Institute of Nuclear Energy Research

Vol’no-Nadezhdinskoye, Russia

Institute of Nuclear Energy Research

Vol’no-Nadezhdinskoye, Russia
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PubMed | National Health Research Institute, Institute of Nuclear Energy Research and Taipei Veterans General Hospital
Type: Journal Article | Journal: Molecular and clinical oncology | Year: 2014

Rhenium-188 (

Cheng C.-C.,National Yang Ming University | Guan S.-S.,Institute of Nuclear Energy Research | Yang H.-J.,Institute of Nuclear Energy Research | Chang C.-C.,Taipei Medical University Hospital | And 3 more authors.
Journal of Biomedical Science | Year: 2016

Background: Hypoxia in tumor niche is one of important factors to start regeneration of blood vessels, leading to increase survival, proliferation, and invasion in cancer cells. Under hypoxia microenvironment, furthermore, steadily increased hypoxia-inducible factor-1aα (HIF-1aα) is observed, and can increase vascular endothelial growth factor (VEGF) expression and promote angiogenesis. Zinc protoporphyrin (ZnPP), a heme oxygenase-1 (HO-1) inhibitor, is potential to inhibit tumor proliferation and progression. However, the mechanism of ZnPP in inhibition of tumor is not completely clear. We hypothesize that ZnPP may modulate HIF-1aα through inhibiting HO-1, and then inhibit angiogenesis and tumor progression. This study aimed to dissect the mechanism of ZnPP in tumor suppression. Results: We observed the amount of VEGF was increased in the sera of the colorectal cancer (CRC) patients (n = 34, p < 0.05). Furthermore, increased VEGF expression was also measured in colorectal cancer cells, HCT-15, culturing under mimicking hypoxic condition. It suggested that hypoxia induced VEGF production from cancer cells. VEGF production was significantly reduced from HCT-15 cells after exposure to HIF-1aα inhibitor KC7F2, suggesting that HIF-1aα regulated VEGF production. Moreover, we observed that the HO-1inhibitor ZnPP inhibited the expressions of HIF-1aα and VEGF coupled with cell proliferations of HCT-15 cells, suggesting that ZnPP blocked HIF-1aα expression, and then inhibited the consequent VEGF production. In the xenograft model, we also observed that the animals exposed to ZnPP displayed much smaller tumor nodules and less degree of angiogenesis with decreased expression of the angiogenesis marker, aαvβ3 integrin, compared to that in normal control. Conclusions: This study demonstrated that VEGF level in serum was elevated in the patients with CRC. The HO-1 inhibitor, ZnPP, possessed the properties of anti-tumor agent by decreasing HIF-1aα levels, blocking VEGF production, impairing tumor angiogenesis, and inhibiting tumor growth. © 2016 Cheng et al.

Yang C.-W.,Institute of Nuclear Energy Research | Yang L.-C.,Institute of Nuclear Energy Research | Cheng T.-C.,Institute of Nuclear Energy Research | Jou Y.-T.,Chung Yuan Christian University | Chiou S.-W.,Chung Yuan Christian University
Nuclear Engineering and Design | Year: 2012

Computerized procedure (CP) system has been developed in nuclear power plant (NPP) instrumentation and control (I&C) system. The system may include normal operating procedures (OPs), abnormal operating procedures (AOPs), alarm response procedures (ARPs), surveillance test procedures (STPs) and/or emergency operating procedures (EOPs). While there are many ways to evaluate computerized procedures design, the user's mental workload and situation awareness (SA) are particularly important considerations in the supervisory control of safety-critical systems. Users' mental workload and situation awareness may be influenced by human factor issues relating to computerized procedures, e.g.; level of automation, dealing with (partially) unavailable I&C, switching to back-up system (e.g.; paper-based procedures). Some of the positive impacts of CPs on operator performance include the following: tasks can be performed more quickly; overall workload can be reduced; cognitive workload can be minimized; fewer errors may be made in transitioning through or between procedures. However, various challenges have also been identified with CP systems. These should be addressed in the design and implementation of CPs where they are applicable. For example, narrower "field of view" provided by CP systems than with paper-based procedures could reduce crew communications and crewmember awareness of the status and progress through the procedure. Based on a human factors experiment in which each participant monitored and controlled multiple simulated reactors, this study applied the NASA-TLX instrument for assessing mental workload. For the assessment of situation awareness (SA), the present research used the situational awareness rating technique (SART). In support of summarizing the results of user interface evaluation along multiple dimensions (e.g.; workload, SA), we propose advantages for the CPs compared to the paper-based procedures. © 2012 Elsevier B.V.

Tseng Y.-S.,Institute of Nuclear Energy Research | Wang J.-R.,Institute of Nuclear Energy Research | Tsai F.P.,Cool Tec Co. | Cheng Y.-H.,National Tsing Hua University | Shih C.,National Tsing Hua University
Annals of Nuclear Energy | Year: 2011

This study numerically investigated the thermal performance of a new tube-type dry-storage system (DSS) with 61 BWR spent nuclear fuels (SNFs) by utilizing the Computational Fluid Dynamics (CFD) code FLUENT. Through a minimized and necessary assumption of modeling process (e.g.; the lumped model of fuel assembly), a geometry model was employed to solve the problem of conjugate heat transfer coupled with thermal radiation. The simulation results show that the maximum temperature is 333 °C, and the minimum temperature margins are 81.5 °C and 12.3 °C for the fuel assembly and concrete structure, respectively. The results further demonstrate that the new tube-type DSS meets the thermal requirements in the NUREG-1536 guidelines and the temperature limitation of structure material. Finally, the CFD simulation can be a powerful tool for thermal-hydraulic analysis, which can provide useful information for design improving, such as the accuracy temperature values, location of hot spots in each component and the flow field characteristics.

Hung Z.-Y.,National Tsing Hua University | Huang Y.-K.,National Tsing Hua University | Pei B.-S.,National Tsing Hua University | Hsu W.-S.,National Tsing Hua University | Chen Y.-S.,Institute of Nuclear Energy Research
Kerntechnik | Year: 2015

The accident that occurred at the Fukushima Daiichi Nuclear Power Plant is a reminder of the danger of hydrogen explosion within a reactor building. Sufficiently high hydrogen concentration may cause an explosion that could damage the structure, resulting in the release of radioisotopes into the environment. In the first part of this study, a gas diffusion experiment was performed, in which helium was used as the working fluid. An analytical model was also developed using the GOTHIC code and the model predictions of the helium distribution were found to be in good agreement with the experimentally measured data. In the second part of the study, a model of the Mark HI containment of the Kuosheng Plant in Taiwan was developed, and was applied to a long-term station blackout (SBO) accident similar to that of the Fukushima plant. The hydrogen generation was calculated using the Modular Accident Analysis Program and was used as the boundary condition for the GOTHIC containment model. The simulation results revealed that the hydrogen concentration at the first floor of the wetwell in the containment reached 4% 9.7 h after the accident. This indicated the possibility of dangerous conditions inside the containment. Although active hydrogen ignitors are already installed in the Kuosheng plant, the findings of this study indicate that it may be necessary to add passive recombiners to prolong an SBO event. © Carl Hanser Verlag, München © Carl Hanser Verlag GmbH & Co. KG.

Tseng Y.-S.,Institute of Nuclear Energy Research | Wang J.-R.,Institute of Nuclear Energy Research | Lin C.-H.,Institute of Nuclear Energy Research | Shin C.,National Tsing Hua University | Tsai F.P.,Cool Tec Co.
American Society of Mechanical Engineers, Power Division (Publication) POWER | Year: 2010

Chinshan Nuclear Power Plant (CSNPP) is a two-unit BWR4 plant with 1804MWt power per unit. Taipower Co., the owner of the plant is preparing the life extension procedure to extend the CSNPP operation time. In order to meet the life extension requirement, many issues need to be solved before life extension licensing, such as the spent nuclear fuel management, structure aging etc. For the spent nuclear fuel management, ROC Atomic Energy Council (ROCAEC) certified method is employed to analyze the thermal behaviors of Dry Storage System (DSS). This method uses ANSYS coupled with RELAP5-3D to solve the thermal characteristic and successfully accomplish the licensing procedure of the Chinshan Nuclear Dry Storage Project. However, further validation results demonstrate that the coupled method still exists uncertainty and deficiency. In this study, a new Computational Fluid Dynamics (CFD) numerical model for spend nuclear fuel (NSF) dry storage system (DSS) has been developed to improve the accuracy of DSS thermal analysis results. Its accuracy has been validated by comparing the temperature predictions with the experimental results of VSC-17 DSS. It has been found that the thermal behaviors and physical phenomena in the DSS could be predicted with good agreement for the measurements. Moreover, the uncertainty and reasonableness of results in previous method can be improved by the new thermal analyses methodology. Copyright © 2010 by ASME.

Chen Y.-S.,Institute of Nuclear Energy Research | Yuann Y.-R.,Institute of Nuclear Energy Research | Dai L.-C.,Institute of Nuclear Energy Research
Nuclear Engineering and Design | Year: 2012

Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3°C, which is greater than the designed temperature of 171.1°C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity during the MSLB event can be maintained. © 2012 Elsevier B.V. All rights reserved.

Chen Y.-S.,Institute of Nuclear Energy Research | Yuann Y.-R.,Institute of Nuclear Energy Research | Dai L.-C.,Institute of Nuclear Energy Research | Lin Y.-P.,Institute of Nuclear Energy Research
Nuclear Engineering and Design | Year: 2011

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition. © 2011 Elsevier B.V. All rights reserved.

Liu C.-K.,Institute of Nuclear Energy Research | Lee R.-Y.,Institute of Nuclear Energy Research | Tsai K.-C.,Institute of Nuclear Energy Research | Wu S.-H.,Institute of Nuclear Energy Research | Lin K.-F.,Institute of Nuclear Energy Research of Taiwan
Ceramic Engineering and Science Proceedings | Year: 2014

The crystalline properties and thermal stabilities of a novel borosilicate glass (GC9) developed at INER, have been investigated for use as high-temperature seals in solid oxide fuel cells (SOFCs). The kinetics of isothermal and non-isothermal crystallization of the GC9 glass were examined by TG/DTA and XRD at various crystallization temperatures of 700~900°C and heating rates of 2.5~50 °C/min, respectively. A sandwich specimen of metallic interconnect and MEA joined by the GC9 glass was aged at 800°C for 1,000 hours to examine the compatibility and interfacial stability. High-temperature leak rate measurements of the GC9 glass were performed under the condition of aging (800°C) and thermal cycling (RT~800°C). The glass transition temperature, softening temperature, and coefficient of thermal expansion of the GC9 glass are 652°C, 745°C, and 12.5 ppm/°C, respectively. In the study, there was about 50% of fine ceramic phases, mainly Ba3La6(SiO4)6, embedded in the GC9 glass matrix and thus resulted in a superior mechanical strength at elevated temperatures. No obvious interfacial interaction or diffusion between the GC9 seals and the adjacent plates was found after long-term aging. Additionally, the average leakage rates were 2.25×I0-5 and 5.58×10-5 mbar·l/s/cm corresponding to the aging and thermal cycling tests. Copyright © 2015 by The American Ceramic Society.

Hsiao T.-Y.,National Tsing Hua University | Hsiao T.-Y.,Institute of Nuclear Energy Research | Lin C.,National Tsing Hua University
Progress in Nuclear Energy | Year: 2012

Once an event in a pressurized water reactor (PWR) occurs and the operator has identified it, the system may then be returned to safe operational status. Usually, monitoring the thermal state of the reactor, e.g.; the temperature and pressure of the reactor cooling system (RCS), allows the operator to recognize an event in progress. In this study, the temperature difference between hot and cold legs and the pressure of the steam generator (SG) are also adopted as the parameters relative to identification. The event data are generated by the best estimated model RELAP5. The cumulative distribution function (CDF) is constructed and the cumulative probability from 0.05 to 1.00, with 0.05 increments, are chosen for identification. Considering the measurement noise, the probabilistic approach is adopted. The random value is added to the simulated data, and then a 95% confidence interval is obtained. To identify an event, eighty data points, i.e.; twenty data points for each parameter and four parameters for each event, should match the stored data. Since the identification procedure is simple, the process is very fast. This method will be beneficial in the context of executing an emergency operating procedure more effectively. © 2012 Elsevier Ltd. All rights reserved.

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