays, which photo-generate charge carriers in the CH3NH3PbBr3 SC. The induced photo-carriers contribute to the electric current, which can easily be measured, providing information on the radiation intensity of thermal neutrons. The dependence on the beam size, bias voltage and the converting distance is investigated. To ensure stable and efficient charge extraction, the perovskite SCs were equipped with carbon electrodes. Furthermore, other types of conversion layers were also tested, including borated polyethylene sheets as well as Gd grains and Gd2O3 pellets directly engulfed into the SCs. Monte Carlo N-Particle (MCNP) radiation transport code calculations quantitatively confirmed the detection mechanism herein proposed.

Tailoring neutron beam properties by target-moderator-reflector optimisation

Compact accelerator-driven neutron sources allow to operate multiple optimised target-moderator-reflector (TMR) units adapted to the requirements of the respective instruments. The compact design of the TMR units allows an efficient coupling of neutron production, neutron moderation and extraction, but requires a novel way of optimisation. The neutronic performance of different TMR units based on polyethylene, heavy water and a mixture of heavy and light water moderators together with Pb and Be reflectors and a borated polyethylene absorber is discussed. Extraction channels for thermal and cold neutrons are investigated regarding the energy and time spectra.

Optimization Simulation of Neutron Shielding Performance of Iron/Borated-Polyethylene Composite Structure

To obtain the optimization design of neutron shielding by iron/borated-polyethylene composite structure, we built a neutron shielding model of two layers of iron and borated-polyethylene. For neutron with various energies, the neutron transmission coefficients of iron/borated-polyethylene composite shield with different thicknesses were obtained by using Monte Carlo method. The simulation results show that, when neutron energy is 14 MeV and the total thickness of the composite shield is 40 cm, 60 cm and 80 cm, the optimal thickness ratio of iron to borated-polyethylene is about 0.7: 0.3, 0.725: 0.275, and 0.75: 0.25, respectively. The optimal thickness ratio of iron to borated-polyethylene is usually higher than the case of iron/polyethylene composite structure.

MODELING THE RADIATION SHIELDING OF BORON NEUTRON CAPTURE THERAPY BASED ON 2.4 MEV D-D NEUTRON GENERATOR FACILITY

Radiation shield at Boron Neutron 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 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 Beam Shaping Assembly (BSA) yang telah didesain sebelumnya. Pemodelan ini dilakukan untuk memperoleh suatu desain perisai radiasi untuk fasilitas BNCT berbasis generator neutron 2,4 MeV. Pemodelan dilakukan dengan cara memvariasikan bahan dan ketebalan perisasi radiasi. Bahan yang dipilih adalah beton barit, parafin, polietilen terborasi dan timbal. Perhitungan dilakukan menggunakan program MCNPX dengan tally F4 untuk menentukan laju dosis yang keluar dari perisai radiasi. Desain periasi radiasi dinyatakan optimal jika radiasi yang dihasilkan diluar perisai radiasi tidak melebihi Nilai Batas Dosis (NBD) yang telah ditentukan oleh BAPETEN. Hasilnya, diperoleh suatu desain perisai radiasi menggunakan lapisan utama beton barit setebal 100 cm yang mengelilingi ruangan 100 cm x 100 cm x 166,4 cm dan polietilen terborasi 40 cm yang mengelilingi bahan beton barit. Kemudian ditambahkan beton barit 10 cm dan polietilen terborasi 10 cm untuk mengurangi radiasi primer yang lurus dari BSA setelah keluar dari lapisan utama. Laju dosis terbesar adalah 4,58 μSv·jam-1 pada sel 227 dan laju dosis rata-rata yang dihasilkan adalah sebesar 0,65 µSv·jam-1. Nilai laju dosis tersebut masih dibawah ambang batas NBD yang diperbolehkan oleh BAPETEN untuk pekerja radiasi.Kata kunci: Perisai radiasi, tally, laju dosis radiasi, BSA, BNCT

Measurements of the Reactor Antineutrino with Solid State Scintillation Detector

Measurements of reactor antineutrino play an important role in the efforts at the frontier of the modern physics. The DANSS collaboration presents preliminary results of a one year run with a cubic meter solid state detector placed below 3.1 GW industrial light water reactor. The experiment is sensitive to sterile neutrino in the most interesting region of mixing parameter space. 2500 scintillation strips of the sensitive volume of the detector have multilayer passive shielding of copper, lead and borated polyethylene and active muon veto. Detector position below the reactor gives an advantage of overburden about 50 m of water equivalent providing factor of six in cosmic muon suppression and eliminating fast neutrons.The detector is placed on a vertically movable platform which allows to change the distance to the reactor core center in the range 10.7-12.7 m within a few minutes. The strips are read out individually by SiPMs and in groups of 50 by PMTs. 5000 inverse beta-decay events per day are collected in the fiducial volume, which is 78% of the whole detector, at the position closest to the reactor. Overburden, active veto and good segmentation of the detector result in an excellent signal to background ratio. The talk is dedicated to the data analysis and preliminary results. The experiment status is also presented.

Analysis of Radiation Effects on Workers and Environment Pilot Plant Boron Neutron Capture Therapy (BNCT)

BNCT is a new method in nuclear technology. The aim of BNCT application is to reduce human risk which used to kills cell targeting characteristic. The impact of using this technology should be considered before it is applied, among the effects of radiation on workers and the surrounding environment BNCT pilot plant. A research on modeling of BNCT pilot plant used a collimator for a 30 MeV cyclotron neutron sources which had been designed from the past research. Radiation shielding modeling for treatment room used MCNPX software. The radiation shielding was concrete baryte on each side that includes coated borated polyethylene 2 cm thick and it is featured with a sliding door with dimensions 220 × 87 × 200 cm coated with stainless steels 2 cm thick. Results obtained value equivalent dose rate of neutron and gamma of each 41.5 µSv.h^-1 and 2.05 µSv.h^-1. Effects of radiation received by workers in the form of deterministic effects did not have a significant are impact.

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

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.