The Experts below are selected from a list of 270 Experts worldwide ranked by ideXlab platform

Douglas B Kothe - One of the best experts on this subject based on the ideXlab platform.

  • modeling and simulation challenges pursued by the consortium for advanced simulation of Light Water Reactors casl
    Journal of Computational Physics, 2016
    Co-Authors: Paul J. Turinsky, Douglas B Kothe
    Abstract:

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics "core simulator" based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M&S capabilities, which is in progress, will assist in addressing long-standing and future operational and safety challenges of the nuclear industry. Complexity of physics based modeling of Light Water reactor cores being addressed.Capability developed to help address problems that have challenged the nuclear power industry.Simulation capabilities that take advantage of high performance computing developed.

  • casl the consortium for advanced simulation of Light Water Reactors
    Bulletin of the American Physical Society, 2010
    Co-Authors: Douglas B Kothe
    Abstract:

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear Reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of Light Water Reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized Water Reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of Reactors. CASL's mission is to more » develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance. « less

Isao Kataoka - One of the best experts on this subject based on the ideXlab platform.

  • review of thermal hydraulic researches in severe accidents in Light Water Reactors
    Journal of Nuclear Science and Technology, 2013
    Co-Authors: Isao Kataoka
    Abstract:

    The Tohoku Region Pacific Coast Earthquake and subsequent severe accident (SA) in Fukushima Daiichi Nuclear Power Station caused unprecedented disaster in Japan. Before this accident, considerable researches on SAs had been carried out in Japan. However, unfortunately, such researches could not prevent the accident due to the unexpected huge Tsunami. However, the researches on SAs become more and more important in order to make clear the causes of the accident in Fukushima and improve the safety of nuclear power plants in Japan. In view of this, review on researches on thermal hydraulics in SAs in Light Water Reactors was carried out. Important thermal-hydraulic phenomena in SAs were identified. Research activities on each phenomenon were surveyed mainly based on the articles published in Journal of Nuclear Science and Technology of Atomic Energy Society of Japan.

Raj Pathania - One of the best experts on this subject based on the ideXlab platform.

  • Irradiation Assisted Stress Corrosion Cracking (IASCC) of Nickel-Base Alloys in Light Water Reactors Environments Part II: Stress Corrosion Cracking
    Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, 2018
    Co-Authors: Mi Wang, Gary S. Was, L. Nelson, Miao Song, Raj Pathania
    Abstract:

    Radiation-induced microstructural changes control the Irradiation Assisted Stress Corrosion Cracking (IASCC)Irradiation-assisted stress corrosion cracking of core materials, which is a key factor in the extension of the operating lifetime of Light Water Reactors (LWRs). Nickel-base alloys are considered as potential structural materials to replace highly IASCC susceptible austenitic stainless steels. Constant extension rate tensile (CERT) tests were conducted on proton irradiated high strength nickel-base alloy 625 with two different heat treatment conditions (625Plus and 625DA) in both simulated BWR NWC and PWR primary Water. Crack length per unit area and fraction of grain boundaries that cracked were used to assess the IASCC susceptibility. Both 625Plus and 625DA showed a very high IASCC susceptibility. 625DA also exhibited greater changes in all microstructure features than 625Plus.

  • irradiation assisted stress corrosion cracking iascc of nickel base alloys in Light Water Reactors environments part ii stress corrosion cracking
    18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors 2017, 2017
    Co-Authors: Mi Wang, L. Nelson, Miao Song, Raj Pathania
    Abstract:

    Radiation-induced microstructural changes control the Irradiation Assisted Stress Corrosion Cracking (IASCC) of core materials, which is a key factor in the extension of the operating lifetime of Light Water Reactors (LWRs). Nickel-base alloys are considered as potential structural materials to replace highly IASCC susceptible austenitic stainless steels. Constant extension rate tensile (CERT) tests were conducted on proton irradiated high strength nickel-base alloy 625 with two different heat treatment conditions (625Plus and 625DA) in both simulated BWR NWC and PWR primary Water. Crack length per unit area and fraction of grain boundaries that cracked were used to assess the IASCC susceptibility. Both 625Plus and 625DA showed a very high IASCC susceptibility. 625DA also exhibited greater changes in all microstructure features than 625Plus.

  • Environmental controls for higher temperature direct-cycle Light Water Reactors
    Journal of Nuclear Materials, 2009
    Co-Authors: John Wilson, Raj Pathania, Samson Hettiarachchi
    Abstract:

    Radiolysis modeling is used to estimate the minimum hydrogen concentration to activate platinum catalysts and reduce the electrochemical corrosion potential in Light Water Reactors. Platinum catalysts are used in boiling Water Reactors to catalyze hydrogen and oxygen recombination, which reduces the corrosion potential and the susceptibility of austenitic structural materials to intergranular stress corrosion cracking. Two environmental challenges for material performance in higher temperature Light Water Reactors are the increased susceptibility of austenitic materials to stress corrosion cracking and the higher production rate of oxidizing radiolytic species. For a reference supercritical Water reactor, a hydrogen addition rate of 2 standard cubic feet per minute is needed to significantly reduce the susceptibility of austenitic materials to stress corrosion cracking. Also, for a reference higher temperature boiling Water reactor, a hydrogen addition rate of 10 standard cubic feet per minute of hydrogen reduces the stress corrosion crack susceptibility of austenitic materials located in the lower portion of the reactor vessel.

Mi Wang - One of the best experts on this subject based on the ideXlab platform.

S V Ivanova - One of the best experts on this subject based on the ideXlab platform.

  • hydrogen effected defects evolution in zirconium items of Light Water Reactors
    International Journal of Hydrogen Energy, 2006
    Co-Authors: S V Ivanova
    Abstract:

    Delayed hydride cracking (DHC) of the reactor core zirconium items of Light-Water Reactors and factors affecting this process (composition and structure state of alloy, hydrogenation level of zirconium item, direction of a defect evolution in it as well as conditions (temperature, stresses, irradiation) and regime (isothermal, thermal cycling) of operation were investigated. Influence of irradiation on DHC of zirconium items was studied. Possibility of evolution of the manufacturing defects (cracks) available in tubes of zirconium items in the process of their long storage under the effect of hydrogen available in tubes and residual stresses remained in them after fabrication was established.