Rong-Jiun Sheu

Evaluation of the root and root canal systems of mandibular first premolars in northern Taiwanese patients using cone-beam computed tomography

BACKGROUND/PURPOSE Cone-beam computed tomography (CBCT) can provide valuable data for root canal systems of human teeth in vivo. This study used CBCT to evaluate the number of roots and canals of 300 mandibular first premolars in 150 northern Taiwanese patients. METHODS The root canal systems of 300 mandibular first premolars in 150 northern Taiwanese patients with bilateral premolars were analyzed by CBCT. RESULTS Of the 300 mandibular first premolars, 197 (65.7%) had one root with one canal (1R1C), 49 (16.3%) had one root with two canals (1R2C), 51 (17.0%) had two roots with one canal in each root (2R2C), and three (1.0%) had three roots with one canal in each root (3R3C). Statistical analyses showed that women had a significantly higher incidence of 1R1C mandibular first premolars (71.4%) than men (58.8%, p = 0.031), and men had a significantly higher incidence of 2R2C mandibular first premolars (27.2%) than women (8.5%, p < 0.001). One hundred and twenty-two (81.3%) of the 150 patients had a symmetrical root and root canal system between the right and left mandibular first premolars. Men had a significantly higher symmetrical rate of 2R2C mandibular first premolars (26.5%) than women (8.2%, p = 0.013). CONCLUSION Approximately 82% of mandibular first premolars in northern Taiwanese patients have one root with either one or two canals. There are significant differences in the number of roots and canals and symmetry of the root canal system of bilateral mandibular first premolars between male and female northern Taiwanese patients.

Characteristics of Prompt Radiation Field and Shielding Design for Taiwan Photon Source

Abstract This study investigates the characteristics of the prompt radiation field due to the operation of the proposed Taiwan Photon Source (TPS). Two extreme beam loss cases are considered to bound the possible beam loss scenarios in the tunnel; i.e., all electrons are lost at one point, or they are lost uniformly along the whole electron orbit. Energy spectra and dose distributions of the prompt radiation field for the shielding design of the TPS are studied using analytic estimations and Monte Carlo simulations. The radiation levels of photons and neutrons outside the shielding wall are estimated for various operation modes and beam loss scenarios. The calculated results show that the preliminary shielding design of the TPS is highly practicable to achieve its annual design dose limit of 1 mSv for personnel. Meanwhile, the radiation impact on the environment is also far below the regulatory requirement.

Comparing standard Bonner spheres and high-sensitivity Bonner cylinders

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.

Study of the capability of energy dispersive small angle x-ray scattering with synchrotron radiation

Energy dispersive small angle x-ray scattering is a unique method that takes advantage of a continuous spectrum of a synchrotron radiation source. In this study, the capability of this scattering measurement is explored. Using a detector mask with three acceptance rings of different ring radius and a sample to detector distance of 440 mm with the usable x-ray energy ranging from 4 to 20 keV, we can cover a q range from 0.01 to 0.4 A−1. However, sample transmission and detector response function limit the application range at low energy. Slit scattering is the dominant source of background. The signal to background ratio for a standard cross-linked polyethylene sample can be more than 100. For a time-resolved experiment, data acquisition rate is limited by the throughput of the detector.

Robust artificial neural network for reliability and sensitivity analyses of complex non-linear systems

Artificial Neural Networks (ANNs) are commonly used in place of expensive models to reduce the computational burden required for uncertainty quantification, reliability and sensitivity analyses. ANN with selected architecture is trained with the back-propagation algorithm from few data representatives of the input/output relationship of the underlying model of interest. However, different performing ANNs might be obtained with the same training data as a result of the random initialization of the weight parameters in each of the network, leading to an uncertainty in selecting the best performing ANN. On the other hand, using cross-validation to select the best performing ANN based on the ANN with the highest R2 value can lead to biassing in the prediction. This is as a result of the fact that the use of R2 cannot determine if the prediction made by ANN is biased. Additionally, R2 does not indicate if a model is adequate, as it is possible to have a low R2 for a good model and a high R2 for a bad model. Hence, in this paper, we propose an approach to improve the robustness of a prediction made by ANN. The approach is based on a systematic combination of identical trained ANNs, by coupling the Bayesian framework and model averaging. Additionally, the uncertainties of the robust prediction derived from the approach are quantified in terms of confidence intervals. To demonstrate the applicability of the proposed approach, two synthetic numerical examples are presented. Finally, the proposed approach is used to perform a reliability and sensitivity analyses on a process simulation model of a UK nuclear effluent treatment plant developed by National Nuclear Laboratory (NNL) and treated in this study as a black-box employing a set of training data as a test case. This model has been extensively validated against plant and experimental data and used to support the UK effluent discharge strategy.

Cosmic-ray neutron simulations and measurements in Taiwan.

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.

Source terms and attenuation lengths for estimating shielding requirements or dose analyses of proton therapy accelerators.

Abstract 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.

Surface Dose Rate Calculations of a Spent-Fuel Storage Cask by Using MAVRIC and Its Comparison with SAS4 and MCNP

Abstract Based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology, MAVRIC is a new computational sequence in the SCALE6 code package that is designed to perform efficient Monte Carlo simulation for a complicated and difficult shielding problem. This study aimed to evaluate the performance of MAVRIC with the latest cross-section library in calculating the surface dose rates of a realistic spent-fuel storage cask. Detailed dose rate profiles over the cask side and top surfaces were calculated, and the results were compared with our previous work using SAS4 and MCNP. In order to duplicate the same source model, the MAVRIC code has been modified to accommodate a user-defined axial source distribution. The comparison among the three codes was evaluated in terms of their accuracies and computational efficiencies. For the gamma-ray sources, the MAVRIC-calculated results are more accurate than SAS4 and consistent with those predicted by the continuous-energy MCNP calculations. Meanwhile, its computational efficiencies are comparable to the performance of the TORT-coupled MCNP calculations. For the fuel neutron source, the MAVRIC calculation with broad-group cross sections cannot give satisfactory result, and its computational performance is also a factor of ˜10 less efficient than that of TORT-coupled MCNP. With a fine-group cross-section library, MAVRIC can provide a better prediction but still underestimates the surface dose rates of the cask by 15 to 30%.

Predicting induced radioactivity for the accelerator operations at the Taiwan Photon Source.

This study investigates the characteristics of induced radioactivity due to the operations of a 3-GeV electron accelerator at the Taiwan Photon Source (TPS). According to the beam loss analysis, the authors set two representative irradiation conditions for the activation analysis. The FLUKA Monte Carlo code has been used to predict the isotope inventories, residual activities, and remanent dose rates as a function of time. The calculation model itself is simple but conservative for the evaluation of induced radioactivity in a light source facility. This study highlights the importance of beam loss scenarios and demonstrates the great advantage of using FLUKA in comparing the predicted radioactivity with corresponding regulatory limits. The calculated results lead to the conclusion that, due to fairly low electron consumption, the radioactivity induced in the accelerator components and surrounding concrete walls of the TPS is rather moderate and manageable, while the possible activation of air and cooling water in the tunnel and their environmental releases are negligible.

EVALUATION OF RADIONUCLIDES IN CONCRETE SHIELDING FOR NUCLEAR POWER PLANT DECOMMISSIONING

Abstract 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.