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The overabundance of gas molecules in the coldest regions of space points to a nonthermal desorption process. Laboratory simulations show the efficient desorption of CO ice that is exposed to ultraviolet radiation, known as photodesorption, which decreases with increasing ice deposition temperature. However, our understanding of this abnormal phenomenon still remains elusive. In this work, we show that the same phenomenon—and, in particular, a dramatic drop in the photodesorption yield—is observed when the deposition temperature is 19 K and higher. Also, the minimum ice thickness that accounts for a constant photodesorption yield of CO ice is dependent on the deposition temperature, an observation that is reported here for the first time. We propose that the key parameters dominating the absorbed photon energy transfer in CO ice, and contributing to the measured photodesorption yields, are the energy transfer length, the desorption yield contributed by a single ice layer, and the relative effective surface area. These parameters should be incorporated into astrophysical models that simulate the photodesorption of the top CO-rich ice layer in icy dust populations, with a size distribution that is related to the ice thickness.

Digital mammography has become a first-line diagnostic tool for clinical breast cancer screening due to its high sensitivity and specificity. Mammographic compression force is closely associated with image quality and patient comfort. Therefore, optimizing breast compression parameters is essential. Subjects were recruited for digital mammography and breast magnetic resonance imaging (MRI) within a month. Breast MRI images were used to calculate breast volume and volumetric breast density (VBD) and construct finite element models. Finite element analysis was performed to simulate breast compression. Simulated compressed breast thickness (CBT) was compared with clinical CBT and the relationships between compression force, CBT, breast volume, and VBD were established. Simulated CBT had a good linear correlation with the clinical CBT (R2 = 0.9433) at the clinical compression force. At 10, 12, 14, and 16 daN, the mean simulated CBT of the breast models was 5.67, 5.13, 4.66, and 4.26 cm, respectively. Simulated CBT was positively correlated with breast volume (r > 0.868) and negatively correlated with VBD (r < –0.338). The results of this study provides a subject-specific and evidence-based suggestion of mammographic compression force for radiographers considering image quality and patient comfort.

Because the Taiwan domestic boiling water reactor nuclear power plant is under shutdown and going to be decommissioned, the development of radioactive waste container used for containing the radioactive waste from nuclear facility decommissioning and dismantling is a critical issue in recent years. The main purpose of this study is to design a metal container which can provide the highest storage efficiency for the low level radioactive waste, and ensure the structure meeting the safety requirements of related regulations. This paper first introduces the related regulations for the container in Taiwan. To determine the optimal storage efficiency of the container, the shielding analyses were performed by varying dimensions of 3 cubic meter containers to select the best one, and conduct the structural analyses. Then, a series of structural analyses for operational conditions such as static standstill analysis, lifting test and stacking test with full load were performed, as well as the drop impact analysis, to ensure the container structure meeting the safety requirements of regulations. Present work can provide design information of the development of radioactive waste container for decommissioned nuclear facility in Taiwan.

The probabilistic fracture mechanics method is widely adopted in nuclear power plant industry, especially for the structural integrity assessment of reactor pressure vessels. In this work, the plant specific analyses of Taiwan's boiling water reactors, BWR/4 and BWR/6, under the design transients were performed by the FAVOR (Fracture Analysis of Vessel – Oak Ridge) code. The difference of reactor pressure vessel geometry, alloying elements, neutron fluence, and loading conditions were taken into account in the probabilistic fracture mechanics analyses. The failure probabilities of the axial welds at 64 effective full-power year (EFPY) for the analyzed BWR/4 and BWR/6 are less than 2.8 × 10−10/year and 1.5 × 10−4/year, respectively. Furthermore, the circumferential welds present much smaller failure probabilities. The results of this work are useful for the subsequent aging management of the operating reactor pressure vessels.

With the development of probabilistic fracture mechanics (PFM) methods in recent years, the risk-informed approach has gradually been used to evaluate the structural integrity and reliability of the reactor pressure vessels (RPV) in many countries. For boiling water reactor (BWR) pressure vessels, it has been demonstrated that it is not necessary to perform the inservice inspections of beltline circumferential welds to maintain the required safety margins because their probability of failure is orders of magnitude less than that of beltline vertical welds, thus may well reduce the associated substantial cost and person-rem exposure. In Taiwan, however, the inservice inspections of shell welds still have to be performed every ten years per ASME Boiler and Pressure Vessel Code, Section XI inspection requirements for a BWR type Chinshan nuclear power station. In this work, a very conservative PFM model of FAVOR code consistent with that USNRC used for regulation is built with the plant specific parameters concerning the beltline shell welds of RPVs of Chinshan nuclear power station. Meanwhile, a hypothetical transient of low temperature over-pressure (LTOP) event which challenges the BWR RPV integrity most severely is also assumed as the loading condition for conducting the PFM analyses. Further, the effects of performance of inservice inspection are also studied to determine the benefit of the costly inspection effort. The computed low probability of failure indicates that the analyzed RPVs can provide sufficient reliability even without performing any inservice inspection on the circumferential welds. It also indicates that performing the inservice inspections can not promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on RPV degradation and also be helpful for the regulation of BWR plants in Taiwan.

The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast neutron fluence exposure. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture analysis of vessels—Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, alloying elements, neutron fluence, and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08 has the probability of failure. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region and provide the aging analysis results for the life extension and the license renewal applications.

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.Copyright © 2012 by ASME

The chemistry concentration uncertainty of cooper and nickel significantly affects the shift in reference nil-ductility transition temperature (ΔRTNDT ). The uncertainty comes from the methods and equipments applied in measurements, the lack of specimen in surveillance capsule, and the non-homogeneous of material. The variations of ΔRTNDT result in the differences of failure probability of reactor pressure vessel. In this study, the structural integrity of Chinshan boiling water reactor RPV shell welds was evaluated by probabilistic fracture mechanics code-Fracture Analysis of Vessel – Oak Ridge (FAVOR). The influence of chemistry concentration uncertainty on the fracture probability of Chinshan nuclear power plant RPV with 32 and 64 effective full power years (EFPY) operation was discussed. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region, and provide the aging assessment of reactor pressure vessel.Copyright © 2011 by ASME

The Chinshan boiling water reactor (BWR) units 1 and 2, owned by Taiwan Power Company (TPC), started commercial operations in 1978 and 1979, respectively. The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast-neutron fluence exposure. This effect should be considered in the life extension and license renewal application. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture Analysis of Vessels – Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, chemistry components, neutron fluence and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08, is found to have the probability of failure. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region, and provide the aging analysis results for the life extension and the license renewal applications.Copyright © 2010 by ASME