Neutron Dose Rates Research Articles

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor.

The HMGU combined neutron and photon dosemeter

Recently, two independent electronic dosimeters for individual monitoring have been developed at the Helmholtz Center Munich, Germany, one for photons and one for neutrons. Calibration of these dosimeters was performed using mono-energetic neutrons (energy range: 144 keV–14.8 MeV) and photons (12 keV–1.3 MeV), and the resulting calibration curves were compared to Monte Carlo simulations using the GEANT4 tool kit. In the present paper, an integrated dosimeter prototype is described which combines both dosimeters and therefore allows measurement of total dose in mixed neutron/photon radiation fields. It is demonstrated that the neutron channel of the combined dosimeter is insensitive to photons, while the photon channel is insensitive to neutrons. The prototype was tested in the field of an Am-Be neutron source which also includes a photon component. When the Am-Be neutron source was shielded by 9 mm lead, the combined dosimeter indicated a dose rate of 122.6 ± 1.0 μSv/h from neutrons and 7.3 ± 0.1 μSv/h from photons. In contrast, when the Am-Be neutron source was not shielded by lead, the combined dosimeter indicated a dose rate of 125.8 ± 1.0 μSv/h from neutrons and 65.9 ± 0.1 μSv/h from photons indicating a significant contribution from photons to total dose. In a second experiment when the Am-Be neutron source was closer to the combined dosimeter and a 137Cs source was added, a neutron dose rate of 421 ± 20 μSv/h and a photon dose rate of 849 ± 6 μSv/h was obtained. It is concluded that this instrument can be used in mixed photon-neutron fields with doses and dose rates typical for radiation protection purposes.

Shielding evaluation of neutron generator hall by Monte Carlo simulations

A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10 9 n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 µSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper.

NEUTRON AND PHOTON DOSE MAPPING OF A DD NEUTRON GENERATOR.

Neutron generators are an excellent tool that can be effectively utilized in educational institutions for applications such as neutron activation analysis, neutron radiography, and profiling and irradiation effects. For safety purposes, it is imperative that appropriate measures are taken in order to minimize the radiation dose from such devices to the operators, students and the public. This work presents the simulation and measurement results for the neutron and photon dose rates in the vicinity of the neutron generator installed at the University of Sharjah. A very good agreement is found between the simulated and measured dose rates. All of the public dose constraints were found to be met. The occupational dose constraint was also met after imposing a 200 cm no entry zone around the generator room.

Photoneutron radiation field of ducts in barrier of 15 MV medical electron accelerators

Shielding body of the high-energy medical electron accelerators is always penetrated by ducts, which would influence the shielding capability of local barrier. In order to quantitatively analyze the duct's impact on shielding of the photoneutron from 15MV accelerators, the ambient dose equivalent rate and energy spectrum at the center of a typical duct and the external mouth of duct were calculated based on MCNP program for the first time. The results demonstrate that leakage neutrons at the external mouth of duct are mainly thermal neutron, and its dose rate is decreased with the increase of the intersection angle between duct and wall as well as the reduction of duct diameter. When a duct in a diameter no more than 30cm penetrates the wall unidirectionally and the inclined Angle (θ) is 60°, neutron dose rate at the external mouth of duct could meet the requirements of protection. At last, according to the calculation results, some suggestions are proposed for the shielding design of ducts in walls.

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Dose Measurements through the Concrete and Iron Shields under the 100 to 400 MeV Quasi-Monoenergetic Neutron Field (at RCNP, Osaka Univ.)

Shielding benchmark experiments are useful to verify the accuracy of calculation methods for the radiation shielding designs of high-energy accelerator facilities. In the present work, the benchmark experiments were carried out for 244- and 387-MeV quasi-monoenergetic neutron field at RCNP of Osaka University. Neutron dose rates through the test shields, 100-300 cm thick concrete and 40-100 cm thick iron, were measured by four kinds of neutron dose equivalent monitors, three kinds of wide-energy range monitors applied to high-energy neutron fields above 20 MeV and a conventional type rem monitor for neutrons up to 20 MeV, placed behind the test shields. Measured dose rates were compared one another. Measured results with the wide-energy range monitors were in agreement one another for both the concrete and the iron shields. For the conventional type rem monitor, measured results are smaller than those with the wide-energy range monitors for the concrete shields, while that are in agreements for the iron shields. The attenuation lengths were obtained from the measurements. The lengths from all the monitors are in agreement on the whole, though some differences are shown. These results are almost same as those from others measured at several hundred MeV neutron fields.

Neutron irradiation and damage assessment of plastic scintillators of the Tile Calorimeter

Following the comparative study of proton induced radiation damage on various plastic scintillator samples from the ATLAS-CERN detector, a study on neutron irradiation and damage assessment on the same type of samples will be conducted. The samples will be irradiated with different dose rates of neutrons produced in favourable nuclear reactions using a radiofrequency linear particle accelerator as well as from the SAFARI nuclear reactor at NECSA. The MCNP 5 code will be utilized in simulating the neutron transport for determining the dose rate. Light transmission and light yield tests will be performed in order to assess the radiation damage on the scintillators. In addition, Raman spectroscopy and Electron Paramagnetic Resonance (EPR) analysis will be used to characterize the samples after irradiation. The project aims to extent these studies to include radiation assessment damage of any component that processes the scintillating light and deteriorates the quantum efficiency of the Tilecal detector, namely, photomultiplier tubes, wavelength shifting optical fibres and the readout electronics. They will also be exposed to neutron irradiation and the damage assessed in the same manner.

Radioprotection shielding for neutrons induced by the reaction (2H (40 MeV), 12C)

In the framework of design studies for SPIRAL2, the simulation of the neutron flux generated by 40 MeV deuterons on a thick 12 C target was performed and compared to experimental data. The calculation of the dose rate of these neutrons allowed to compare four materials being considered for radioprotection shielding: barites, gypsum, ordinary concrete and heavy concrete. The simulated map of the neutron dose rate in the production building shows a very high dose rate around the neutron source and in the environment of some of the accelerator equipment.

Determination of neutron spectrum outside the lateral shielding of Taiwan Photon Source using high-sensitivity Bonner cylinders

A homemade Bonner cylinder spectrometer was used to determine the neutron spectrum from thermal energy up to several GeV at a heavily-shielded light source facility. The spectrometer, similar to the design of Bonner spheres, features high sensitivity of neutron detection because of using a long cylindrical 3 He proportional counter. The spectrum measurement was performed during the facility commissioning by intentionally parking the injected electrons at the septum of the storage ring. Based on a high-fidelity FLUKA simulation, the predicted neutron spectrum at the location of measurement was adopted as the initial guess of spectrum unfolding. The unfolded result indicated an underestimation of the calculated neutron spectrum in the high-energy portion, leading to a substantial revision of the neutron dose rate at the location.

Development of a multi-element microdosimetric detector based on a thick gas electron multiplier

A prototype multi-element gaseous microdosimetric detector was developed using the Thick Gas Electron Multiplier (THGEM) technique. The detector aims at measuring neutron and gamma-ray dose rates for weak neutron-gamma radiation fields. The multi-element design was employed to increase the neutron detection efficiency. The prototype THGEM multi-element detector consists of three layers of tissue equivalent plastic hexagons and each layer houses a hexagonal array of seven cylindrical gas cavity elements with equal heights and diameters of 17mm. The final detector structure incorporates 21 gaseous volumes. Owing to the absence of wire electrodes, the THGEM multi-element detector offers flexible and convenient fabrication. The detector responses to neutron and gamma-ray were investigated using the McMaster Tandetron 7Li(p,n) neutron source. The dosimetric performance of the detector is presented in contrast to the response of a commercial tissue equivalent proportional counter. Compared to the standard TEPC response, the detector gave a consistent microdosimetric response with an average discrepancy of 8 % in measured neutron absorbed dose. An improvement of a factor of 3.0 in neutron detection efficiency has been accomplished with only a small degradation in energy resolution. However, its low energy cut off is about 6keV/μm, which is not sufficient to measure the gamma-ray dose. This problem will be addressed by increasing the electron multiplication gain using double THGEM layers.

Investigating a cyclotron HM-30 based neutron source for BNCT of deep-seated tumors by using shifting method

We have successfully proposed a simulation of a neutron beam-shaping assembly using MCNPX Code. This simulation study deals with designing a compact, optimized, and geometrically simple beam shaping assembly for a neutron source based on a proton cyclotron for BNCT purpose. Shifting method was applied in order to lower the fast neutron energy to the epithermal energy range by choosing appropriate materials. Based on a set of MCNPX simulations, it has been found that the best materials for beam shaping assembly are 3 cm Ni layered with 7 cm Pb as the reflector and 13 cm AlF3 the moderator. Our proposed beam shaping assembly configuration satisfies 2 of 5 of the IAEA criteria, namely the epithermal neutron flux 1.25 × 109 n.cm-2 s-1 and the gamma dose over the epithermal neutron flux is 0.18×10 -13 Gy.cm 2 n -1. However, the ratio of the fast neutron dose rate over neutron epithermal flux is still too high. We recommended that the shifting method must be accompanied by the filter method to reduce the fast neutron flux.

Biological Shielding Design of In Vitro Test Facility Boron Neutron Capture Therapy (BNCT) at Radial Piercing Beam Port of Kartini Research Reactor Using MCNPX

&lt;p&gt;This research aimed to determine material specification of radiation biological shielding design of neutron and gamma output on radial piercing beamport of Kartini Research Reactor using MCNPX code, as safety radiation protection purpose of in vitro / in vivo irradiation test facility. Refference input data using parameters of neutron and gamma as result simmulation researcher before as follows Ф&lt;sub&gt;th&lt;/sub&gt; is 5.00 x 10&lt;sup&gt;8 &lt;/sup&gt;n.cm&lt;sup&gt;-2&lt;/sup&gt;s&lt;sup&gt;-1&lt;/sup&gt;, Ф&lt;sub&gt;epi&lt;/sub&gt; is 1.23 x 10&lt;sup&gt;8&lt;/sup&gt; n.cm&lt;sup&gt;-2&lt;/sup&gt;s&lt;sup&gt;-1&lt;/sup&gt;, Ф&lt;sub&gt;fast &lt;/sub&gt;is 1.35 x 10&lt;sup&gt;9&lt;/sup&gt; n.cm&lt;sup&gt;-2&lt;/sup&gt;s&lt;sup&gt;-1&lt;/sup&gt;, Ḋ&lt;sub&gt;γ&lt;/sub&gt; is 2.49 x 10&lt;sup&gt;-3&lt;/sup&gt; Sv.s&lt;sup&gt;-1&lt;/sup&gt;, Ḋ&lt;sub&gt;n&lt;/sub&gt; is 3.63 x 10&lt;sup&gt;-1&lt;/sup&gt; Sv.s&lt;sup&gt;-1&lt;/sup&gt;. [ Dwi W.]&lt;/p&gt;Optimation result of simulation using MCNPX to determine specification material for radiation protection safety are parafin block with thickness 75 cm for neutron shielding and covered by material lead (Pb) with thickness 15 cm for gamma shieding. By this optimation for the both thickness materials, determination for neutron and gamma dose rate are as follow , Ḋ&lt;sub&gt;n&lt;/sub&gt; = 1,1 x Sv.s&lt;sup&gt;-1&lt;/sup&gt; and with Ḋ&lt;sub&gt;γ &lt;/sub&gt;= 1,1 x 10&lt;sup&gt;-09 &lt;/sup&gt;Sv.s&lt;sup&gt;-1&lt;/sup&gt;. Those radiation dose rate of neutron and gamma parameters are still under requirement of safety dose standard that is 2.78 x 10&lt;sup&gt;-9&lt;/sup&gt; Sv.s&lt;sup&gt;-1&lt;/sup&gt; as regulation of Regulatory Body BAPETEN. [Perka BAPETEN No. 4 Tahun 2014]. It can be concluded that by the Biological Shielding Design of In Vitro Test Facility Boron Neutron Capture Therapy (BNCT)&lt;em&gt; &lt;/em&gt;at Radial Piercing Beam Port of Kartini Research Reactor Using MCNPX can operated safely.

Source terms, shielding calculations and soil activation for a medical cyclotron

Calculations of the shielding and estimates of soil activation for a medical cyclotron are presented in this work. Based on the neutron source term from the 18O(p,n)18F reaction produced by a 28 MeV proton beam, neutron and gamma dose rates outside the building were estimated with the Monte Carlo code MCNP6 (Goorley et al 2012 Nucl. Technol. 180 298–315). The neutron source term was calculated with the MCNP6 code and FLUKA (Ferrari et al 2005 INFN/TC_05/11, SLAC-R-773) code as well as with supplied data by the manufacturer. MCNP and FLUKA calculations yielded comparable results, while the neutron yield obtained using the manufacturer-supplied information is about a factor of 5 smaller. The difference is attributed to the missing channels in the manufacturer-supplied neutron source terms which considers only the 18O(p,n)18F reaction, whereas the MCNP and FLUKA calculations include additional neutron reaction channels. Soil activation was performed using the FLUKA code.The estimated dose rate based on MCNP6 calculations in the public area is about 0.035 µSv h−1 and thus significantly below the reference value of 0.5 µSv h−1 (2011 Strahlenschutzverordnung, 9 Auflage vom 01.11.2011, Bundesanzeiger Verlag). After 5 years of continuous beam operation and a subsequent decay time of 30 d, the activity concentration of the soil is about 0.34 Bq g−1.

Characterization of the radiation background at the Spallation Neutron Source

We present a survey of the radiation background at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory, TN, USA during routine daily operation. A broad range of detectors was used to characterize primarily the neutron and photon fields throughout the facility. These include a WENDI-2 extended range dosimeter, a thermoscientific NRD, an Arktis 4He detector, and a standard NaI photon detector. The information gathered from the detectors was used to map out the neutron dose rates throughout the facility and also the neutron dose rate and flux profiles of several different beamlines. The survey provides detailed information useful for developing future shielding concepts at spallation neutron sources, such as the European Spallation Source (ESS), currently under construction in Lund, Sweden.

Different dose rate-dependent responses of human melanoma cells and fibroblasts to low dose fast neutrons

Purpose: To analyze the dose rate influence in hyper-radiosensitivity (HRS) of human melanoma cells to very low doses of fast neutrons and to compare to the behaviour of normal human skin fibroblasts.Materials and methods: We explored different neutron dose rates as well as possible implication of DNA double-strand breaks (DSB), apoptosis, and energy-provider adenosine-triphosphate (ATP) levels during HRS.Results: HRS in melanoma cells appears only at a very low dose rate (VLDR), while a high dose rate (HDR) induces an initial cell-radioresistance (ICRR). HRS does not seem to be due either to DSB or to apoptosis. Both phenomena (HRS and ICRR) appear to be related to ATP availability for triggering cell repair. Fibroblast survival after neutron irradiation is also dose rate-dependent but without HRS.Conclusions: Melanoma cells or fibroblasts exert their own survival behaviour at very low doses of neutrons, suggesting that in some cases there is a differential between cancer and normal cells radiation responses. Only the survival of fibroblasts at HDR fits the linear no-threshold model. This new insight into human cell responses to very low doses of neutrons, concerns natural radiations, surroundings of accelerators, proton-therapy devices, flights at high altitude. Furthermore, ATP inhibitors could increase HRS during high-linear energy transfer (high-LET) irradiation.

Neutron production and dose rate in the IFMIF/EVEDA LIPAc injector beam commissioning

The construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress in Rokkasho, Japan, and the deuteron beam commissioning of the injector began in July 2015. Due to the huge beam current of 125mA, a large amount of d-D neutrons are produced in the commissioning. The neutron streaming effect through pipe penetrations and underground pits may dominate the radiation dose at the outside of the accelerator vault during the injector operation. In the present study the effective dose rate expected during the injector commissioning was analyzed by a Monte Carlo calculation and compared with the measured value. For the comparison it is necessary to know the total neutron production yield in the accelerator vault, thus a dedicated neutron production yield monitoring system was developed. The yield obtained was smaller than that previously reported in a literature by a factor of a few and seems to depend on some beam conditions. From the comparison it was proved that the calculation always provides a conservative estimate and the dose rates in places where occupational works can always access and the controlled area boundary are expected to be far less than the legal criteria throughout the injector commissioning.

Neutron Doses in the EAST Hall Under the NBI Plasma Discharge Mode

Nuclear fusion based on Tokamak has been well and attentively studied during the past years in many countries. There is no denying that some radiation particles and rays will be produced during each nuclear fusion discharge, especially neutrons which will be largely emitted out with high energies. To well obtain the neutrons’ dose rates during the plasma discharge, 3He-based neutron detector is adopted. To fully obtain the neutron dose rates in the EAST hall, three neutron detectors are set at three monitoring sites which have long and different distances to the monitoring personal computer, and the monitoring system of neutrons based on the RS-485 differential communication mode is designed and set up. Then, based on the neutron dose rates saved by the monitoring system, the relations between the NBI’s powers and neutron doses are obtained and schematically presented, also some discussions are made.

Design, Construction, and Modeling of a252Cf Neutron Irradiator

Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s) in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

Dosimetry in Thermal Neutron Irradiation Facility at BMRR

Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single more » organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7 % lower than the statistical mean derived from the Monte-Carlo modeling (5% uncertainty). The dose rate measured by ion chambers was 6 - 10 % lower than the output tallies (7% uncertainty). The detailed dosimetry that was performed at the TNIF for the NCT will be described. « less