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This work presents an interactive code, WRNS, based on data base interpolation for preliminary neutron shielding design. A wide-range data base of transmission factors is also generated for nine shielding materials having thicknesses up to 200 cm and for neutron energies up to 400 MeV by using adjoint calculations of a one-dimensional discrete ordinates code. In addition, calculation details of the interpolation code are clearly interpreted along with WRNS code applications presented as well.

Standard Bonner spheres and proposed high-sensitivity Bonner cylinders were calibrated in a neutron calibration room, using a (252)Cf source. The Bonner sphere system consists of 11 polyethylene (PE) spheres of various diameters and 4 extended spheres that comprise embedded metal shells. Similar to the design of Bonner spheres, a set of Bonner cylinders was assembled using a large cylindrical (3)He tube as the central probe, which was wrapped using various thicknesses of PE. A layer of lead was employed inside one of the PE cylinders to increase the detection efficiency of high-energy neutrons. The central neutron probe used in the Bonner cylinders exhibited an efficiency of ∼17.9 times higher than that of the Bonner spheres. However, compared with the Bonner spheres, the Bonner cylinders are not fully symmetric in their geometry, exhibiting angular dependence in their responses to incoming neutrons. Using a series of calculations and measurements, this study presents a systematic comparison between Bonner spheres and cylinders in terms of their response functions, detection efficiencies, angular dependences and spectrum unfolding.

Proton therapy accelerators in the energy range of 100-300 MeV could potentially produce intense secondary radiation, which must be carefully evaluated and shielded for the purpose of radiation safety in a densely populated hospital. Monte Carlo simulations are generally the most accurate method for accelerator shielding design. However, simplified approaches such as the commonly used point-source line-of-sight model are usually preferable on many practical occasions, especially for scoping shielding design or quick sensitivity studies. This work provides a set of reliable shielding data with reasonable coverage of common target and shielding materials for 100-300 MeV proton accelerators. The shielding data, including source terms and attenuation lengths, were derived from a consistent curve fitting process of a number of depth-dose distributions within the shield, which were systematically calculated by using MCNPX for various beam-target shield configurations. The general characteristics and qualities of this data set are presented. Possible applications in cases of single- and double-layer shielding are considered and demonstrated.

A dedicated empirical formula for evaluating γ-ray buildup factors of double-layered shields for a point isotropic source has been proposed for the first time by fitting it to the data of Monte Carlo calculations using EGS4 code. The calculated results from the present formula were compared with those using other empirical formulae. The extension of the present formula in the application for multilayered shields was tested for a number of combinations of stratified shields. By using infinite-medium buildup factors of single materials, the present formula will deliver buildup factors of stratified shields without suffering boundary effect instead of the ones of transmission type.

Gamma-ray buildup factors for a point isotropic source in stratified spherical shields have been studied experimentally. Energy absorption rates for an experimental setup of iron spheres in water have been measured using the thermoluminescent dosimeter (TLD)-200. The measured absorbed doses in the TLD have been converted to the absorbed doses in iron and water by using cavity ionization theory. The measured absorbed doses and energy absorption buildup factors in iron and water were then compared with those calculated using the BIGGI-4T and EGS4 codes.

High energy proton beam (>8MeV) is favorable for producing neutrons with high yield. However, the produced neutrons are of high energies. These high energy neutrons can cause severe fast neutron contamination and degrade the BNCT treatment quality if they are not appropriately moderated. Hence, this study aims to briefly discuss the issue, from the viewpoint of fast neutron contamination control, whether high energy proton beam is ideal for AB-BNCT or not. In this study, D2O, PbF4, CaF2, and Fluental(™) were used standalone as moderator materials to slow down 1-, 6-, and 10-MeV parallelly incident neutrons. From the calculated results, we concluded that neutrons produced by high energy proton beam could not be easily moderated by a single moderator to an acceptable contamination level and still with reasonable epithermal neutron beam intensity. Hence, much more complicated and sophisticated designs of beam shaping assembly have to be developed when using high energy proton beams.

Dose estimation of animal experiments affects many subsequent derived quantities, such as RBE and CBE values. It is important to ensure the trustiness of calculated dose of the irradiated animals. However, the dose estimation was normally calculated using simplified geometries and tissue compositions, which led to rough results. This paper introduces the use of treatment planning systems NCTplan and Xplan for the dose estimation. A mouse was taken as an example and it was brought to hospital for micro-PET/CT scan. It was found that the critical organ doses of an irradiated mouse calculated by simplified model were unreliable in comparison to Xplan voxel model. The difference could reach the extent of several tenths percent. It is recommended that a treatment planning system should be introduced to future animal experiments to upgrade the data quality.

In order to monitor the neutron dose in the environment of the Synchrotron Radiation Research Center, the sensitivity of the monitor system to the natural background cosmic-ray neutrons must be established. Large-size BF/sub 3/ counters with and without an 8.5-cm polyethylene enclosure were used to measure the cosmic-ray neutrons in the Nuclear Engineering Building, NTHU. The response functions of the BF/sub 3/ counters were calculated by using ANISN code performed in adjoint mode. By applying the cosmic-ray neutron spectrum and the detector response functions the measured counting rates of the BF/sub 3/ counters were converted to deliver the annual dose of cosmic-ray neutrons. The shielding effect of the concrete slabs and windows of the building was determined by the measured data and compared to that predicted by neutron transport calculations. >

This paper aims to evaluate the performance of the source description of the THOR BNCT beam via different measurement techniques in different phantoms. The measurement included (1) the absolute reaction rate measurement of a set of triple activation foils, (2) the neutron and gamma-ray dose rates measured using the paired ionization chamber method, and (3) the relative reaction rate distributions obtained using the indirect neutron radiography. Three source descriptions, THOR-Y09, surface source file RSSA, and THOR-50C, were tested. The comparison results concluded that THOR-Y09 is a well-tested source description not only for neutron components, but also for gamma-ray component.

Various quality control (QC) and quality assurance (QA) procedures of the boron neutron capture therapy (BNCT) beam at the Tsing Hua Open-pool Reactor (THOR) are established to ensure beam availability and quality. The QC/QA methods mainly employ foil activation and paired ionization chambers, respectively, for beam intensity check and dose assessment. Beam intensity is monitored on-line by using three dead-time corrected fission chambers. In addition to the periodic QC/QA activities regarding beam quality and the monitoring system, the quick QC/QA performed in an all-in-one phantom will be executed less than 70 min before the clinical treatment to guarantee beam quality. The QC/QA procedures have been gradually established and the actual performance satisfied the preset criteria defined for the BNCT facility at THOR.

A refined dose assessment method has been used now in the THOR BNCT facility, which takes into account more delicate corrections, carefully handled calibration factors, and the spectrum- and kerma-weighted k t value. The refined method solved the previous problem of negative derived neutron dose in phantom at deeper positions. With the improved dose assessment, the calculated and measured gamma-ray dose rates match perfectly in a 15×15×15 cm 3 PMMA phantom.

This paper aims to evaluate the effective dose as well as equivalent doses of several organs of an adult hermaphrodite mathematical phantom according to the definition of ICRP Publication 60 for BNCT treatments of brain tumors in the epithermal neutron beam at THOR. The MCNP5 Monte Carlo code was used for the calculation of the average absorbed dose of each organ. The effective doses for a typical brain tumor treatment with a tumor treatment dose of 20 Gy-eq were evaluated to be 0.59 and 0.35 Sv for the LLAT and TOP irradiation geometries, respectively. In addition to the stochastic effect, it was found that it is also likely to produce deterministic effects, such as cataracts and depression of haematopoiesis.

This study aims to measure the two-dimensional (2D) neutron spatial distribution inside a cubic and a head PMMA phantom for the purpose of further comparison with the treatment planning. The measurements were made by using the indirect neutron radiography (INR), which utilized a thin copper foil and the imaging plate. The developed image provides satisfactory spatial resolution and very low statistical error (< 1%). As to the time cost, the whole procedure normally takes less than 3 h. The result shows that the indirect neutron radiography can be a quick and reliable method to provide a 2D neutron spatial distribution inside a phantom.

Four different kinds of microelectronical elements, namely A-111 (functioning as a combined charge-sensitive preamplifier, shaping amplifier and discriminator), A-203 and A-225 (both functioning as charge-sensitive preamplifiers and shaping amplifiers) and A-206 (voltage amplifier and discriminator), supplied by AMPTEK Inc., were used to replace conventional NIM modules to construct radiation-instrumentation systems. The specifications of these microelectronical IC chips were tested. Radiation detection systems using three sets of microelectronical circuits, namely A-111/A-206, A-203/A-206 and A-225/A-206, respectively, coupled with various kinds of detectors, were constructed, whose performances were investigated and compared with those of conventional systems with NIM modules.

The dose rate distributions in the 29,000-Ci 6 0 Co irradiation facility in National Tsing Hua University were investigated by measurements and calculations. The dose rate measurements were performed using radiochromic dye films and an Exradin A2 ion chamber mounted on a PC-controlled motorised vertical translation stage. The calculations were made by using the three-dimensional point kernel code QAD-CGGP with detailed source composition and geometry modelling. The scattered gamma rays from the walls of the irradiation cell were also evaluated by using the Monte Carlo code MCNP.

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the inadequate maze design. The double differential neutron source from the 1 8 O(p,n) 1 8 F reaction was adopted for the calculation. The weight window variance reduction technique, where the weight window was set by applying the adjoint flux, has been implemented in the MCNP run to facilitate the calculation of the dose rates outside the cyclotron room. Dose rates including neutron and gamma-ray components were calculated for some maze shielding modifications.

The electron-photon cascade Monte Carlo code EGS4 has been used to study the energy response of the detector-grade low-cost PIN Si photodiode, which is used as the detector for personal dosimetry applications. The energy deposition distribution in the active volume of the photodiode was calculated. The effect of the discrimination level and the dimension of the active volume on the sensitivity of the photodiode was studied. The sensitivities of the photodiode with a Cu, Sn, Cu+Sn, and Pb compensation filter, respectively, were investigated in detail. The energy response of the Si photodiode was also measured in the X-ray calibration field with effective energies from 46 to 202 keV and by using Cs-137 and Co-60 gamma -ray sources. Good agreement has been obtained between results of the measurement and the Monte Carlo calculation. >

A general review of investigations of gamma-ray buildup factors is presented. Gamma-ray buildup factors for a point isotropic source in stratified spherical shields have been calculated using the o...

A one-dimensional gamma-ray transport code BIGGI 4T has been used to calculate gamma-ray attenuation in single layer and multiple layers of lead and water slabs from a plane monodirectional source....