The Experts below are selected from a list of 294 Experts worldwide ranked by ideXlab platform

Akio Yamamoto - One of the best experts on this subject based on the ideXlab platform.

  • Compression of Cross-Section Data Size for High-Resolution Core Analysis Using Dimensionality Reduction Technique
    Nuclear Science and Engineering, 2020
    Co-Authors: Masato Yamamoto, Tomohiro Endo, Akio Yamamoto
    Abstract:

    Compression of cross-section data used for high-resolution Core Analysis is performed using a dimensionality reduction technique based on the singular value decomposition (SVD) and low-rank approxi...

  • Subchannel void distribution correction model for the two-stage Core Analysis method in boiling water reactors
    Annals of Nuclear Energy, 2018
    Co-Authors: Takeshi Mitsuyasu, Motoo Aoyama, Akio Yamamoto
    Abstract:

    Abstract The two-stage Core Analysis method is widely used for BWR Core Analysis. The purpose of this study is to develop a subchannel void distribution correction model for the two-stage Core Analysis method using an assembly-based thermal-hydraulics calculation in the Core Analysis stage. The model assumes two kinds of subchannel void distribution gradients along with the two diagonal lines in the horizontal cross section of a BWR fuel assembly. The model appends and tabulates the difference of the subchannel perturbation condition from the base condition in the lattice physics, and evaluates the tilts within the 2D lattice physics scheme, and couples those results with 3D subchannel Analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. The developed model is evaluated using a heterogeneous and a small Core problem. The model gives a better power distribution compared with that of the authors’ previous model. As a result, the model can incorporate the subchannel effect into the current two-stage Core calculation method.

  • a coupling model for the two stage Core calculation method with subchannel Analysis for boiling water reactors
    Annals of Nuclear Energy, 2017
    Co-Authors: Takeshi Mitsuyasu, Motoo Aoyama, Akio Yamamoto
    Abstract:

    Abstract The two-stage Core Analysis method is widely used for BWR Core Analysis. The purpose of this study is to develop a Core Analysis model coupled with subchannel Analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the Core Analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel Analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted condition. The present model can incorporate the subchannel effect into the current two-stage Core calculation method.

  • A new technique for spectral interference correction on pin-by-pin BWR Core Analysis
    Journal of Nuclear Science and Technology, 2014
    Co-Authors: Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto
    Abstract:

    A new correction technique to capture the spectral interference effect on collapsed cross sections, which focuses on application to the pin-by-pin boiling water reactor (BWR) Core Analysis, is proposed. The spectral interference effect, which is caused by adjacent loadings of different types of fuel assemblies, has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. Variation of neutron leakage affects neutron spectrum and thus the neutron leakage is considered to be important to correct coarse-group cross sections used in Core calculations. We focus on the neutron leakage in each pin-cell and use it as a correction index (i.e., a leakage index (LI)), which is defined as the volume-averaged neutron leakage in a pin-cell. By utilizing the leakage index, we represent the variations of coarse-group cross sections as the linear combination of LIs. In order to verify and discuss the applicability of the present correction technique, two-dimensional benchmark ...

  • A macroscopic cross-section model for BWR pin-by-pin Core Analysis
    Journal of Nuclear Science and Technology, 2013
    Co-Authors: Tatsuya Fujita, Tomohiro Endo, Akio Yamamoto
    Abstract:

    A macroscopic cross-section model used in boiling water reactor (BWR) pin-by-pin Core Analysis is studied. In the pin-by-pin Core calculation method, pin-cell averaged cross sections are calculated for many combinations of Core state and depletion history variables and are tabulated prior to Core calculations. Variations of cross sections in a Core simulator are caused by two different phenomena (i.e. instantaneous and history effects). We treat them through the Core state variables and the exposure-averaged Core state variables, respectively. Furthermore, the cross-term effect among the Core state and the depletion history variables is considered. In order to confirm the calculation accuracy and discuss the treatment of the cross-term effect, the k-infinity and the pin-by-pin fission rate distributions in a single fuel assembly geometry are compared. Some cross-term effects could be negligible since the impacts of them are sufficiently small. However, the cross-term effects among the control rod history ...

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

  • RMC – A Monte Carlo code for reactor Core Analysis
    Annals of Nuclear Energy, 2015
    Co-Authors: Kan Wang, Ding She, Jingang Liang, Yishu Qiu, Jialong Sun, Xiao Fan
    Abstract:

    Abstract A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor Core Analysis on high-performance computing platforms. To meet the requirements of reactor Analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the efficiency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor Analysis problems with good performances.

  • rmc a monte carlo code for reactor Core Analysis
    International Conference on Supercomputing, 2014
    Co-Authors: Ka Wang, Ding She, Jingang Liang, Yishu Qiu
    Abstract:

    A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor Core Analysis on high-performance computing platforms. To meet the requirements of reactor Analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the effciency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor Analysis problems with good performances.

Deokjung Lee - One of the best experts on this subject based on the ideXlab platform.

  • VHTR Core Analysis with McCARD and DeCART codes for high temperature engineering test reactor benchmark
    Annals of Nuclear Energy, 2018
    Co-Authors: Jinsu Park, Tae Young Han, Deokjung Lee, Hyun Chul Lee
    Abstract:

    Abstract This paper presents a Very High Temperature Gas-Cooled Reactor (VHTR) Core Analysis of the McCARD and DeCART-2D with High Temperature Engineering Test Reactor (HTTR) benchmark problem. The numerical results of the HTTR benchmark problem are demonstrated, the capability for VHTR Core Analysis of McCARD is validated by using the experiment data, and that of DeCART-2D is verified by comparing the numerical results with those of McCARD. The detail HTTR benchmark problem and specification of Core geometry and material are also described. Also, a method of approximation for the complex geometry of the VHTR Core is described for modelling in computational code. In terms of the multiplication factor and temperature coefficient, the numerical results of McCARD are easily compared with the HTTR experiment data. Because the DeCART-2D can solve 2-dimensional (2D) geometry, the simulation of 2D HTTR fuel rod, block, and Core problem is performed with DeCART-2D and McCARD. The numerical results between them are around 200–400 pcm, meaning that DeCART-2D can be used to analyze VHTR Core.

  • On the diffusion coefficient calculation in two-step light water reactor Core Analysis
    Journal of Nuclear Science and Technology, 2017
    Co-Authors: Sooyoung Choi, Kord Smith, Hanjoo Kim, Taewoo Tak, Deokjung Lee
    Abstract:

    ABSTRACTThis paper presents consistent and rigorous accuracy assessments of various methods for calculating the diffusion coefficients in a two-step reactor Core Analysis of light water reactors (LWRs). The diffusion coefficients are significantly affected by the transport correction and critical spectrum calculations. There are various methods for the transport corrections (inflow/outflow/hybrid corrections) and critical spectrum calculations (B1/P1/CASMO-4E methods) so that it is necessary to decide the best combination to achieve a high accuracy in the transport/diffusion two-step Analysis. Numerical tests are performed step-by-step to search for the best combination of the methods by comparing each other the transport one-step results, transport/diffusion two-step results, and Monte Carlo results. Numerical test results with a large and a small LWR Core show that the combination of inflow transport correction and CASMO-4E critical spectrum calculation is most accurate than the other combinations in te...

  • Whole Core Analysis of molten salt breeder reactor with online fuel reprocessing
    International Journal of Energy Research, 2015
    Co-Authors: Jinsu Park, Hyun Chul Lee, Yongjin Jeong, Deokjung Lee
    Abstract:

    Summary The simulation of whole Core depletion and continuous reprocessing of a molten salt breeder reactor (MSBR) was performed. The MSBR model was built using MCNP6, and the depletion and reprocessing simulations were modeled using CINDER90 and a Python script. The Python script was introduced to implement online reprocessing of molten salt fuel and the feeding of new fertile material with 3-day depletion intervals during the simulations. The simulation started with the reference composition from the original Oak Ridge National Laboratory MSBR design. Equilibrium compositions were obtained from depletion and reprocessing simulations 7000 days worth of data. The MSBR whole Core Analysis was performed at the initial and equilibrium Core conditions, for various reactor design parameters such as multiplication factors, neutron flux distributions, temperature coefficients, rod worths, and power distributions. The neutronic Core characteristics were analyzed using a four-factor formula applied to the two zones of the Core separately. Copyright © 2015 John Wiley & Sons, Ltd.

  • Whole Core Analysis of molten salt breeder reactor with online fuel reprocessing
    International Journal of Energy Research, 2015
    Co-Authors: Jinsu Park, Hyun Chul Lee, Yongjin Jeong, Deokjung Lee
    Abstract:

    The simulation of whole Core depletion and continuous reprocessing of a molten salt breeder reactor (MSBR) was performed. The MSBR model was built using MCNP6, and the depletion and reprocessing simulations were modeled using CINDER90 and a Python script. The Python script was introduced to implement online reprocessing of molten salt fuel and the feeding of new fertile material with 3-day depletion intervals during the simulations. The simulation started with the reference composition from the original Oak Ridge National Laboratory MSBR design. Equilibrium compositions were obtained from depletion and reprocessing simulations 7000days worth of data. The MSBR whole Core Analysis was performed at the initial and equilibrium Core conditions, for various reactor design parameters such as multiplication factors, neutron flux distributions, temperature coefficients, rod worths, and power distributions. The neutronic Core characteristics were analyzed using a four-factor formula applied to the two zones of the Core separately. © 2015 John Wiley & Sons, Ltdclose0

Izaskun Zubizarreta - One of the best experts on this subject based on the ideXlab platform.

  • Example of a Core Analysis Programme
    Developments in Petroleum Science, 2015
    Co-Authors: Colin Mcphee, Jules Reed, Izaskun Zubizarreta
    Abstract:

    Abstract Poor planning and ill-considered programme design are the two principal reasons why so much Core Analysis data are unreliable or unrepresentative. This results in under-utilisation, poor appreciation and misapplication of the resultant Core data. Proper planning, design and supervision of a Core Analysis programme can do much to reduce the data redundancy rate. This chapter provides the key steps in planning and designing both routine Core Analysis (RCA) and special Core Analysis (SCAL) programmes and provides test recommendations. These can be used as a basis of design or template that might be adapted to design and formalise specific RCA and SCAL programmes for oil and gas reservoirs.

  • Best Practice in Coring and Core Analysis
    Developments in Petroleum Science, 2015
    Co-Authors: Colin Mcphee, Jules Reed, Izaskun Zubizarreta
    Abstract:

    Abstract Core Analysis provides the only direct and quantitative measurement of “intact” oil and gas reservoir properties. It should provide the foundation of formation evaluation for building static and dynamic reservoir models. However, it is estimated that approximately 70% of Core Analysis data are unfit for purpose as a consequence of ill-conceived test programme design, lack of planning, poor laboratory practice, inadequate reporting standards and unrepresentative test samples or test conditions. This chapter explains why a best practice guide to Core Analysis data acquisition is needed, and outlines a Core Analysis management framework to ensure effective quality control of test data and minimise data uncertainties. The objectives are to provide reservoir engineers, geoscientists and petrophysicists with a solid grounding in the acquisition, evaluation and implementation of reliable and representative routine Core Analysis and special Core Analysis data, and the essential knowledge to be able to judge the quality and reliability of Core Analysis data that are used in reservoir evaluation and characterisation.

  • Preparation for Special Core Analysis
    Developments in Petroleum Science, 2015
    Co-Authors: Colin Mcphee, Jules Reed, Izaskun Zubizarreta
    Abstract:

    Abstract Correct planning and preparation for special Core Analysis laboratory (SCAL) tests are fundamental to the acquisition of high-quality data. This chapter describes the methods and procedures to prepare and characterise test fluids used in SCAL (oil, water and gas); how to select and screen representative test plugs; and the methods and data used to determine the test confining stress that is most representative of reservoir effective stress conditions.

  • Chapter 5 – Routine Core Analysis
    Developments in Petroleum Science, 2015
    Co-Authors: Colin Mcphee, Jules Reed, Izaskun Zubizarreta
    Abstract:

    Abstract Routine (or basic or conventional) Core Analysis typically involves fluid saturation measurements and petrophysical measurements on dry plugs and samples at ambient or laboratory conditions. The data are principally used to characterise the reservoir properties and for log–Core integration. This chapter provides details of sample preparation, test equipment, test procedures and data reporting requirements for the principal RCA tests: fluid saturations (retort and Dean–Stark); helium porosity; and steady-state and unsteady-state nitrogen/helium permeability measurements. Darcy's law is derived for steady-state gas flow, and corrections for non-Darcy flow (Klinkenberg corrections and Forchheimer's effect) are described. The chapter also includes details of recommended best practice for single-phase liquid permeability, high-resolution probe permeability, and basic porosity and permeability measurements on whole Core samples. The advantages and drawbacks/issues associated with each test are described, and quality control checks and diagnostics are summarised.

  • chapter 5 routine Core Analysis
    Developments in Petroleum Science, 2015
    Co-Authors: Colin Mcphee, Jules Reed, Izaskun Zubizarreta
    Abstract:

    Abstract Routine (or basic or conventional) Core Analysis typically involves fluid saturation measurements and petrophysical measurements on dry plugs and samples at ambient or laboratory conditions. The data are principally used to characterise the reservoir properties and for log–Core integration. This chapter provides details of sample preparation, test equipment, test procedures and data reporting requirements for the principal RCA tests: fluid saturations (retort and Dean–Stark); helium porosity; and steady-state and unsteady-state nitrogen/helium permeability measurements. Darcy's law is derived for steady-state gas flow, and corrections for non-Darcy flow (Klinkenberg corrections and Forchheimer's effect) are described. The chapter also includes details of recommended best practice for single-phase liquid permeability, high-resolution probe permeability, and basic porosity and permeability measurements on whole Core samples. The advantages and drawbacks/issues associated with each test are described, and quality control checks and diagnostics are summarised.

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

  • rmc a monte carlo code for reactor Core Analysis
    International Conference on Supercomputing, 2014
    Co-Authors: Ka Wang, Ding She, Jingang Liang, Yishu Qiu
    Abstract:

    A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor Core Analysis on high-performance computing platforms. To meet the requirements of reactor Analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the effciency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor Analysis problems with good performances.