The Experts below are selected from a list of 312 Experts worldwide ranked by ideXlab platform

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

  • Sodium Fast Reactors with Closed Fuel Cycle
    2015
    Co-Authors: Baldev Raj, P. Chellapandi, P. R. Vasudeva Rao
    Abstract:

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled Fast Reactors with closed fuel cycles, this book: Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspects Features a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspection, and simulators Addresses design essentials with a focus on reactor assembly including core and coolant circuits, fuel handling, instrumentation and control, energy conversion, and containment systems Provides design codes and standards with sufficient background information to ensure a solid understanding of the underlying mechanics Supplies guidelines for concept selection, design, analysis, and validation Sodium Fast Reactors with Closed Fuel Cycle is a valuable reference for industry professionals involved in the construction of Fast-reactor power plants, as well as graduate-level engineering students of the design and development of sodium-cooled Fast-reactor systems and components.

  • Knowledge Management in Fast Reactors
    Energy Procedia, 2011
    Co-Authors: K. K. Kuriakose, S. A. V. Satya Murty, P. Swaminathan, Baldev Raj
    Abstract:

    Abstract This paper highlights the work that is being carried out in Knowledge Management of Fast Reactors at Indira Gandhi Centre for Atomic Research (IGCAR) including a few examples of how the knowledge acquired because of various incidents in the initial years has been utilized for the successful opera tion of Fast Breeder Test Reactor. It also briefly refers to the features of the IAEA initiative on the preservation of Knowledge in the area of Fast Reactors in the form of “Fast Reactor Knowledge Organization System” (FR-KOS), which is based on a taxonomy for storage and mining of Fast Reactor Knowledge.

  • current status of Fast Reactors and future plans in india
    Energy Procedia, 2011
    Co-Authors: S C Chetal, P. Chellapandi, P Puthiyavinayagam, S Raghupathy, V Balasubramaniyan, P Selvaraj, P Mohanakrishnan, Baldev Raj
    Abstract:

    Abstract In the Indian energy scenario projections for the future, the nuclear power through Fast Reactors is expected to play an important role of ∼ 20% of total installed capacity by 2052. Successful operation of 40 MWt/13 MWe capacity Fast Breeder Test Reactor(FBTR) since 1985, strong R&D executed in multidisciplinary domain backed up by manufacturing technology and construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) based on indigenous design have provided high confidence on the success of sodium cooled Fast reactor technology. PFBR is a pool type MoX fuelled reactor designed with 2 primary sodium pumps, 2 secondary loops, 8 single wall integrated once through steam generators, and a rectangular containment. PFBR is presently under advanced stage of construction. Beyond PFBR, it is planned to construct 6 more FBRs of 500 MWe capacity each. Towards this, a systematic roadmap has been drawn for improved economy and enhanced safety through a number of measures. Roadmap for necessary R&D and manufacturing technology has been well detailed. The major features incorporated are twin unit concept, plant life increased to 60 years in comparison to 40 years for PFBR, reduction in number of steam generators from 8 to 6, reduction in special steel specific weight requirements, integrated primary sodium purification, enhanced reliability of shutdown systems, enhanced diversity in decay heat removal systems, enhanced in -service inspection, and compact plant layout. Beyond 2025, a series of 1000 MWe capacity metallic fuel with higher breeding potential are planned. R&D activities have been systematically formulated for metallic fuel development of both sodium bonded and mechanical bonded design. The paper addresses the highlights of current operating experience of FBTR and its life extension, construction status of PFBR, and design features of future sodium cooled Fast Reactors in India.

  • Ferritic steels for sodium-cooled Fast Reactors: Design principles and challenges
    JOM, 2010
    Co-Authors: Baldev Raj, M. Vijayalakshmi
    Abstract:

    An overview of the current status of development of ferritic steels for emerging Fast reactor technologies is presented in this paper. The creep-resistant 9–12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of Fast Reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the Fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  • Materials science research for sodium cooled Fast Reactors
    Bulletin of Materials Science, 2009
    Co-Authors: Baldev Raj
    Abstract:

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled Fast Reactors towards extending the life of Reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing Fast reactor electricity at an affordable and competitive price.

Suizheng Qiu - One of the best experts on this subject based on the ideXlab platform.

  • numerical approach to study the thermal hydraulic characteristics of reactor vessel cooling system in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2019
    Co-Authors: Ping Song, Dalin Zhang, Jing Chen, Xinan Wang, Shibao Wang, Yapei Zhang, Tangtao Feng, Xiuli Xue, Mingjun Wang, Suizheng Qiu
    Abstract:

    Abstract The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled Fast Reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled Fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled Fast reactor.

  • the development and validation of the inter wrapper flow model in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2018
    Co-Authors: Nina Yue, Dalin Zhang, Jing Chen, Ping Song, Xinan Wang, Shibao Wang, Suizheng Qiu, Yapei Zhang
    Abstract:

    Abstract The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled Fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental Fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled Fast Reactors.

Ping Song - One of the best experts on this subject based on the ideXlab platform.

  • numerical approach to study the thermal hydraulic characteristics of reactor vessel cooling system in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2019
    Co-Authors: Ping Song, Dalin Zhang, Jing Chen, Xinan Wang, Shibao Wang, Yapei Zhang, Tangtao Feng, Xiuli Xue, Mingjun Wang, Suizheng Qiu
    Abstract:

    Abstract The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled Fast Reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled Fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled Fast reactor.

  • the development and validation of the inter wrapper flow model in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2018
    Co-Authors: Nina Yue, Dalin Zhang, Jing Chen, Ping Song, Xinan Wang, Shibao Wang, Suizheng Qiu, Yapei Zhang
    Abstract:

    Abstract The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled Fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental Fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled Fast Reactors.

Yapei Zhang - One of the best experts on this subject based on the ideXlab platform.

  • numerical approach to study the thermal hydraulic characteristics of reactor vessel cooling system in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2019
    Co-Authors: Ping Song, Dalin Zhang, Jing Chen, Xinan Wang, Shibao Wang, Yapei Zhang, Tangtao Feng, Xiuli Xue, Mingjun Wang, Suizheng Qiu
    Abstract:

    Abstract The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled Fast Reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled Fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled Fast reactor.

  • the development and validation of the inter wrapper flow model in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2018
    Co-Authors: Nina Yue, Dalin Zhang, Jing Chen, Ping Song, Xinan Wang, Shibao Wang, Suizheng Qiu, Yapei Zhang
    Abstract:

    Abstract The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled Fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental Fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled Fast Reactors.

Dalin Zhang - One of the best experts on this subject based on the ideXlab platform.

  • numerical approach to study the thermal hydraulic characteristics of reactor vessel cooling system in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2019
    Co-Authors: Ping Song, Dalin Zhang, Jing Chen, Xinan Wang, Shibao Wang, Yapei Zhang, Tangtao Feng, Xiuli Xue, Mingjun Wang, Suizheng Qiu
    Abstract:

    Abstract The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled Fast Reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled Fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled Fast reactor.

  • the development and validation of the inter wrapper flow model in sodium cooled Fast Reactors
    Progress in Nuclear Energy, 2018
    Co-Authors: Nina Yue, Dalin Zhang, Jing Chen, Ping Song, Xinan Wang, Shibao Wang, Suizheng Qiu, Yapei Zhang
    Abstract:

    Abstract The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled Fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental Fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled Fast Reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled Fast Reactors.