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Smoking rates in the military are evaluated through questionnaire surveying. Because the accurate identification of smokers facilitates the provision of smoking cessation services, this study conducted urine cotinine concentration testing to verify the accuracy of self-reported smoking behavior by female volunteer soldiers and analyzed the effects of second-hand smoking on urine cotinine concentrations. This study is a cross-sectional study conducted using purposive sampling on female volunteer soldiers receiving training at the Taichung Recruit Training Center in May 2014. This study simultaneously collected questionnaires and urine samples, and urine samples were analyzed with an enzyme-linked immunosorbent assay. The self-reported smoking rate of female volunteer soldiers was 19.3%, whereas the smoking rate as determined by urine cotinine concentration testing was 26.3%, indicating an overall underestimation of 7.0%. Chi-square (χ2) goodness of fit test results indicated that the distribution of self-reported smoking behaviors and that verified from urine cotinine concentration testing were significantly different. The sensitivity of self-reported smoking behavior was 66.7% with a specificity of 97.6%. There was no significant association between second-hand smoking and urine cotinine concentrations. Questionnaire survey self-reporting methods could underestimate the smoking behavior of female volunteer soldiers and routine testing with biochemical verification is necessary.

TLD-400 (CaF2:Mn) chips were applied for the gamma-ray dose measurement in a PMMA phantom exposed to a BNCT beam because of their very low neutron sensitivity. Since TLD-400 chips possess an adequate amount of Mn activator they have been employed in this work simultaneously for neuron activation measurement. The self-irradiation TL signals owing to the decay of the neutron induced 56Mn activity have been applied for a calibration of the TLD-400 chip in situ, where the activities were measured by an HPGe detector system and the energy deposition per disintegration of 56Mn was calculated by applying a Monte Carlo code. It was accidentally found that the irradiated TLD-400 chips were capable of emitting prominent scintillation lights owing to the induced 56Mn activity, which can easily be recorded by the TLD reader without heating and after a calibration can be used to determine the 56Mn activity.

In this work TLD-200 (CaF2:Dy) chips were used to measure the gamma-ray doses in a PMMA phantom exposed to the BNCT beam at Tsing Hua Open-pool Reactor (THOR). The self-irradiation component induced by the decay of Dy-165 was corrected. The neutron dose contamination was less than 0.3%. The Dy content in the TLD-200 chip was determined by using the modified absolute calibration method of NAA. The self-irradiation TL signal was also applied for the in situ calibration.

In this work a modified absolute calibration method of neutron activation analysis (NAA) was proposed. Reaction rates per atom of elements of interest were calculated by using MCNPX Monte Carlo code with a well established disk surface neutron source. Time dependency of the absolute reaction rate per atom produced during the irradiation was achieved by utilizing a well calibrated neutron monitoring system. Three activation foils with given compositions, namely, Au(1%)/Al, Mn(88%)/Ni and Cu(100%) and a SS304 sample were measured to test the adequacy and performance of this proposed modified absolute calibration method.

The neutron beam monitoring system is indispensable to BNCT facility in order to achieve an accurate patient dose delivery. The neutron beam monitoring of a reactor-based BNCT (RB-BNCT) facility can be implemented through the instrumentation and control system of a reactor provided that the reactor power level remains constant during reactor operation. However, since the neutron flux in reactor core is highly correlative to complicated reactor kinetics resulting from such as fuel depletion, poison production, and control blade movement, some extent of variation may occur in the spatial distribution of neutron flux in reactor core. Therefore, a dedicated neutron beam monitoring system is needed to be installed in the vicinity of the beam path close to the beam exit of the RB-BNCT facility, where it can measure the BNCT beam intensity as closely as possible and be free from the influence of the objects present around the beam exit. In this study, in order to demonstrate the importance of a dedicated BNCT neutron beam monitoring system, the signals originating from the two in-core neutron detectors installed at THOR were extracted and compared with the three dedicated neutron beam monitors of the THOR BNCT facility. The correlation of the readings between the in-core neutron detectors and the BNCT neutron beam monitors was established to evaluate the improvable quality of the beam intensity measurement inferred by the in-core neutron detectors. In 29 sampled intervals within 16 days of measurement, the fluctuations in the mean value of the normalized ratios between readings of the three BNCT neutron beam monitors lay within 0.2%. However, the normalized ratios of readings of the two in-core neutron detectors to one of the BNCT neutron beam monitors show great fluctuations of 5.9% and 17.5%, respectively.

The paired ionization chambers (ICs) technique is commonly employed to determine neutron and photon doses in radiology or radiotherapy neutron beams, where neutron dose shows very strong dependence on the accuracy of accompanying high energy photon dose. During the dose derivation, it is an important issue to evaluate the photon and electron response functions of two commercially available ionization chambers, denoted as TE(TE) and Mg(Ar), used in our reactor based epithermal neutron beam. Nowadays, most perturbation corrections for accurate dose determination and many treatment planning systems are based on the Monte Carlo technique. We used general purposed Monte Carlo codes, MCNP5, EGSnrc, FLUKA or GEANT4 for benchmark verifications among them and carefully measured values for a precise estimation of chamber current from absorbed dose rate of cavity gas. Also, energy dependent response functions of two chambers were calculated in a parallel beam with mono-energies from 20 keV to 20 MeV photons and electrons by using the optimal simple spherical and detailed IC models. The measurements were performed in the well-defined (a) four primary M-80, M-100, M120 and M150 X-ray calibration fields, (b) primary 60Co calibration beam, (c) 6 MV and 10 MV photon, (d) 6 MeV and 18 MeV electron LINACs in hospital and (e) BNCT clinical trials neutron beam. For the TE(TE) chamber, all codes were almost identical over the whole photon energy range. In the Mg(Ar) chamber, MCNP5 showed lower response than other codes for photon energy region below 0.1 MeV and presented similar response above 0.2 MeV (agreed within 5% in the simple spherical model). With the increase of electron energy, the response difference between MCNP5 and other codes became larger in both chambers. Compared with the measured currents, MCNP5 had the difference from the measurement data within 5% for the 60Co, 6 MV, 10 MV, 6 MeV and 18 MeV LINACs beams. But for the Mg(Ar) chamber, the derivations reached 7.8–16.5% below 120 kVp X-ray beams. In this study, we were especially interested in BNCT doses where low energy photon contribution is less to ignore, MCNP model is recognized as the most suitable to simulate wide photon–electron and neutron energy distributed responses of the paired ICs. Also, MCNP provides the best prediction of BNCT source adjustment by the detector’s neutron and photon responses.

Reference man has been widely used for external and internal dose evaluation of radiation protection. The parameters of the mathematical model of organs suggested by the International Commission of Radiological Protection (ICRP) are adopted from the average data of Caucasians. However, the organ masses of Asians are significantly different from the data of Caucasians, leading to potentially dosimetric errors. In this study, a total of 40 volunteers whose heights and weights corresponded to the statistical average of Taiwanese adults were recruited. Magnetic resonance imaging was performed, and T2-weighted images were acquired. The Taiwanese reference man and woman were constructed according to the measured organ masses. The dose conversion coefficients (DCFs) for anterior-posterior (AP), posterior-anterior (PA), right lateral (RLAT) and left lateral (LLAT) irradiation geometries were simulated. For the Taiwanese reference man, the average differences of the DCFs compared with the results of ICRP-74 were 7.6, 5.1 and 11.1 % for 0.1, 1 and 10 MeV photons irradiated in the AP direction. The maximum difference reached 51.7 % for the testes irradiated by 10 MeV photons. The size of the trunk, the volume and the geometric position of organs can cause a significant impact on the DCFs for external exposure of radiation. The constructed Taiwanese reference man and woman can be used in radiation protection to increase the accuracy of dose evaluation for the Taiwanese population.

Herndon proposed a nuclear fission reactor at the center of the Earth to explain changes in the geomagnetic field and the 3 He/ 4 He ratios observed from deep mantle sources. This study investigated the neutronic properties of the planetary-scale reactor by performing rigorous depletion simulations over geologic time by using a modified TRITON sequence in SCALE6. We also conducted analytical calculations of the rates of change of various actinides in the reactor core to identify the primary mechanisms involved in the nuclear system as a function of the operating time. The sound agreement between analytical and TRITON calculations on the predicted variations of the amounts of important actinides revealed that (1) the hypothetical nuclear georeactor is a fast-spectrum converter reactor burning only 235 U; (2) the efficiency of fuel conversion approaches 0.9, and can be sustained for billions of years based on the cycle of 238 U/ 239 Pu/ 235 U, rather than of 238 U/ 239 Pu or 232 Th/ 233 U; and (3) under appropriate conditions, the georeactor can operate at a constant power of 3 TW for up to 6.5 billion years.

In this work neutron activation surveys on 26 patients of recurrent head and neck cancer following boron neutron capture therapy (BNCT) were performed. By means of qualitative neutron activation analysis (NAA), radioactive nuclides activated from neutron capture reaction within patients were identified. The estimated activities of the corresponding radioactive nuclides were achieved based on several assumptions. Gamma-ray dose rate measurements utilizing a survey meter were taken at three chosen distances, namely, 0 cm, 30 cm and 100 cm away from the irradiated region of patients. The variations of gamma-ray dose rate against time were obtained. An approach to estimating the gamma-ray dose rate resulted from each individual radioactive nuclide is also presented.

Reference phantoms are widely applied to evaluate the radiation dose for external exposure. However, the frequently used reference phantoms are based on Caucasians. Dose estimation for Asians using a Caucasian phantom can result in significant errors. This study recruited 40 volunteers whose body sizes are close to the average Taiwanese population. Magnetic resonance imaging was performed to obtain the organ volume for construction of the Taiwanese reference man (TRM) and Taiwanese reference woman (TRW). The dose conversion coefficients (DCC) resulting from photo beams in anterior-posterior, posterior-anterior, right-lateral, left-lateral, and isotropic irradiation geometries were estimated. In the anterior-posterior geometry, the mean DCC differences among organs between the TRM and ORNL phantom at 0.1, 1, and 10 MeV were 7.3%, 5.8%, and 5.2%, respectively. For the TRW, the mean differences from the ORNL phantom at the three energies were 10.6%, 7.4%, and 8.3%. The DCCs of the Taiwanese reference phantoms and the ORNL phantom presented similar trends in other geometries. The torso size of the phantom and the mass and geometric location of the organ have a significant influence on the DCC. The Taiwanese reference phantoms can be used to establish dose guidelines and regulations for radiation protection from external exposure.

For BNCT treatments, in addition to tumor target doses, non-negligible doses will result in all the remaining organs of the body. This work aims to evaluate the effective dose as well as the average absorbed doses of each of organs of patients with brain tumor treated in the BNCT epithermal neutron beam at THOR. The effective doses were evaluated according to the definitions of ICRP Publications 60 and 103 for the reference male and female computational phantoms developed in ICRP Publication 110 by using the MCNP5 Monte Carlo code with the THOR-Y09 beam source. The effective dose acquired in this work was compared with the results of our previous work calculated for an adult hermaphrodite mathematical phantom. It was found that the effective dose for the female voxel phantom is larger than that for the male voxel phantom by a factor of 1.2-1.5 and the effective dose for the voxel phantom is larger than that for the mathematical phantom by a factor of 1.3-1.6. For a typical brain tumor BNCT, the effective dose was calculated to be 1.51Sv and the average absorbed dose for eye lenses was 1.07Gy.

To introduce our experience of treating locally and regionally recurrent head and neck cancer patients with BNCT at Tsing Hua Open-Pool Reactor in Taiwan, 12 patients (M/F=10/2, median age 55.5 Y/O) were enrolled and 11 received two fractions of treatment. Fractionated BNCT at 30-day interval with adaptive planning according to changed T/N ratios was feasible, effective and safe for selected recurrent head and neck cancer in this trial.

This study used simulations of galactic cosmic ray in the atmosphere to investigate the neutron background environment in Taiwan, emphasising its altitude dependence and spectrum variation near interfaces. The calculated results were analysed and compared with two measurements. The first measurement was a mobile neutron survey from sea level up to 3275 m in altitude conducted using a car-mounted high-sensitivity neutron detector. The second was a previous measured result focusing on the changes in neutron spectra near air/ground and air/water interfaces. The attenuation length of cosmic-ray neutrons in the lower atmosphere was estimated to be 163 g cm(-2) in Taiwan. Cosmic-ray neutron spectra vary with altitude and especially near interfaces. The determined spectra near the air/ground and air/water interfaces agree well with measurements for neutrons below 10 MeV. However, the high-energy portion of spectra was observed to be much higher than our previous estimation. Because high-energy neutrons contribute substantially to a dose evaluation, revising the annual sea-level effective dose from cosmic-ray neutrons at ground level in Taiwan to 35 μSv, which corresponds to a neutron flux of 5.30 × 10(-3) n cm(-2) s(-1), was suggested.

The (7)Li(p,xn)(7)Be nuclear reaction, based on the low-energy protons, could produce soft neutrons for accelerator-based boron neutron capture therapy (AB-BNCT). Based on the fact that the induced neutron field is relatively divergent, the relationship between the incident angle of proton beam and the neutron beam quality was evaluated in this study. To provide an intense epithermal neutron beam, a beam-shaping assembly (BSA) was designed. And a modified Snyder head phantom was used in the calculations for evaluating the dosimetric performance. From the calculated results, the intensity of epithermal neutrons increased with the increase in proton incident angle. Hence, either the irradiation time or the required proton current can be reduced. When the incident angle of 2.5-MeV proton beam is 120°, the required proton current is ∼13.3 mA for an irradiation time of half an hour.

In this work, we established a real-time multichannel detector system for large-scale environmental radiation survey. The system consists of a Berkeley Lab cosmic-ray detector for the measurement of muons, a moderated He detector for neutrons, a NaI(Tl) as well as a HPGe spectrometer for terrestrial gamma rays, and a high pressure ionization chamber, which responses to both cosmic rays and terrestrial gamma rays. The measured data of these five detector channels as well as signals from a global position system are acquired and displayed by a LabVIEW program on a notebook computer every one minute (10 min for HPGe and 20 min for NaI(Tl) gamma-ray spectrometers). The advantage and merit of this multichannel detector system have been demonstrated by test runs with the equipment set up both on a train and on a bus to show in real time the variation of cosmic-ray intensity as a function of altitude and the gamma-ray dose rates as well as the relative concentrations of the natural radioactive elements over a large survey area. This real-time multichannel detector system was originally developed for natural background radiation measurements for popular science education to general public, however, due to the specific features of each detector of the system it appears to be an ideal system for radiation survey over a large radiation polluted or contaminated area resulted from a severe nuclear accident, such as Fukushima Dai-Ichi Nuclear Power Plant accident.

Energy spectrum determination of an epithermal neutron beam for BNCT is always a critical issue to its dose accuracy. The spectrum is generally determined by multiple activation foils with few-channel unfolding codes. This study aims to make improvement to the activation-detector-based unfolding technique via introducing specially designed detector complexes in addition to the commonly applied multiple activation foils. The extra detector complexes can obtain valuable information from planned, artificial perturbations posed on the incident epithermal neutron beam, without the need of modifying the unfolding algorithm. Two different types of BNCT beams, one is based on nuclear reactor and the other one is based on low energy accelerator, with two different conditions were tested: under- and over-moderated cases. The spectra adjusted by using the original foil package containing 11 foils and the improved system are compared based on the square root of the sum of deviation squares. The specially designed detector complexes have shown the great ability to significantly reduce the deviations between the unfolded spectra and standard solutions. For a seriously under-moderated case tested in this work, the improvement can even reach more than 400% in deviation reduction.

Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.

High energy photon (over 10 MeV) and neutron beams adopted in radiobiology and radiotherapy always produce mixed neutron/gamma-ray fields. The Mg(Ar) ionization chambers are commonly applied to determine the gamma-ray dose because of its neutron insensitive characteristic. Nowadays, many perturbation corrections for accurate dose estimation and lots of treatment planning systems are based on Monte Carlo technique. The Monte Carlo codes EGSnrc, FLUKA, GEANT4, MCNP5, and MCNPX were used to evaluate energy dependent response functions of the Exradin M2 Mg(Ar) ionization chamber to a parallel photon beam with mono-energies from 20 keV to 20 MeV. For the sake of validation, measurements were carefully performed in well-defined (a) primary M-100 X-ray calibration field, (b) primary 60Co calibration beam, (c) 6-MV, and (d) 10-MV therapeutic beams in hospital. At energy region below 100 keV, MCNP5 and MCNPX both had lower responses than other codes. For energies above 1 MeV, the MCNP ITS-mode greatly resembled other three codes and the differences were within 5%. Comparing to the measured currents, MCNP5 and MCNPX using ITS-mode had perfect agreement with the 60Co, and 10-MV beams. But at X-ray energy region, the derivations reached 17%. This work shows us a better insight into the performance of different Monte Carlo codes in photon–electron transport calculation. Regarding the application of the mixed field dosimetry like BNCT, MCNP with ITS-mode is recognized as the most suitable tool by this work.

This study aims to measure the neutronic characteristics of the 14-cm diameter boron neutron capture therapy beam at the Tsing Hua Open-pool Reactor and to provide essential information for the neutron source adjustment and validation. The measurements were performed both in air and inside a cubic PMMA phantom at the beam, with and without an 18-cm long extended PE collimator. The neutron intensity was determined by neutron activation analysis; the indirect neutron radiography and the cadmium difference method were coupled to provide the two-dimensional neutron flux distributions. According to the measurements, the angular distribution is highly forward for epithermal neutrons but much more divergent for thermal ones. A PE extended collimator will modify the beam characteristics and could benefit the treatment for head and neck tumors, which often needs higher therapeutical boron dose at the shallow region.

This paper presents a way to determine the angular and spatial distributions of the thermal neutron source strength of a boron neutron capture therapy (BNCT) beam. The experiments applied 1) the indirect neutron radiography, 2) the cadmium difference method, and 3) the instrumental neutron activation analysis. The measured data were processed by the spectrum deconvolution technique to resolve into a proper set of angular and spatial distributions. This paper took the epithermal neutron beam of the BNCT facility at the Tsing Hua Open-pool Reactor as an example.

This study is a part of the beam comparison campaign, inter-center dose comparison, between boron neutron capture therapy facilities at the Tsing Hua Open-pool Reactor and the High Flux Reactor. The clinical information exchange can improve the dosimetry uncertainty for medical physics in a mixed field. The method of paired Mg(Ar) and TE(TE) ionization chambers was used to determine the gamma-ray and neutron dose rates. Furthermore, activation foils, including gold, copper, and manganese, were employed to estimate the thermal and epithermal neutron fluxes. Measurements were performed free in air and also in a PMMA phantom. All the chambers were calibrated using a 60Co primary standard source at the Institute of Nuclear Energy Research, Taiwan. Spectrum dependent neutron sensitivity of TE(TE) chamber is one of the important parameters to evaluate dose components. The requested neutron spectra were calculated by the Monte Carlo code MCNP. The measured thermal neutron fluxes, gamma-ray and neutron dose rates of the THOR beam in the phantom were 2.6, 2.2, and 2.1 times of the HFR beam at 2.5-cm depth, respectively. The higher thermal neutron flux and neutron and gamma-ray dose rates are due to the higher epithermal neutron beam intensity of the THOR.

This study aims to determine the activation contamination in the metal-based ionization chambers, i.e. Mg(Ar)- and Al(Ar)-chambers. The contamination was estimated by directly measuring the signal caused by the radioactive nuclides after the irradiation. The INAA gamma ray spectra clearly showed that the ionization chambers were activated. The activation contamination could comprise more than 2% of the total measured current of the Mg(Ar)-chamber, and more than 50% of the Al(Ar)-chamber, both irradiated free-in-air at the THOR epithermal neutron beam. The activation contamination provides concrete evidence of the necessity of the thermal neutron sensitivity correction.

This work aims to measure different components of natural background radiation on a train. A radiation measurement system consisting of four types of radiation detectors, namely, a Berkeley Lab cosmic-ray detector, moderated (3)He detector, high-pressure ionisation chamber and NaI(Tl) spectrometer, associated with a global positioning system unit was established for this purpose. For the commissioning of the system, a test measurement on a train along the railway around the northern Taiwan coast from Hsinchu to Hualien with a distance of ∼ 275 km was carried out. No significant variation of the intensities of the different components of natural background radiation was observed, except when the train went underground or in the tunnels. The average external dose rate received by the crew of the train was estimated to be 62 nSv h(-1).

In this work a Berkeley Lab cosmic ray detector was used to measure the angular distribution of the cosmic-ray muon fluence rate. Angular response functions of the detector at each measurement orientation were calculated by using the FLUKA Monte Carlo code, where no energy attenuation was taken into account. Coincidence counting rates were measured at ten orientations with equiangular intervals. The muon angular fluence rate spectrum was unfolded from the measured counting rates associated with the angular response functions using both the MAXED code and the parameter adjusting method.

For the nuclear safeguard purpose, this work aims to nondestructively determine the plutonium content in the Taiwan Research Reactor (TRR) spent fuel rods in the storage pool before the stabilization process, which transforms the metal spent fuel rods into oxide powder. A SPent-fuel-Neutron-Counter (SPNC) system was designed and constructed to carry out underwater scan measurements of neutrons emitting from the spent fuel rod, from which the 240 Pu mass in the fuel rod will be determined. The SAS2 H control module of the SCALE 5.1 code package was applied to calculate the 240 Pu-to-Pu mass ratio in the TRR spent fuel rod according to the given power history. This paper presents the methodology and design of our detector system as well as the measurements of four TRR spent fuel rods in the storage pool and the comparison of the measured results with the facility declared values.

This paper aims to measure the neutron flux distribution over the cross section of the boron neutron capture therapy epithermal neutron beam at the Tsing Hua Open-pool Reactor (THOR) by using indir...

The paper aims to estimate the residual activity in the concrete shielding of a nuclear power plant (NPP) after 40 yr of design service life and to determine if the whole massive concrete shielding must be treated as radioactive waste for future decommissioning. The process was a combination of experiment and calculation. Nonradioactive concrete samples collected from the Lungmen NPP were measured to determine the initial concentrations of major, minor, and trace elements in the concrete shielding by neutron activation analysis, inductively coupled plasma-mass spectrometry, and elemental analysis. The neutron flux distribution and depth-dependent cross sections, which were generated by SAS1, in the 60-cm-thick reactor shielding wall and 200-cm-thick dry well wall of the Lungmen NPP were fed to the ORIGEN-S code to calculate the activity distribution in the concrete shielding after 40 yr of reactor full-power operation. Comparing the activity with the exemption levels, it was found that the dry well wall of the Lungmen NPP can be handled as construction waste for immediate decommissioning. However, most of the reactor shielding wall must be treated as radioactive waste even after a 25-yr cooling time.

The site dose rate of a spent-fuel storage facility to the populace is a major concern in a radiation protection project. Shielding analysis of the facility must be performed to ensure that the nearby dose rates are within regulation limitations. The purpose of this study was to simulate an independent spent-fuel storage installation (ISFSI) storage facility with different methods and different conditions for validation and analysis. The discrete ordinates code DORT and the SKYSHINE III code were used for the cask surface flux estimation and the site dose rate calculation, respectively. The Monte Carlo code MCNP was also utilized to estimate the surface dose rate and site dose rate by its subsequent calculation. Various cask decay heats (23, 14, and 7 kW/cask) were considered as the source conditions. Afacility layout composed of 30 casks was also simulated by the MCNP code and analyzed for the cask self-shielding effect to a certain detecting point. For a single storage cask, comparisons of the site dose rates calculated by different methods were carried out at variant distances. For the layout simulation, the calculated results indicated that the self-shielding effect could be roughly classified into several groups according to the location of the storage cask, and a factor could be assigned to each group. These classified factors might help to infer the site dose in variant layout designs. The site dose rates calculated by different codes were compared for the whole facility, too. In spite of the similar dose rates on the cask surface, the difference of site dose rates changes with decay heat. The layout study of the ISFSI facility could offer information to make the site dose estimation more efficient as many layout assessments are needed.

The purpose of this research is to estimate the wall effect of spherical graphite-walled cavity chambers using the Monte Carlo method to establish a 60Co air-kerma standard at the National Radiatio...

This paper aims to measure the spectra of HB11 (high flux reactor, HFR) and the Tsing Hua open-pool reactor (THOR) boron neutron capture therapy (BNCT) beams by multiple activation foils. The self-shielding corrections were made with the aid of MCNP calculations. The initial spectra were adjusted by a sophisticated process named coarse-scaling adjustment using SAND-EX, which can adjust a given coarse-group spectrum into a fine-group structure, i.e. 640 groups, with excellent continuity. The epithermal neutron flux of the THOR beam is about three times of HB11. The thermal neutron flux, boron and gold reaction rates along the central axis of a PMMA phantom are calculated for both adjusted spectra for comparison.

Activation foils, stacked in a package, together with a spectrum adjustment code are generally the means to characterize the neutron spectrum of a BNCT epithermal neutron beam. When using foil stacks, the phenomenon of neutron scattering is appreciably enhanced, especially in an epithermal neutron beam, as compared to a thermal neutron field. Contrary to the conventional idea of the self-shielding effect, self-shielding correction factors in most of the foils in the stack are larger than one. This paper demonstrates a way to correct the complex self-shielding effect using groupwise cross-sections modified for self-shielding, as calculated by the Monte Carlo neutron transport code, MCNP. This methodology can solve the perturbation problems caused by the foil stack applied in an epithermal neutron beam. In this study, the adjusted spectrum of the HB11 epithermal neutron beam, dedicated for BNCT research in Petten, The Netherlands, is utilized and two adjusted spectra are presented with and without the self-shielding correction. The difference between the spectra is highly significant and demonstrates that the self-shielding correction cannot be ignored.

In order to provide an improved and reliable neutron source description for treatment planning in boron neutron capture therapy (BNCT), a spectrum adjustment procedure named coarse-scaling adjustment has been developed and applied to the neutron spectrum measurements of both the Tsing Hua Open-pool Reactor (THOR) epithermal neutron beam in Taiwan and the High Flux Reactor (HFR) in The Netherlands, using multiple activation detectors. The coarse-scaling adjustment utilizes a similar idea as the well-known two-foil method, which adjusts the thermal and epithermal neutron fluxes according to the Maxwellian distribution for thermal neutrons and 1/ E distribution over the epithermal neutron energy region. The coarse-scaling adjustment can effectively suppress the number of oscillations appearing in the adjusted spectrum and provide better smoothness. This paper also presents a sophisticated 9-step process utilizing twice the coarse-scaling adjustment which can adjust a given coarse-group spectrum into a fine-group structure, i.e. 640 groups, with satisfactory continuity and excellently matched reaction rates between measurements and calculation. The spectrum adjustment algorithm applied in this study is the same as the well-known SAND-II.

This paper presents a detailed comparison of the surface dose rate calculations for the NAC-UMS spent fuel storage cask by using MCNP and SAS4 computer codes. Their accuracy and computation efficiencies are compared. For such a real world deep penetration and streaming problem, effective variance reduction techniques are indispensable for a Monte Carlo simulation to obtain results of small statistic errors within reasonable computing time. The TORT-coupled MCNP calculation based on the CADIS methodology has been used in this study. The main differences between MCNP and SAS4 calculations are the underlying cross-section libraries and the adjoint functions used for variance reduction in Monte Carlo simulations. The cross-section libraries and their formats should be the root cause for some significant discrepancies between the MCNP and SAS4 results. In addition, limited by the 1D adjoint biasing scheme, SAS4 is inefficient in calculating the dose rates near inlet/outlet apertures. Considering all the computer time spent and the statistical errors of results obtained, the overall computation efficiency by using the TORT-coupled MCNP is better than SAS4 in the shielding calculations of spent fuel storage casks. More specifically, although the SAS4 efficiency is better when the cask side calculation is the only concern, the TORT-coupled MCNP technique is more efficient for the gamma-ray transport in cask top configurations and almost all the vent-streaming problems.

In this study, discrete ordinates and Monte Carlo methods were applied to solve the radiation transport problem for a simplified spent fuel storage cask considering fixed neutron and gamma-ray sources. The results were compared, and the causes for their differences were investigated. In addition, a hybrid method based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been adopted to accelerate the Monte Carlo simulations. CADIS utilizes a deterministic adjoint function for variance reduction through source biasing and consistent transport biasing. The problem encountered and its possible solution for applying the source biasing in such a large volume source are described. Compared with the unbiased case, the computational efficiency is improved by a factor of several tens for neutron transport, and the efficiency is increased tremendously by about five orders of magnitude for gamma-ray transport. It has been demonstrated that the biasing scheme applied here is very effective in the shielding calculations for a spent fuel storage cask using the Monte Carlo method.

The Cyclotron Center of Shin Kong Memorial Hospital is the most active [ 18 F]FDG producer in Taiwan. Any improvement of the production efficiency of the MINItrace™ cyclotron is valuable. A theoretical activity of [ 18 F]FDG has been calculated from the proton stopping power and the interaction cross-section of the 18 O(p,n) 18 F reaction to estimate the efficiency of the production process. A high-sensitivity moderated 3 He neutron detector was installed to investigate the real-time condition of the target irradiation by monitoring the associated neutron production. Based on the results of this study, a graphical user interface program has been developed for operators to arrange the operation conditions and to analyze the performance of the production process.

The radiation doses resulting from diagnostic X-ray examinations are routinely measured in terms of entrance skin exposure (ESE). In this study, for the purpose of radiation protection, the radiation doses received from chest and abdomen X-ray tests were evaluated in terms of equivalent dose and effective dose. The dose calculations were conducted by using the MCNP Monte Carlo code and an adult hermaphrodite mathematical phantom. The effects of both operating high voltage and projection geometry on the effective dose were investigated. The absolute values of the effective doses may be provided from the national average ESE.

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted.

This study investigated the characteristics of gas bremsstrahlung and induced photoneutrons from the electron storage ring in National Synchrotron Radiation Research Center (NSRRC). Gas bremsstrahlung is generated in the vacuum chamber of an electron storage ring by interactions of the circulating electrons with the atoms of the residual gas. The emitted photons with sufficiently high energies can further produce neutrons by means of photonuclear interactions. Photoneutrons produced in a lead target struck by gas bremsstrahlung were measured as a function of target thickness by using high-efficiency neutron detectors in a well-defined configuration. A Monte Carlo simulation of the whole process including generation of gas bremsstrahlung, production of photoneutrons and subsequent neutron transport has been performed in an integrated manner and the results were directly compared with the measurements on an absolute basis. The agreement between them is very satisfactory and hence validates the evaluation of the characteristics of gas bremsstrahlung and induced photoneutrons in NSRRC.

The radon progeny concentration in raindrops as well as in cloud droplets has been determined based on a simplified rainout model from the in situ measurement of gamma-ray intensity using an HPGe detector. The radon concentration in cloud air was deduced from the radon progeny concentration in cloud droplets by applying the Junge's equation. From the measured radon progeny concentration in raindrops the additional exposure rate due to rainfall has been evaluated and compared with the HPIC monitoring data.

A simplified method, based on the integral of the first collision kernel, is presented for performing gamma-ray skyshine calculations for the collimated sources. The first collision kernels were calculated in air for a reference air density by use of the EGS4 Monte Carlo code. These kernels can be applied to other air densities by applying density corrections. The integral first collision kernel (IFCK) method has been used to calculate two of the ANSI/ANS skyshine benchmark problems and the results were compared with a number of other commonly used codes. Our results were generally in good agreement with others but only spend a small fraction of the computation time required by the Monte Carlo calculations. The scheme of the IFCK method for dealing with lots of source collimation geometry is also presented in this study.

A simplified rainout model is presented in this study. According to this model formulations have been derived to relate the exposure rate on the ground to the radon progeny concentration in raindrops as well as the radon concentration in cloud air. By normalizing the saturation exposure rate of the calculation to the HPIC measured value and by the use of reasonably assumed values for some other parameters, the radon progeny concentration in raindrops as well as the radon concentration in cloud air has been determined.

The production of Thallium-201 via the 203 Tl(p,3n) 201 Pb→ 201 Tl process in the compact cyclotron TR30/15 of the Institute of Nuclear Energy Research (INER) was evaluated based on excitation functions. This study involves a selection of a series of parameters including proton energy, target thickness, and irradiation time that will maximize the yield of the product and minimize the radioactive impurities. The excitation functions of the 203 Tl(p, x n)Pb reactions have been calculated by using the ALICE-91 and FLUKA codes, respectively. Their results were compared with that recommended in the IAEA-TECDOC-1211 report. The discrepancies among them were identified and the resulting effects on the integral yields were discussed. Based on the recommended nuclear data, a user-friendly program has been developed to estimate the activities of the product and impurities during the production process of 201 Tl in INER’s compact cyclotron (INERCC). This helps in understanding the effects of various parameters selected in the production process. The results reveal that the current status of the production process of 201 Tl in INERCC appears to be quite optimal and the production yield is close to its theoretical prediction.

The ineffectiveness of the oven annealing for the LiF:Mg,Cu,P thermoluminescent dosimeter with a thick annealing tray was demonstrated to be due to the fact that the temperature achieved on the annealing tray was significantly lower than the oven preset temperature. The 240 °C/10 min annealing using a special-designed hot plate, for which the heating temperature on the thin annealing tray can be accurately controlled, has been presented to be capable to reduce the residual signal of the TLD-700H, TLD-7000, and TLD-2007 detectors to less than 0.02% and achieve a reproducibility for 10-cycle measurements of ∼0.36% when an optimal readout cycle has been applied.

The purpose of this study is to evaluate the impact of the tissue heterogeneity on the dose distribution for a single beam of the gamma knife. The EGS4 Monte Carlo code was used to simulate both depth and radial profiles of the radiation dose in homogeneous and heterogeneous phantoms, respectively. The results are compared with the dose distribution calculated using the mathematical model of GammaPlan®, the treatment planning system of the gamma knife. The skull and sinus heterogeneity were simulated by a TeflonTM shell and an air shell, respectively. It was found that the tissue heterogeneity caused significant perturbation on the absolute depth dose at the focus as well as on the depth-dose distribution near the phantom surface and/or at the interface but little effect on the radial dose distribution. The effect of the beam aperture on the depth-dose distribution was also investigated in this study.

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.