The Experts below are selected from a list of 111 Experts worldwide ranked by ideXlab platform

S. P. Kelley - One of the best experts on this subject based on the ideXlab platform.

  • Evidence for excess argon during high Pressure metamorphism in the dora maira massif (Western Alps, Italy), using an ultra-violet laser ablation microprobe ^40Ar-^39Ar technique
    Contributions to Mineralogy and Petrology, 1995
    Co-Authors: N. O. Arnaud, S. P. Kelley
    Abstract:

    Ultra-high Pressure eclogite/amphibolite grade metamorphism of the Dora Maira Massif in the western Alps is a well established and intensively studied Event. However, the age of peak metamorphism and early cooling remains controversial. The ^40Ar-^39 Ar step-heating and laser spot ages from high Pressure phengites yield plateau ages as old as 110 Ma which have been interpreted as the time of early cooling after the high Pressure Event. Recent U/Pb and Sm/Nd results challenge this assertion, indicating a much younger age for the Event, around 45 Ma, and hence a radically different timing for the tectonic evolution of the western Alps. In a new approach to the problem, samples from the undeformed Hercynian metagranite, Brossasco, were studied using an ultra-violet laser ablation microprobe technique for ^40Ar-^39 Ar dating. The new technique allowed selective in situ analysis, at a spatial resolution of 50 urn, of quartz, phengite, biotite and K-feldspar. The results demonstrate the frequent occurrence of excess argon with high ^40Ar-^39 Ar ratios (1000–10000) and a strong relationship between apparent ages and metamorphic textures. The highest excess argon ratios are always associated with high closure temperature minerals or large diffusion domains within single mineral phases. The best interpretation of this relationship seems to be that excess argon was incorporated in all phases during the high Pressure Event, then mixed with an atmospheric component during rapid cooling and retrogression, producing a wide range of argon concentrations and ^40Ar/^36Ar ratios. Step-heating analysis of minerals with this mixture would produce linear arrays on a^40Ar/^36Ar versus ^40Ar/^36 Ar correlation diagram, leading to geologically meaningless plateau ages, older than the true closure age. In the present case, some ages in the range 60–110 Ma could be explained by the presence of excess argon incorporated around 40–50 Ma ago. Similar results found in other high Pressure terrains in the Alps may reconcile the argon geochronometer with other systems such as Rb/Sr, U/Pb or Sm/Nd. This study therefore calls for an increasing use of high resolution in situ sampling techniques to clarify the meaning of ^40Ar/ ^36 Ar ages in many high Pressure terrains.

Boyi Chen - One of the best experts on this subject based on the ideXlab platform.

  • Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event
    Volume 7: Operations Applications and Components, 2012
    Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi Chen
    Abstract:

    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

  • probabilistic fracture analysis for boiling water reactor Pressure vessels subjected to low temperature over Pressure Event
    Annals of Nuclear Energy, 2012
    Co-Authors: Chincheng Huang, Hsoungwei Chou, Boyi Chen
    Abstract:

    Abstract This paper performs the probabilistic fracture analyses for boiling water reactor Pressure vessels subjected to the low temperature over-Pressure Event. Using the FAVOR code, which was developed by the Oak Ridge National Laboratory in America, a conservative model with the plant specific data of domestic BWRs in Taiwan to the USNRC requirements is built. A hypothetical transient of low temperature over-Pressure Event which could severely challenge the integrity of reactor Pressure vessel of BWR is assumed as the loading condition for the probabilistic fracture mechanics analyses. The effects of performing inservice inspection on the probability of failure are also simulated. The computed low probability of failure indicates that the analyzed reactor Pressure vessel can maintain sufficient reliability even without performing any inservice inspections for the circumferential welds. It also indicates that extensive inservice inspections on shell welds cannot promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on reactor Pressure vessel degradation. The developed probabilistic fracture mechanics model is helpful for the subsequent BWR regulation in Taiwan.

  • Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event
    ASME 2010 Pressure Vessels and Piping Conference: Volume 7, 2010
    Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi Chen
    Abstract:

    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.© 2010 ASME

N. O. Arnaud - One of the best experts on this subject based on the ideXlab platform.

  • Evidence for excess argon during high Pressure metamorphism in the dora maira massif (Western Alps, Italy), using an ultra-violet laser ablation microprobe ^40Ar-^39Ar technique
    Contributions to Mineralogy and Petrology, 1995
    Co-Authors: N. O. Arnaud, S. P. Kelley
    Abstract:

    Ultra-high Pressure eclogite/amphibolite grade metamorphism of the Dora Maira Massif in the western Alps is a well established and intensively studied Event. However, the age of peak metamorphism and early cooling remains controversial. The ^40Ar-^39 Ar step-heating and laser spot ages from high Pressure phengites yield plateau ages as old as 110 Ma which have been interpreted as the time of early cooling after the high Pressure Event. Recent U/Pb and Sm/Nd results challenge this assertion, indicating a much younger age for the Event, around 45 Ma, and hence a radically different timing for the tectonic evolution of the western Alps. In a new approach to the problem, samples from the undeformed Hercynian metagranite, Brossasco, were studied using an ultra-violet laser ablation microprobe technique for ^40Ar-^39 Ar dating. The new technique allowed selective in situ analysis, at a spatial resolution of 50 urn, of quartz, phengite, biotite and K-feldspar. The results demonstrate the frequent occurrence of excess argon with high ^40Ar-^39 Ar ratios (1000–10000) and a strong relationship between apparent ages and metamorphic textures. The highest excess argon ratios are always associated with high closure temperature minerals or large diffusion domains within single mineral phases. The best interpretation of this relationship seems to be that excess argon was incorporated in all phases during the high Pressure Event, then mixed with an atmospheric component during rapid cooling and retrogression, producing a wide range of argon concentrations and ^40Ar/^36Ar ratios. Step-heating analysis of minerals with this mixture would produce linear arrays on a^40Ar/^36Ar versus ^40Ar/^36 Ar correlation diagram, leading to geologically meaningless plateau ages, older than the true closure age. In the present case, some ages in the range 60–110 Ma could be explained by the presence of excess argon incorporated around 40–50 Ma ago. Similar results found in other high Pressure terrains in the Alps may reconcile the argon geochronometer with other systems such as Rb/Sr, U/Pb or Sm/Nd. This study therefore calls for an increasing use of high resolution in situ sampling techniques to clarify the meaning of ^40Ar/ ^36 Ar ages in many high Pressure terrains.

Hsoungwei Chou - One of the best experts on this subject based on the ideXlab platform.

  • Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event
    Volume 7: Operations Applications and Components, 2012
    Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi Chen
    Abstract:

    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

  • probabilistic fracture analysis for boiling water reactor Pressure vessels subjected to low temperature over Pressure Event
    Annals of Nuclear Energy, 2012
    Co-Authors: Chincheng Huang, Hsoungwei Chou, Boyi Chen
    Abstract:

    Abstract This paper performs the probabilistic fracture analyses for boiling water reactor Pressure vessels subjected to the low temperature over-Pressure Event. Using the FAVOR code, which was developed by the Oak Ridge National Laboratory in America, a conservative model with the plant specific data of domestic BWRs in Taiwan to the USNRC requirements is built. A hypothetical transient of low temperature over-Pressure Event which could severely challenge the integrity of reactor Pressure vessel of BWR is assumed as the loading condition for the probabilistic fracture mechanics analyses. The effects of performing inservice inspection on the probability of failure are also simulated. The computed low probability of failure indicates that the analyzed reactor Pressure vessel can maintain sufficient reliability even without performing any inservice inspections for the circumferential welds. It also indicates that extensive inservice inspections on shell welds cannot promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on reactor Pressure vessel degradation. The developed probabilistic fracture mechanics model is helpful for the subsequent BWR regulation in Taiwan.

  • Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event
    ASME 2010 Pressure Vessels and Piping Conference: Volume 7, 2010
    Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi Chen
    Abstract:

    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.© 2010 ASME

Chincheng Huang - One of the best experts on this subject based on the ideXlab platform.

  • Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event
    Volume 7: Operations Applications and Components, 2012
    Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi Chen
    Abstract:

    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

  • probabilistic fracture analysis for boiling water reactor Pressure vessels subjected to low temperature over Pressure Event
    Annals of Nuclear Energy, 2012
    Co-Authors: Chincheng Huang, Hsoungwei Chou, Boyi Chen
    Abstract:

    Abstract This paper performs the probabilistic fracture analyses for boiling water reactor Pressure vessels subjected to the low temperature over-Pressure Event. Using the FAVOR code, which was developed by the Oak Ridge National Laboratory in America, a conservative model with the plant specific data of domestic BWRs in Taiwan to the USNRC requirements is built. A hypothetical transient of low temperature over-Pressure Event which could severely challenge the integrity of reactor Pressure vessel of BWR is assumed as the loading condition for the probabilistic fracture mechanics analyses. The effects of performing inservice inspection on the probability of failure are also simulated. The computed low probability of failure indicates that the analyzed reactor Pressure vessel can maintain sufficient reliability even without performing any inservice inspections for the circumferential welds. It also indicates that extensive inservice inspections on shell welds cannot promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on reactor Pressure vessel degradation. The developed probabilistic fracture mechanics model is helpful for the subsequent BWR regulation in Taiwan.

  • Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event
    ASME 2010 Pressure Vessels and Piping Conference: Volume 7, 2010
    Co-Authors: Hsoungwei Chou, Chincheng Huang, Boyi Chen
    Abstract:

    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.© 2010 ASME